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Sample records for irradiated materials similarly

  1. Microstructure of irradiated materials

    International Nuclear Information System (INIS)

    Robertson, I.M.

    1995-01-01

    The focus of the symposium was on the changes produced in the microstructure of metals, ceramics, and semiconductors by irradiation with energetic particles. the symposium brought together those working in the different material systems, which revealed that there are a remarkable number of similarities in the irradiation-produced microstructures in the different classes of materials. Experimental, computational and theoretical contributions were intermixed in all of the sessions. This provided an opportunity for these groups, which should interact, to do so. Separate abstracts were prepared for 58 papers in this book

  2. Materials modified by irradiation

    International Nuclear Information System (INIS)

    Chmielewski, A.G.

    2007-01-01

    Application of radiation in pharmaceutical sciences and cosmetology, polymer materials, food industry, environment, health camre products and packing production is described. Nano-technology is described more detailed, because it is less known as irradiation using technology. Economic influence of the irradiation on the materials value addition is shown

  3. Analysis of irradiated materials

    International Nuclear Information System (INIS)

    Bellamy, B.A.

    1988-01-01

    Papers presented at the UKAEA Conference on Materials Analysis by Physical Techniques (1987) covered a wide range of techniques as applied to the analysis of irradiated materials. These varied from reactor component materials, materials associated with the Authority's radwaste disposal programme, fission products and products associated with the decommissioning of nuclear reactors. An invited paper giving a very comprehensive review of Laser Ablation Microprobe Mass Spectroscopy (LAMMS) was included in the programme. (author)

  4. NSUF Irradiated Materials Library

    Energy Technology Data Exchange (ETDEWEB)

    Cole, James Irvin [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The Nuclear Science User Facilities has been in the process of establishing an innovative Irradiated Materials Library concept for maximizing the value of previous and on-going materials and nuclear fuels irradiation test campaigns, including utilization of real-world components retrieved from current and decommissioned reactors. When the ATR national scientific user facility was established in 2007 one of the goals of the program was to establish a library of irradiated samples for users to access and conduct research through competitively reviewed proposal process. As part of the initial effort, staff at the user facility identified legacy materials from previous programs that are still being stored in laboratories and hot-cell facilities at the INL. In addition other materials of interest were identified that are being stored outside the INL that the current owners have volunteered to enter into the library. Finally, over the course of the last several years, the ATR NSUF has irradiated more than 3500 specimens as part of NSUF competitively awarded research projects. The Logistics of managing this large inventory of highly radioactive poses unique challenges. This document will describe materials in the library, outline the policy for accessing these materials and put forth a strategy for making new additions to the library as well as establishing guidelines for minimum pedigree needed to be included in the library to limit the amount of material stored indefinitely without identified value.

  5. Irradiation environment and materials behavior

    International Nuclear Information System (INIS)

    Ishino, Shiori

    1992-01-01

    Irradiation environment is unique for materials used in a nuclear energy system. Material itself as well as irradiation and environmental conditions determine the material behaviour. In this review, general directions of research and development of materials in an irradiation environment together with the role of materials science are discussed first, and then recent materials problems are described for energy systems which are already existing (LWR), under development (FBR) and to be realized in the future (CTR). Topics selected are (1) irradiation embrittlement of pressure vessel steels for LWRs, (2) high fluence performance of cladding and wrapper materials for fuel subassemblies of FBRs and (3) high fluence irradiation effects in the first wall and blanket structural materials of a fusion reactor. Several common topics in those materials issues are selected and discussed. Suggestions are made on some elements of radiation effects which might be purposely utilized in the process of preparing innovative materials. (J.P.N.) 69 refs

  6. Irradiating strand material

    International Nuclear Information System (INIS)

    Austin, J.R.; Brown, M.J.; Loan, L.D.

    1975-01-01

    Conductors covered with insulation which is to be irradiated are passed between two groups of coaxial sheaves mounted rotatably individually. Successive sections of the conductors are advanced past the window of one accelerator head, around the associated sheave or sheaves, and then past the window of another accelerator head. The accelerators face in substantially opposite directions and are staggered along the paths of the conductors to avoid any substantial overlap of the electron beams associated therewith. The windows extend vertically to encompass all the generally horizontal passes of the conductors as between the two groups of sheaves. Preferably, conductors are strung-up between the sheaves in a modified figure eight pattern. The pattern is a figure eight modified to intermittently include a pass between the sheaves which is parallel to a line joining the axes of the two groups of sheaves. This reverses the direction of travel of the conductors and optimizes the uniformity of exposure of the cross sectional area of the insulation of the conductors to irradiation. The use of a figure eight path for the conductors causes the successive sections of the conductor to turn about the longitudinal axes thereof as they are advanced around the sheaves. In this way the insulation is more uniformly irradiated. In a preferred embodiment, twisted conductor pairs may be irradiated. The twist accentuates the longitudinal turning of the conductor pair. The irradiation of twisted pairs achieves obvious manufacturing economies while avoiding the necessity of having to twist irradiation cross-linked conductors

  7. Microstructural processes in irradiated materials

    Science.gov (United States)

    Byun, Thak Sang; Morgan, Dane; Jiao, Zhijie; Almer, Jonathan; Brown, Donald

    2016-04-01

    These proceedings contain the papers presented at two symposia, the Microstructural Processes in Irradiated Materials (MPIM) and Characterization of Nuclear Reactor Materials and Components with Neutron and Synchrotron Radiation, held in the TMS 2015, 144th Annual Meeting & Exhibition at Walt Disney World, Orlando, Florida, USA on March 15-19, 2015.

  8. Microstructural processes in irradiated materials

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Thak Sang; Morgan, Dane; Jiao, Zhijie; Almer, Jonathan; Brown, Donald

    2016-04-01

    This is an editorial article (preface) for the publication of symposium papers in the Journal of Nuclear materials: These proceedings contain the papers presented at two symposia, the Microstructural Processes in Irradiated Materials (MPIM) and Characterization of Nuclear Reactor Materials and Components with Neutron and Synchrotron Radiation, held in the TMS 2015, 144th Annual Meeting & Exhibition at Walt Disney World, Orlando, Florida, USA on March 15–19, 2015.

  9. Irradiation plant for flowable material

    International Nuclear Information System (INIS)

    Bosshard, E.

    1975-01-01

    The irradiation plant can be used to treat various flowable materials including effluent or sewage sludge. The plant contains a concrete vessel in which a partition is mounted to form two coaxial irradiation chambers through which the flowable material can be circulated by means of an impeller. The partition can be formed to house tubes of radiation sources and to provide a venturi-like member about the impeller. The operation of the impeller is reversed periodically to assure movement of both heavy and light particles in the flow. (U.S.)

  10. Materials irradiation research in neutron science

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-11-01

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

  11. Neutron irradiation effects on plasma facing materials

    Science.gov (United States)

    Barabash, V.; Federici, G.; Rödig, M.; Snead, L. L.; Wu, C. H.

    2000-12-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed.

  12. Neutron irradiation effects on plasma facing materials

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Roedig, M.; Snead, L.L.; Wu, C.H.

    2000-01-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed

  13. Microstructural processes in irradiated materials

    Science.gov (United States)

    Byun, Thak Sang; Kaoumi, Djamel; Bai, Xian-Ming

    2017-12-01

    The 8th symposium on Microstructural Progresses in Irradiated Materials (MPIM) was held at San Diego Convention Center and Marriott Marquis & Marina, San Diego, California, USA, February 26-March 2, 2017, as part of the TMS 2017 146th Annual Meeting and Exhibition. Since 2003, when the first MPIM symposium was held in the same place, the symposium has been held in odd years and has grown to one of the biggest symposia in the TMS Annual Meeting which invites more than sixty symposia. In the 8th MPIM symposium, a total of 106 oral and poster presentations, including 16 invited talks, were delivered for 4 days.

  14. Fusion Materials Irradiation Test Facility

    International Nuclear Information System (INIS)

    Kemp, E.L.; Trego, A.L.

    1979-01-01

    A Fusion Materials Irradiation Test Facility is being designed to be constructed at Hanford, Washington, The system is designed to produce about 10 15 n/cm-s in a volume of approx. 10 cc and 10 14 n/cm-s in a volume of 500 cc. The lithium and target systems are being developed and designed by HEDL while the 35-MeV, 100-mA cw accelerator is being designed by LASL. The accelerator components will be fabricated by US industry. The total estimated cost of the FMIT is $105 million. The facility is scheduled to begin operation in September 1984

  15. Research and materials irradiation reactors

    International Nuclear Information System (INIS)

    Ballagny, A.; Guigon, B.

    2004-01-01

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

  16. SEM analysis for irradiated materials

    International Nuclear Information System (INIS)

    Liu Xiaosong; Yao Liang

    2008-06-01

    A radiation-proof Scanning Electron Microscope (SEM) system is introduced. It has been widely used in various areas. For analyzing radioactive samples, normal SEM system needs lots of alterations. Based on KYKY-2800B SEM, the sample room, belt line, operating table and aerator were updated. New radiation-proof SEM system has used to analytic surface contaminated samples and RPV materials samples. An elementary means of SEM analysis for radioactive samples was studied, and this examination supported some available references for further irradiated fuel researches. (authors)

  17. High energy electron irradiation of flowable materials

    International Nuclear Information System (INIS)

    Offermann, B.P.

    1975-01-01

    In order to efficiently irradiate a flowable material with high energy electrons, a hollow body is disposed in a container for the material and the material is caused to flow in the form of a thin layer across a surface of the body from or to the interior of the container while the material flowing across the body surface is irradiated. (U.S.)

  18. Irradiated film material and method of the irradiation

    International Nuclear Information System (INIS)

    1978-01-01

    The irradiation of polymer film material is a strengthening procedure. To obtain a substantial uniformity in the radiation dosage profile, the film is irradiated in a trough having lateral deflection blocks adjacent to the film edges. These deflect the electrons towards the surface of the trough bottom for further deflection towards the film edge. (C.F.)

  19. Irradiation probe and laboratory for irradiated material evaluation

    International Nuclear Information System (INIS)

    Smutny, S.; Kupca, L.; Beno, P.; Stubna, M.; Mrva, V.; Chmelo, P.

    1975-09-01

    The survey and assessment are given of the tasks carried out in the years 1971 to 1975 within the development of methods for structural materials irradiation and of a probe for the irradiation thereof in the A-1 reactor. The programme and implementation of laboratory tests of the irradiation probe are described. In the actual reactor irradiation, the pulse tube length between the pressure governor and the irradiation probe is approximately 20 m, the diameter is 2.2 mm. Temperature reaches 800 degC while the pressure control system operates at 20 degC. The laboratory tests (carried out at 20 degC) showed that the response time of the pressure control system to a stepwise pressure change in the irradiation probe from 0 to 22 at. is 0.5 s. Pressure changes were also studied in the irradiation probe and in the entire system resulting from temperature changes in the irradiation probe. Temperature distribution in the body of the irradiation probe heating furnace was determined. (B.S.)

  20. Evaluation of irradiated coating material specimens

    International Nuclear Information System (INIS)

    Lee, Yong Jin; Nam, Seok Woo; Cho, Lee Moon

    2007-12-01

    Evaluation result of irradiated coating material specimens - Coating material specimens radiated Gamma Energy(Co 60) in air condition. - Evaluation conditions was above 1 X 10 4 Gy/hr, and radiated TID 2.0 X 10 6 Gy. - The radiated coating material specimens, No Checking, Cracking, Flaking, Delamination, Peeling and Blistering. - Coating system at the Kori no. 1 and APR 1400 Nuclear power plant, evaluation of irradiated coating materials is in accordance with owner's requirement(2.0 X 10 6 Gy)

  1. Low cycle fatigue of irradiated LMFBR materials

    International Nuclear Information System (INIS)

    Blackburn, L.D.

    1976-01-01

    A review of low cycle fatigue data on irradiated LMFBR materials was conducted and extensive graphical representations of available data are presented. Representative postirradiation tensile properties of annealed 304 and 316 SS are selected and employed in several predictive methods to estimate irradiated material fatigue curves. Experimental fatigue data confirm the use of predictive methods for establishing conservative design curves over the range of service conditions relevant to such CRBRP components as core former, fixed radial shielding, core barrel, lower inlet module and upper internals structures. New experimental data on fatigue curves and creep-fatigue interaction in irradiated 20 percent cold worked (CW) 316 SS and Alloy 718 would support the design of removable radial shielding and upper internals in CRBRP. New experimental information on notched fatigue behavior and cyclic stress-strain curves of all these materials in the irradiated condition could provide significant design data

  2. Mechanical properties of irradiated materials

    International Nuclear Information System (INIS)

    Robertson, I.M.; Robach, J.; Wirth, B.

    2001-01-01

    The effect of irradiation on the mechanical properties of metals is considered with particular attention being paid to the development of defect-free channels following uniaxial tensile loading. The in situ transmission electron microscope deformation technique is coupled with dislocation dynamic computer simulations to reveal the fundamental processes governing the elimination of defects by glissile dislocations. The observations of preliminary experiments are reported.(author)

  3. Structural material irradiations in FFTF

    International Nuclear Information System (INIS)

    1985-01-01

    Information is presented concerning the Materials Open Test Assembly (MOTA); instrumentation and control system; MOTA neutronic data; pressurized tube specimens; stress-rupture measurements for reactor materials; miniature specimen design; the Interim Examination and Maintenance (IEM) cell at the FFTF; support services; and general information concerning the FFTF

  4. Minimizing material damage using low temperature irradiation

    International Nuclear Information System (INIS)

    Craven, E.; Hasanain, F.; Winters, M.

    2012-01-01

    Scientific advancements in healthcare driven both by technological breakthroughs and an aging and increasingly obese population have lead to a changing medical device market. Complex products and devices are being developed to meet the demands of leading edge medical procedures. Specialized materials in these medical devices, including pharmaceuticals and biologics as well as exotic polymers present a challenge for radiation sterilization as many of these components cannot withstand conventional irradiation methods. The irradiation of materials at dry ice temperatures has emerged as a technique that can be used to decrease the radiation sensitivity of materials. The purpose of this study is to examine the effect of low temperature irradiation on a variety of polymer materials, and over a range of temperatures from 0 °C down to −80 °C. The effectiveness of microbial kill is also investigated under each of these conditions. The results of the study show that the effect of low temperature irradiation is material dependent and can alter the balance between crosslinking and chain scission of the polymer. Low temperatures also increase the dose required to achieve an equivalent microbiological kill, therefore dose setting exercises must be performed under the environmental conditions of use. - Highlights: ► A study is performed to quantify low temperature irradiation effects on polymer materials and BIs. ► Low temperature irradiation alters the balance of cross-linking and chain scissoning in polymers. ► Low temperatures provide radioprotection for BIs. ► Benefits of low temperatures are application specific and must be considered when dose setting.

  5. The construction of irradiated material examination facility

    International Nuclear Information System (INIS)

    Ro, Seung Gy; Lee, Key Soon; Herr, Young Hoi

    1990-03-01

    A detail design of the examination process, the hot cell facility and the annexed facility of the irradiated material examination facility (IMEF) which will be utilized to examine and evaluate physical and mechanical properties of neutron-irradiated materials, has been performed. Also a start-up work of the underground structure construction has been launched out. The project management and tasks required for the license application were duly carried out. The resultant detail design data will be used for the next step. (author)

  6. Polymeric materials obtained by electron beam irradiation

    International Nuclear Information System (INIS)

    Dragusin, M.; Moraru, R.; Martin, D.; Radoiu, M.; Marghitu, S.; Oproiu, C.

    1995-01-01

    Research activities in the field of electron beam irradiation of monomer aqueous solution to produce polymeric materials used for waste waters treatment, agriculture and medicine are presented. The technologies and special features of these polymeric materials are also described. The influence of the chemical composition of the solution to ba irradiated, absorbed dose level and absorbed dose rate level are discussed. Two kinds of polyelectrolytes, PA and PV types and three kinds of hydrogels, pAAm, pAAmNa and pNaAc types, the production of which was first developed with IETI-10000 Co-60 source and then adapted to the linacs built in Accelerator Laboratory, are described. (author)

  7. FMIT - the fusion materials irradiation test facility

    International Nuclear Information System (INIS)

    Liska, D.J.

    1980-01-01

    A joint effort by the Hanford Engineering Development Laboratory (HEDL) and Los Alamos Scientific Laboratory (LASL) has produced a preliminary design for a Fusion Materials Irradiation Test Facility (FMIT) that uses a high-power linear accelerator to fire a deuteron beam into a high-speed jet of molten lithium. The result is a continuous energy spectrum of neutrons with a 14-MeV average energy which can irradiate material samples to projected end-of-life levels in about 3 years, with a total accumulated fluence of 10 21 to 10 22 n/cm 2

  8. Post irradiation examinations on HTTR materials

    International Nuclear Information System (INIS)

    Sakai, Haruyuki; Ohmi, Masao; Eto, Motokuni; Watanabe, Katsutoshi

    1995-01-01

    The HTTR (High Temperature engineering Test Reactor) is being constructed at Oarai Research Establishment of the Japan Atomic Energy Research Institute. In order to develop necessary materials for the HTTR, after irradiations in the JMTR, PIEs are being carried out on these materials in the JMTRHL (JMTR Hot Laboratory). Impact test, tensile test, fatigue test, creep test, metallography and so on were performed for irradiated 2 1/4Cr 1Mo steel as the pressure vessel material and Alloy 800H as the cladding material of the control rod. A fatigue testing machine and four creep testing machines newly designed were fabricated and installed in the steel cells in order to evaluate the integrity of the HTTR materials. The development process and PIE results obtained with these machines are given in this paper

  9. Assessment of repair welding technologies of irradiated materials

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-08-15

    Damages of reactor internals of stainless steels caused by SCC and fatigue were identified in aged BWR plants. Repair-welding is one of the practical countermeasure candidates to restore the soundness of components and structures. The project of 'Assessment of Repair welding Technologies of Irradiated Materials' is being carried out to develop the technical guideline regarding the repair-welding of reactor internals. In fiscal 2011, we investigated the weldability of stainless steel 316L irradiated by welding (TIG) tungsten inert gas. Furthermore, the tensile properties and stress corrosion cracking (SCC) susceptibility of the welds were investigated. Cross-sectional observation of heat affected zone (HAZ) of the bead on plate TIG weldments (heat input 4 kJ/cm) of irradiated SUS316L stainless steel containing 0.026 ~ 0.12appm helium showed degradation of grain boundaries due to helium accumulation. Degree of the degradation depended on the amount of helium. No deterioration of grain boundaries was observed by bead on plate welding with one pass one layer when helium content was 0.039appm. The tensile strengths of welds in non-irradiated and irradiated material were similar. However, the elongation of a weldment by irradiated SUS316L containing 0.124appm Helium was lower than non-irradiated. It was estimated to cause the effects of helium bubbles. The SCC susceptibility of the HAZ was no significant difference compared with other locations. (author)

  10. Minimizing material damage using low temperature irradiation

    Science.gov (United States)

    Craven, E.; Hasanain, F.; Winters, M.

    2012-08-01

    Scientific advancements in healthcare driven both by technological breakthroughs and an aging and increasingly obese population have lead to a changing medical device market. Complex products and devices are being developed to meet the demands of leading edge medical procedures. Specialized materials in these medical devices, including pharmaceuticals and biologics as well as exotic polymers present a challenge for radiation sterilization as many of these components cannot withstand conventional irradiation methods. The irradiation of materials at dry ice temperatures has emerged as a technique that can be used to decrease the radiation sensitivity of materials. The purpose of this study is to examine the effect of low temperature irradiation on a variety of polymer materials, and over a range of temperatures from 0 °C down to -80 °C. The effectiveness of microbial kill is also investigated under each of these conditions. The results of the study show that the effect of low temperature irradiation is material dependent and can alter the balance between crosslinking and chain scission of the polymer. Low temperatures also increase the dose required to achieve an equivalent microbiological kill, therefore dose setting exercises must be performed under the environmental conditions of use.

  11. HFR irradiation testing of fusion materials

    International Nuclear Information System (INIS)

    Conrad, R.; von der Hardt, P.; Loelgen, R.; Scheurer, H.; Zeisser, P.

    1984-01-01

    The present and future role of the High Flux Reactor Petten for fusion materials testing has been assessed. For practical purposes the Tokamak-based fusion reactor is chosen as a point of departure to identify material problems and materials data needs. The identification is largely based on the INTOR and NET design studies, the reported programme strategies of Japan, the U.S.A. and the European Communities for technical development of thermonuclear fusion reactors and on interviews with several experts. Existing and planned irradiation facilities, their capabilities and limitations concerning materials testing have been surveyed and discussed. It is concluded that fission reactors can supply important contributions for fusion materials testing. From the point of view of future availability of fission testing reactors and their performance it appears that the HFR is a useful tool for materials testing for a large variety of materials. Prospects and recommendations for future developments are given

  12. Gamma irradiation technology for composite materials

    International Nuclear Information System (INIS)

    Romero, Guillermo R; Gonzalez, Maria E.

    2003-01-01

    A composite of sugar cane bagasse and low-density polyethylene was prepared. Gamma -radiation of Cobalt-60 (Co 60 ) and reactive additives were used, to make compatible the lignocellulosic fibers with the polymeric matrix. Gamma-radiation was applied in different stages with different purposes: a) Irradiation of cellulosic fibers treated or not with reactive additive, in presence of air, to produce macro radicals increasing their reactivity during extrusion with polyethylene. A homogeneous and fusible material resulted that can be used as raw material in thermoforming processes with cost in between that of its constitutive elements; b) Irradiation of final products, to produce the cross-linking of polymeric chains. The fibers remain trapped in the cross-linked matrix. A homogeneous and infusible material with high mechanical properties was obtained. (author)

  13. Comparison of material irradiation conditions for fusion, spallation, stripping and fission neutron sources

    International Nuclear Information System (INIS)

    Vladimirov, P.; Moeslang, A.

    2004-01-01

    Selection and development of materials capable of sustaining irradiation conditions expected for a future fusion power reactor remain a big challenge for material scientists. Design of other nuclear facilities either in support of the fusion materials testing program or for other scientific purposes presents a similar problem of irradiation resistant material development. The present study is devoted to an evaluation of the irradiation conditions for IFMIF, ESS, XADS, DEMO and typical fission reactors to provide a basis for comparison of the data obtained for different material investigation programs. The results obtained confirm that no facility, except IFMIF, could fit all user requirements imposed for a facility for simulation of the fusion irradiation conditions

  14. Self-organization in irradiated materials

    International Nuclear Information System (INIS)

    Gerasimenko, N.N.; Dzhamanbalin, K.K.; Medetov, N.A.

    2003-01-01

    Full text: By the present time a great deal of experimental material concerning self-organization in irradiated materials is stored. It means that in different materials (single crystal and amorphous semiconductor, metals, polymers) during one process of irradiation with accelerated particles or energetic quanta the structure previously disordered can be reordered to the previous or different order. These processes are considered separately from the processes of radiation-stimulated ordering when the renewal of the structure occurs as the result of extra irradiation, sometimes accompanied with another influence (heating, lighting, application of mechanical tensions). The processes of reordering are divided into two basic classes: the reconstruction of crystalline structure (1) and the formation of space-ordered system (2). The processes of ordering are considered with the use of synergetic approach and are analyzed conformably to the concrete conditions of new order appearance process realization in order to reveal the self-organization factor's role. The concrete experimental results of investigating of the radiation ordering processes are analyzed for different materials: semiconductor, metals, inorganic dielectrics, polymers. The ordering processes are examined from the point of their possible use in the technology of creating nano-dimensional structures general and quantum-dimensional ones in particular

  15. Materials Modification Under Ion Irradiation: JANNUS Project

    International Nuclear Information System (INIS)

    Serruys, Y.; Trocellier, P.; Ruault, M.-O.; Henry, S.; Kaietasov, O.; Trouslard, Ph.

    2004-01-01

    JANNUS (Joint Accelerators for Nano-Science and Nuclear Simulation) is a project designed to study the modification of materials using multiple ion beams and in-situ TEM observation. It will be a unique facility in Europe for the study of irradiation effects, the simulation of material damage due to irradiation and in particular of combined effects. The project is also intended to bring together experimental and modelling teams for a mutual fertilisation of their activities. It will also contribute to the teaching of particle-matter interactions and their applications. JANNUS will be composed of three accelerators with a common experimental chamber and of two accelerators coupled to a 200 kV TEM

  16. Relationship between swelling and irradiation creep in cold-worked PCA stainless steel irradiated to similar 178 dpa at similar 400 C

    International Nuclear Information System (INIS)

    Toloczko, M.B.; Garner, F.A.

    1994-01-01

    The eighth and final irradiation segment for pressurized tubes constructed from the fusion Prime Candidate Alloy (PCA) has been completed in FFTF. At 178 dpa and similar 400 C, the irradiation creep of 20% cold-worked PCA has become dominated by the ''creep disappearance'' phenomenon. The total diametral deformation rate has reached the limiting value of 0.33%/dpa at the three highest stress levels employed in this test. The stress-enhancement of swelling tends to camouflage the onset of creep disappearance, however, requiring the use of several non-traditional techniques to extract the creep coefficients. No failures occurred in these tubes, even though the swelling ranged from similar 20 to 40%. ((orig.))

  17. Relationship between swelling and irradiation creep in cold-worked PCA stainless steel irradiated to similar 178 dpa at similar 400 C

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B. (Department of Chemical and Nuclear Engineering, University of California, Santa Barbara, CA 93106 (United States)); Garner, F.A. (Pacific Northwest Laboratory, Richland, WA 99352 (United States))

    1994-09-01

    The eighth and final irradiation segment for pressurized tubes constructed from the fusion Prime Candidate Alloy (PCA) has been completed in FFTF. At 178 dpa and similar 400 C, the irradiation creep of 20% cold-worked PCA has become dominated by the creep disappearance'' phenomenon. The total diametral deformation rate has reached the limiting value of 0.33%/dpa at the three highest stress levels employed in this test. The stress-enhancement of swelling tends to camouflage the onset of creep disappearance, however, requiring the use of several non-traditional techniques to extract the creep coefficients. No failures occurred in these tubes, even though the swelling ranged from similar 20 to 40%. ((orig.))

  18. Similarity between the effects of carbon-ion irradiation and X-irradiation on the development of rat brain

    International Nuclear Information System (INIS)

    Inouye, Minoru; Hayasaka, Shizu; Murata, Yoshiharu; Takahashi, Sentaro; Kubota, Yoshihisa

    2000-01-01

    The effects of carbon-ion irradiation and X-irradiation on the development of rat brain were compared. Twenty pregnant rats were injected with bromodeoxyuridine (BrdU) at 9 pm on day 18 pregnancy and divided into five groups. Three hours after injection (day 19.0) one group was exposed to 290 MeV/u carbon-ion radiation by a single dose of 1.5 Gy. Other groups were exposed to X-radiation by 1.5, 2.0 or 2.5 Gy, or sham-treated, respectively. Fetuses were removed from one dam in each group 8 h after exposure and examined histologically. Extensive cell death was observed in the brain mantle from the irradiated groups. The cell death after 1.5 Gy carbon-ion irradiation was remarkably more extensive than that after 1.5 Gy X-irradiation, but comparable to that after 2.0 Gy or 2.5 Gy X-irradiation. The remaining rats were allowed to give birth and the offspring were sacrificed at 6 weeks of age. All of the irradiated offspring manifested microcephaly. The size of the brain mantle exposed to 1.5 Gy carbon-ion radiation was significantly smaller than that exposed to 1.5 Gy X-radiation and larger than that exposed to 2.5 Gy X-radiation. A histological examination of the cerebral cortex revealed that cortical layers II-IV were malformed. The defect by 1.5 Gy carbon-ion irradiation was more severe than that by the same dose of X-irradiation. Although the BrdU-incorporated neurons were greatly reduced in number in all irradiated groups, these cells reached the superficial area of the cortex. These findings indicated that the effects of both carbon-ion irradiation and X-irradiation on the development of rat brain are similar in character, and the effect of 1.5 Gy carbon-ion irradiation compares to that of 2.0-2.5 Gy X-irradiation. (author)

  19. Irradiation can for the activation of materials in nuclear reactors

    International Nuclear Information System (INIS)

    Schneider, B.; Findeisen, A.; Katzmann, H.

    1985-01-01

    The invention is concerning with an irradiation can for the activation of materials in nuclear reactors in particular for materials with a high heat generation due to irradiation. A good heat transfer between the irradiated material and the irradiation can environment has been guaranteed by a special can design. The outside of the can consists of a tube or a tube bandle which has been formed as a water guide tube. One or more tubes containing the irradiated materials have been positioned at the inner areas of the irradiated can

  20. Tests on irradiated magnet-insulator materials

    International Nuclear Information System (INIS)

    Schmunk, R.E.; Miller, L.G.; Becker, H.

    1983-01-01

    Fusion-reactor coils, located in areas where they will be only partially shielded, must be fabricated from materials which are as resistant to radiation as possible. They will probably incorporate resistive conductors with either water or cryogenic cooling. Inorganic insulators have been recommended for these situations, but the possibility exists that some organic insulators may be usuable as well. Results were previously reported for irradiation and testing of three glass reinforced epoxies: G-7, G-10, and G-11. Thin disks of these materials, nominally 0.5 mm thick by 11.1 mm diameter, were tested in compressive fatigue, a configuration and loading which represents reasonably well the magnet environment. In that work G-10 was shown to withstand repeated loading to moderately high stress levels without failure, and the material survived better at liquid nitrogen temperature than at room temperature

  1. Intense neutron irradiation facility for fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

    Technical R and D of d-Li stripping type neutron irradiation facilities for development of fusion reactor materials was carried out in Fusion Materials Irradiation Test Facility (FMIT) project and Energy Selective Neutron Irradiation Test Facility (ESNIT) program. Conceptual design activity (CDA) of International Fusion Materials Irradiation Facility (IFMIF), of which concept is an advanced version of FMIT and ESNIT concepts, are being performed. Progress of users` requirements and characteristics of irradiation fields in such neutron irradiation facilities, and outline of baseline conceptual design of IFMIF were described. (author)

  2. Metallographic examination in irradiated materials examination facility

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yong Sun; Lee, Key Soon; Park, Dae Gyu; Ahn, Sang Bok; Yoo, Byoung Ok

    1998-01-01

    It is very important to have equipment of metallographic examination in hot-cell to observe the micro-structure of nuclear fuels and materials irradiated at nuclear power and/or research reactor. Those equipment should be operated by master-slave manipulators, so they are designed, manufactured and modified to make exercise easy and no trouble. The metallographic examination equipment and techniques as well as its operation procedure are described, so an operator can practice the metallography in hot-cell. (author). 5 refs., 7 tabs., 21 figs.

  3. Studies on gamma irradiated rubber materials

    Science.gov (United States)

    Lungu, I. B.; Stelescu, M. D.; Cutrubinis, M.

    2018-01-01

    Due to the increase in use and production of polymer materials, there is a constant pressure of finding a solution to more environmental friendly composites. Beside the constant effort of recycling used materials, it seems more appropriate to manufacture and use biodegradable and renewable row materials. Natural polymers like starch, cellulose, lignin etc are ideal for preparing biodegradable composites. Some of the dynamic markets that use polymer materials are the food and pharmaceutical industries. Because of their desinfastation and sometimes sterility requirements, different treatment processes are applied, one of it being radiation treatment. The scope of this paper is to analyze the mechanical behaviour of rubber based materials irradiated with gamma rays at four medium doses, 30.1 kGy, 60.6 kGy, 91 kGy and 121.8 kGy. The objectives are the following: to identify the optimum radiation dose in order to obtain a good mechanical behaviour and to identify the mechanical behaviour of the material when adding different quantities of natural filler (20 phr, 60 phr and 100 phr).

  4. A new materials irradiation facility at the Kyoto university reactor

    International Nuclear Information System (INIS)

    Yoshiie, T.; Hayashi, Y.; Yanagita, S.; Xu, Q.; Satoh, Y.; Tsujimoto, H.; Kozuka, T.; Kamae, K.; Mishima, K.; Shiroya, S.; Kobayashi, K.; Utsuro, M.; Fujita, Y.

    2003-01-01

    A new materials irradiation facility with improved control capabilities has been installed at the Kyoto University Reactor (KUR). Several deficiencies of conventional fission neutron material irradiation systems have been corrected. The specimen temperature is controlled both by an electric heater and by the helium pressure in the irradiation tube without exposure to neutrons at temperatures different from the design test conditions. The neutron spectrum is varied by the irradiation position. Irradiation dose is changed by pulling the irradiation capsule up and down during irradiation. Several characteristics of the irradiation field were measured. The typical irradiation intensity is 9.4x10 12 n/cm 2 s (>0.1 MeV) and the irradiation temperature of specimens is controllable from 363 to 773 K with a precision of ±2 K

  5. AGC 2 Irradiated Material Properties Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-05-01

    The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

  6. Current investigations of packaging materials used for food irradiation

    International Nuclear Information System (INIS)

    Fiszer, W.

    1996-01-01

    The article reviews current investigations of packaging materials applied for food irradiation. The increasing role of various synthetic materials is described. Author reviews radiation-induced damages in these materials. The article includes the list of materials accepted for food packaging and subsequent irradiation with different doses

  7. Simulation of the welding of irradiated materials

    International Nuclear Information System (INIS)

    Lin, Hua Tay

    1989-07-01

    Helium was uniformly implanted using the ''tritium trick'' technique to levels of 0.18, 2.5, 27, 105 and 256 atomic part per million (appm) for type 316 stainless steel, and 0.3 and 1 appm for Sandvik HT-9 (12 Cr-1MoVW). Both full penetration as well as partial penetration welds were then produced on control and helium-containing materials using the autogenous gas tungsten arc (GTA) welding process under full constraint conditions. For full penetration welds, both materials were successfully welded when they contained less than 0.3 appm helium. However, welds of both materials, when containing greater than 1 appm helium, were found to develop cracks during cooling of the weld. Transmission and scanning electron microscopy indicated that the HAZ cracking was caused by the growth and coalescence of grain boundary (GB) helium bubbles. This cracking occurred as a result of the combination of high temperatures and high shrinkage tensile stresses. The cracking in the fusion zone was found to result from the precipitation of helium along dendrite interfaces. A model based on the kinetics of diffusive cavity growth is presented to explain the observed results. The model proposes a helium bubble growth mechanism which leads to final intergranular rupture in the heat-affected zone. Results of the present study demonstrate that the use of conventional fusion welding techniques to repair materials degraded by exposure to irradiation environments may be difficult if the irradiation results in the generation of helium equal to or greater than 1 appm

  8. Workshop on materials irradiation effects and applications 2012

    International Nuclear Information System (INIS)

    Xu, Qiu; Sato, Koichi; Yoshiie, Toshimasa

    2013-01-01

    For the study of the material irradiation effects, irradiation fields with improved control capabilities, advanced post irradiation experiments and well developed data analyses are required. This workshop aims to discuss new results and to plan the future irradiation research in the KUR. General meeting was held from December 14, 2012 to December 15, 2012 with 44 participants and 28 papers were presented. Especially recent experimental results using irradiation facilities in the KUR such as Materials Controlled Irradiation Facility, Low Temperature Loop and LINAC, and results of computer simulation, and fruitful discussions were performed. This volume contains the summary and selected transparencies presented in the meeting. (author)

  9. Irradiation damage behavior of low alloy steel wrought and weld materials

    International Nuclear Information System (INIS)

    Stofanak, R.J.; Poskie, T.J.; Li, Y.Y.; Wire, G.L.

    1993-01-01

    A study was undertaken to evaluate the irradiation damage response of several different types of low alloy steel: vintage type ASTM A302 Grade B (A302B) plates and welds containing different Ni and Cu concentrations, 3.5% Ni steels similar to ASTM A508 Class 4, welds containing about 1% Ni (similar to type 105S), and 3.5% Ni steels with ''superclean'' composition. All materials were irradiated at several different irradiation damage levels ranging from 0.0003 to 0.06 dpa at 232C (450F). Complete Charpy V-notch impact energy transition temperature curves were generated for all materials before and after irradiation to determine transition temperature at 4IJ (30 ft-lb) or 47J (35 ft-lb) and the upper shelf energy. Irradiation damage behavior was measured by shift in Charpy 41J or 47J transition temperature (ΔTT4 41J or ΔTT 47J ) and lowering of upper shelf Charpy energy at a given irradiation damage level. It was found that chemical composition greatly influenced irradiation damage behavior; highest irradiation damage (greatest ΔTT) was found in an A302B type weld containing 1.28% Ni and 0.20% Cu while the least damage was found in 3.5% Ni, 0.05% Cu, superclean wrought materials. Combination of Ni and Cu was found to affect irradiation damage behavior at higher irradiation damage levels in the A302B welds where the 1.28% Ni, 0.20% Cu weld showed more damage than a 0.60% Ni, 0.31% Cu weld. For the 3.5% Ni steels, fabrication influenced irradiation behavior in that a silicon (Si) killed material showed greater irradiation damage than a low silicon material. In general, the 3.5% Ni materials with low copper showed less irradiation damage than the A302B materials

  10. Modelling irradiation effects in fusion materials

    International Nuclear Information System (INIS)

    Victoria, M.; Dudarev, S.; Boutard, J.L.; Diegele, E.; Laesser, R.; Almazouzi, A.; Caturla, M.J.; Fu, C.C.; Kaellne, J.; Malerba, L.; Nordlund, K.; Perlado, M.; Rieth, M.; Samaras, M.; Schaeublin, R.; Singh, B.N.; Willaime, F.

    2007-01-01

    We review the current status of the European fusion materials modelling programme. We describe recent findings and outline potential areas for future development. Large-scale density functional theory (DFT) calculations reveal the structure of the point defects in α-Fe, and highlight the crucial part played by magnetism. The calculations give accurate migration energies of point defects and the strength of their interaction with He atoms. Kinetic models based on DFT results reproduce the stages of radiation damage recovery in iron, and stages of He-desorption from pre-implanted iron. Experiments aimed at validating the models will be carried out in the future using a multi-beam ion irradiation facility chosen for its versatility and rapid feedback

  11. Modelling irradiation effects in fusion materials

    Energy Technology Data Exchange (ETDEWEB)

    Victoria, M. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, c/Jose Gutierrez Abascal 2, 28006 Madrid (Spain); Dudarev, S. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Oxfordshire OX14 3DB, UK and Department of Physics, Imperial College, Exhibition Road, London SW7 2AZ (United Kingdom); Boutard, J.L. [EFDA-CSU Garching, Boltzmannstrasse 2, D-85748 Garching (Germany)], E-mail: jean-louis.boutard@tech.efda.org; Diegele, E.; Laesser, R. [EFDA-CSU Garching, Boltzmannstrasse 2, D-85748 Garching (Germany); Almazouzi, A. [Structural Materials Expert Group, Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Caturla, M.J. [Departamento de Fisica Aplicada, Universidad de Alicante, 03690 San Vicente de Raspeig (Spain); Fu, C.C. [Service de Metallurgie Physique, CEA/Saclay, F-91191 Gif sur Yvette Cedex (France); Kaellne, J. [Department of Engineering Sciences, Uppsala University, Box 534, S-751 21 Uppsala (Sweden); Malerba, L. [Structural Materials Expert Group, Nuclear Materials Science Institute, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Nordlund, K. [Association EURATOM-Tekes, Accelerator Laboratory, P.O. Box 43, 00014 University of Helsinki (Finland); Perlado, M. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, c/Jose Gutierrez Abascal 2, 28006 Madrid (Spain); Rieth, M. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung I, P.O. Box 3640, D-76021 Karlsruhe (Germany); Samaras, M. [Paul Scherrer Institute, Nuclear Energy and Safety Department, CH-5232 Villigen PSI (Switzerland); Schaeublin, R. [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association Euratom-Confederation Suisse, CH-5232 Villigen PSI (Switzerland); Singh, B.N. [Department of Materials Research, Risoe National Laboratory, DK-4000 Roskilde (Denmark); Willaime, F. [Service de Metallurgie Physique, CEA/Saclay, F-91191 Gif sur Yvette Cedex (France)

    2007-10-15

    We review the current status of the European fusion materials modelling programme. We describe recent findings and outline potential areas for future development. Large-scale density functional theory (DFT) calculations reveal the structure of the point defects in {alpha}-Fe, and highlight the crucial part played by magnetism. The calculations give accurate migration energies of point defects and the strength of their interaction with He atoms. Kinetic models based on DFT results reproduce the stages of radiation damage recovery in iron, and stages of He-desorption from pre-implanted iron. Experiments aimed at validating the models will be carried out in the future using a multi-beam ion irradiation facility chosen for its versatility and rapid feedback.

  12. Testing of irradiated and annealed 15H2MFA materials

    International Nuclear Information System (INIS)

    Gillemot, F.; Uri, G.

    1994-01-01

    A set of surveillance samples made from 15H2MFA material has been studied in the laboratory of AEKI. Miniature notched tensile specimens were cut from some remnants of irradiated and broke surveillance charpy remnants. The Absorbed Specific Fracture Energy (ASFE) was measured on the specimens. A cutting machine and testing technique were elaborated for the measurements. The second part of the Charpy remnants was annealed at 460 deg. C and 490 deg. C for 6-8 hours. The specimens were tested similarity and the results were compared. (author). 5 refs, 9 figs

  13. Stochastic simulation of destruction processes in self-irradiated materials

    Directory of Open Access Journals (Sweden)

    T. Patsahan

    2017-09-01

    Full Text Available Self-irradiation damages resulting from fission processes are common phenomena observed in nuclear fuel containing (NFC materials. Numerous α-decays lead to local structure transformations in NFC materials. The damages appearing due to the impacts of heavy nuclear recoils in the subsurface layer can cause detachments of material particles. Such a behaviour is similar to sputtering processes observed during a bombardment of the material surface by a flux of energetic particles. However, in the NFC material, the impacts are initiated from the bulk. In this work we propose a two-dimensional mesoscopic model to perform a stochastic simulation of the destruction processes occurring in a subsurface region of NFC material. We describe the erosion of the material surface, the evolution of its roughness and predict the detachment of the material particles. Size distributions of the emitted particles are obtained in this study. The simulation results of the model are in a qualitative agreement with the size histogram of particles produced from the material containing lava-like fuel formed during the Chernobyl nuclear power plant disaster.

  14. Dynamic nuclear polarization of irradiated target materials

    International Nuclear Information System (INIS)

    Seely, M.L.

    1982-01-01

    Polarized nucleon targets used in high energy physics experiments usually employ the method of dynamic nuclear polarization (DNP) to polarize the protons or deuterons in an alcohol. DNP requires the presence of paramagnetic centers, which are customarily provided by a chemical dopant. These chemically doped targets have a relatively low polarizable nucleon content and suffer from loss of polarization when subjected to high doses of ionizing radiation. If the paramagnetic centers formed when the target is irradiated can be used in the DNP process, it becomes possible to produce targets using materials which have a relatively high polarizable nucleon content, but which are not easily doped by chemical means. Furthermore, the polarization of such targets may be much more radiation resistant. Dynamic nuclear polarization in ammonia, deuterated ammonia, ammonium hydroxide, methylamine, borane ammonia, butonal, ethane and lithium borohydride has been studied. These studies were conducted at the Stanford Linear Accelerator Center using the Yale-SLAC polarized target system. Results indicate that the use of ammonia and deuterated ammonia as polarized target materials would make significant increases in polarized target performance possible

  15. Fusion Materials Irradiation Test Facility: experimental capabilities and test matrix

    International Nuclear Information System (INIS)

    Opperman, E.K.

    1982-01-01

    This report describes the experimental capabilities of the Fusion Materials Irradiation Test Facility (FMIT) and reference material specimen test matrices. The description of the experimental capabilities and the test matrices has been updated to match the current single test cell facility ad assessed experimenter needs. Sufficient detail has been provided so that the user can plan irradiation experiments and conceptual hardware. The types of experiments, irradiation environment and support services that will be available in FMIT are discussed

  16. Radiation research of materials using irradiation capsules

    International Nuclear Information System (INIS)

    Chamrad, B.

    1976-01-01

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

  17. Standard Guide for Packaging Materials for Foods to Be Irradiated

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This guide provides a format to assist producers and users of food packaging materials in selecting materials that have the desirable characteristics for their intended use and comply with applicable standards or government authorizations. It outlines parameters that should be considered when selecting food-contact packaging materials intended for use during irradiation of prepackaged foods and it examines the criteria for fitness for their use. 1.2 This guide identifies known regulations and regulatory frameworks worldwide pertaining to packaging materials for holding foods during irradiation; but it does not address all regulatory issues associated with the selection and use of packaging materials for foods to be irradiated. It is the responsibility of the user of this guide to determine the pertinent regulatory issues in each country where foods are to be irradiated and where irradiated foods are distributed. 1.3 This guide does not address all of the food safety issues associated with the synergisti...

  18. Neutron irradiation experiments for fusion reactor materials through JUPITER program

    International Nuclear Information System (INIS)

    Abe, K.; Namba, C.; Wiffen, F.W.; Jones, R.H.

    1998-01-01

    A Japan-USA program of irradiation experiments for fusion research, ''JUPITER'', has been established as a 6 year program from 1995 to 2000. The goal is to study ''the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment''. This is phase-three of the collaborative program, which follows RTNS-II program (phase-1: 1982-1986) and FFTF/MOTA program (phase-2: 1987-1994). This program is to provide a scientific basis for application of materials performance data, generated by fission reactor experiments, to anticipated fusion environments. Following the systematic study on cumulative irradiation effects, done through FFTF/MOTA program. JUPITER is emphasizing the importance of dynamic irradiation effects on materials performance in fusion systems. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects. (orig.)

  19. Fusion Materials Irradiation Test Facility: a facility for fusion-materials qualification

    International Nuclear Information System (INIS)

    Trego, A.L.; Hagan, J.W.; Opperman, E.K.; Burke, R.J.

    1983-01-01

    The Fusion Materials Irradiation Test Facility will provide a unique testing environment for irradiation of structural and special purpose materials in support of fusion power systems. The neutron source will be produced by a deuteron-lithium stripping reaction to generate high energy neutrons to ensure damage similar to that of a deuterium-tritium neutron spectrum. The facility design is now ready for the start of construction and much of the supporting lithium system research has been completed. Major testing of key low energy end components of the accelerator is about to commence. The facility, its testing role, and the status and major aspects of its design and supporting system development are described

  20. Capsule Development and Utilization for Material Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Kang, Y. H.; Cho, M. S. (and others)

    2007-06-15

    The essential technology for an irradiation test of materials and nuclear fuel has been successively developed and utilized to meet the user's requirements in Phase I(July 21, 1997 to March 31, 2000). It enables irradiation tests to be performed for a non-fissile material under a temperature control(300{+-}10 .deg. C) in a He gas environment, and most of the irradiation tests for the internal and external users are able to be conducted effectively. The basic technology was established to irradiate a nuclear fuel, and a creep capsule was also developed to measure the creep property of a material during an irradiation test in HANARO in Phase II(April 1, 2000 to March 31, 2003). The development of a specific purpose capsule, essential technology for a re-irradiation of a nuclear fuel, advanced technology for an irradiation of materials and a nuclear fuel were performed in Phase III(April 1, 2003 to February 28, 2007). Therefore, the technology for an irradiation test was established to support the irradiation of materials and a nuclear fuel which is required for the National Nuclear R and D Programs. In addition, an improvement of the existing capsule design and fabrication technology, and the development of an instrumented capsule for a nuclear fuel and a specific purpose will be able to satisfy the user's requirements. In order to support the irradiation test of materials and a nuclear fuel for developing the next generation nuclear system, it is also necessary to continuously improve the design and fabrication technology of the existing capsule and the irradiation technology.

  1. Capsule Development and Utilization for Material Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Kang, Y H; Cho, M S [and others

    2007-06-15

    The essential technology for an irradiation test of materials and nuclear fuel has been successively developed and utilized to meet the user's requirements in Phase I(July 21, 1997 to March 31, 2000). It enables irradiation tests to be performed for a non-fissile material under a temperature control(300{+-}10 .deg. C) in a He gas environment, and most of the irradiation tests for the internal and external users are able to be conducted effectively. The basic technology was established to irradiate a nuclear fuel, and a creep capsule was also developed to measure the creep property of a material during an irradiation test in HANARO in Phase II(April 1, 2000 to March 31, 2003). The development of a specific purpose capsule, essential technology for a re-irradiation of a nuclear fuel, advanced technology for an irradiation of materials and a nuclear fuel were performed in Phase III(April 1, 2003 to February 28, 2007). Therefore, the technology for an irradiation test was established to support the irradiation of materials and a nuclear fuel which is required for the National Nuclear R and D Programs. In addition, an improvement of the existing capsule design and fabrication technology, and the development of an instrumented capsule for a nuclear fuel and a specific purpose will be able to satisfy the user's requirements. In order to support the irradiation test of materials and a nuclear fuel for developing the next generation nuclear system, it is also necessary to continuously improve the design and fabrication technology of the existing capsule and the irradiation technology.

  2. Laser irradiation of carbon–tungsten materials

    International Nuclear Information System (INIS)

    Marcu, A; Lungu, C P; Ursescu, D; Porosnicu, C; Grigoriu, C; Avotina, L; Kizane, G; Marin, A; Osiceanu, P; Grigorescu, C E A; Demitri, N

    2014-01-01

    Carbon–tungsten layers deposited on graphite by thermionic vacuum arc (TVA) were directly irradiated with a femtosecond terawatt laser. The morphological and structural changes produced in the irradiated area by different numbers of pulses were systematically explored, both along the spots and in their depths. Although micro-Raman and Synchrotron-x-ray diffraction investigations have shown no carbide formation, they have shown the unexpected presence of embedded nano-diamonds in the areas irradiated with high fluencies. Scanning electron microscopy images show a cumulative effect of the laser pulses on the morphology through the ablation process. The micro-Raman spatial mapping signalled an increased percentage of sp 3 carbon bonding in the areas irradiated with laser fluencies around the ablation threshold. In-depth x-ray photoelectron spectroscopy investigations suggested a weak cumulative effect on the percentage increase of the sp 2 -sp 3 transitions with the number of laser pulses just for nanometric layer thicknesses. (paper)

  3. On the material durability under irradiation conditions

    International Nuclear Information System (INIS)

    Kiselevskij, V.N.; Kosov, B.D.

    1977-01-01

    The initial principle adopted for the construction of a phenomenological model of the failure of irradiated steel, as proposed in the paper of V.A. Tsykanov and coworkers, is analized and some critical remarks made

  4. Van de Graaff Irradiation of Materials

    Energy Technology Data Exchange (ETDEWEB)

    Quigley, Kevin [Argonne National Lab. (ANL), Argonne, IL (United States); Chemerisov, Sergey [Argonne National Lab. (ANL), Argonne, IL (United States); Tkac, Peter [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, George F. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-10-01

    Through irradiations using our 3 MeV Van de Graaf accelerator, Argonne is testing the radiation stability of components of equipment that are being used to dispense molybdenum solutions for use as feeds to 99mTc generators and in the 99mTc generators themselves. Components have been irradiated by both a direct electron beam and photons generated from a tungsten convertor.

  5. On the Thermal Conductivity Change of Matrix Graphite Materials after Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Kim, Eung-Seon; Sah, Injin; Park, Daegyu; Kim, Youngjun; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this work, the variations of the thermal conductivity of the A3-3 matrix graphite after neutron irradiation is discussed as well as of the IG-110 graphite for comparison. Neutron irradiation of the graphite specimens was carried out as a part of the first irradiation test of KAERI's coated particle fuel specimens by use of Hanaro research reactor. This work can be summarized as follows: 1) In the evaluation of the specific heat of the graphite materials, various literature data were used and the variations of the specific heat data of all the graphite specimens are observed well agreed, irrespectively of the difference in specimens (graphite and matrix graphite and irradiated and un-irradiated). 2) This implies that it should be reasonable that for both structural graphite and fuel matrix graphite, and even for the neuron-irradiated graphite, any of these specific heat data set be used in the calculation of the thermal conductivity. 3) For the irradiated A3-3 matrix graphite specimens, the thermal conductivity decreased on both directions. On the radial direction, the tendency of variation upon temperature is similar to that of unirradiated specimen, i.e., decreasing as the temperature increases. 4) In the German irradiation experiments with A3-27 matrix graphite specimens, the thermal conductivity of the un-irradiated specimen shows a decrease and that of irradiated specimen is nearly constant as the temperature increases. 5) The thermal conductivity of the irradiated IG-110 was considerably decreased compared with that of un-irradiated specimens The difference of the thermal conductivity of un-irradiated and irradiated IG-110 graphite specimens is much larger than that of un-irradiated and irradiated A3-3 matrix graphite specimens.

  6. Opening of new field in material science and technology by materials irradiation research

    Energy Technology Data Exchange (ETDEWEB)

    Kurishita, Hiroaki [Tohoku Univ., Sendai (Japan). Inst. for Materials Research

    1998-03-01

    It is believed that high energy particle irradiation causes severe degradation of materials, and great efforts have been made to reveal the underlying mechanism of such degradation. However, recent progress of the developments of irradiation rigs performed in the Japan Materials Testing Reactor (JMTR) and materials fabrication techniques has enabled to change our understanding of radiation effects on materials from the above pessimistic one to the very challenging one, i.e., irradiation has the beneficial effect of producing new phenomena and/or innovative materials that will not be available without irradiation. An example to be noted is that irradiation with neutrons in JMTR greatly improved the ductility of less ductile metals. This ductility improvement due to irradiation is directly opposite to irradiation embrittlement and is called radiation induced ductilization (RIDU). In this presentation the significance of RIDU and its mechanism will be stated. (author)

  7. Materials for cold neutron sources: Cryogenic and irradiation effects

    International Nuclear Information System (INIS)

    Alexander, D.J.

    1990-01-01

    Materials for the construction of cold neutron sources must satisfy a range of demands. The cryogenic temperature and irradiation create a severe environment. Candidate materials are identified and existing cold sources are briefly surveyed to determine which materials may be used. Aluminum- and magnesium-based alloys are the preferred materials. Existing data for the effects of cryogenic temperature and near-ambient irradiation on the mechanical properties of these alloys are briefly reviewed, and the very limited information on the effects of cryogenic irradiation are outlined. Generating mechanical property data under cold source operating conditions is a daunting prospect. It is clear that the cold source material will be degraded by neutron irradiation, and so the cold source must be designed as a brittle vessel. The continued effective operation of many different cold sources at a number of reactors makes it clear that this can be accomplished. 46 refs., 8 figs., 2 tab

  8. Disk-bend ductility tests for irradiated materials

    International Nuclear Information System (INIS)

    Klueh, R.L.; Braski, D.N.

    1984-01-01

    We modified the HEDL disk-bend test machine and are using it to qualitatively screen alloys that are susceptible to embrittlement caused by irradiation. Tests designed to understand the disk-bend test in relation to a uniaxial test are discussed. Selected results of tests of neutron-irradiated material are also presented

  9. Effect of 60Co γ-irradiation on saccharification of uncooked sweet potato material

    International Nuclear Information System (INIS)

    Hu Tingchun; Xiong Xingyao; Yi Jinqiong; Wang Keqin; Su Xiaojun; Zou Jianfeng

    2010-01-01

    Using the starch and powder of sweet potato of Xiangshu 86 and Xiangshu 541 as materials, the effect of 60 Co γ-irradiation on the structure of starch particle and the efficiency of saccharification were studied. The result showed that some reticulate flaws appeared in the surface of irradiated starch particles, and the reticulate flaws were increased with the increase of irradiation dose. The content of reducing sugar and total soluble sugar in both starch and the powder were obviously increased along with the increase of irradiation dose ranged from 50 to 1200 kGy. The saccharification efficiency of Xiangshu 86 and Xiangshu 541 was obviously difference at the dose lower than 500 kGy, and then the efficiency showed the similar trends at higher dose irradiation, the saccharification rate reached the highest value after the treatment of 1200 kGy irradiation. (authors)

  10. Production behavior of irradiation defects in solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Hirotake; Moritani, Kimikazu [Kyoto Univ. (Japan)

    1998-03-01

    The irradiation effects in solid breeder materials are important for the performance assessment of fusion reactor blanket systems. For a clearer understanding of such effects, we have studied the production behavior of irradiation defects in some lithium ceramics by an in-situ luminescence measurement technique under ion beam irradiation. The luminescence spectra were measured at different temperatures, and the temperature-transient behaviors of luminescence intensity were also measured. The production mechanisms of irradiation defects were discussed on the basis of the observations. (author)

  11. Development status of irradiation devices and instrumentation for material and nuclear fuel irradiation tests in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Sohn, Jae Min; Choo, Kee Nam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-04-15

    The High flux Advanced Neutron Application ReactOr (HANARO), an open-tank-in-pool type reactor, is one of the multi-purpose research reactors in the world. Since the commencement of HANARO's operations in 1995, a significant number of experimental facilities have been developed and installed at HANARO, and continued efforts to develop more facilities are in progress. Owing to the stable operation of the reactor and its frequent utilization, more experimental facilities are being continuously added to satisfy various fields of study and diverse applications. The irradiation testing equipment for nuclear fuels and materials at HANARO can be classified into capsules and the Fuel Test Loop (FTL). Capsules for irradiation tests of nuclear fuels in HANARO have been developed for use under the dry conditions of the coolant and materials at HANARO and are now successfully utilized to perform irradiation tests. The FTL can be used to conduct irradiation testing of a nuclear fuel under the operating conditions of commercial nuclear power plants. During irradiation tests conducted using these capsules in HANARO, instruments such as the thermocouple, Linear Variable Differential Transformer (LVDT), small heater, Fluence Monitor (F/M) and Self-Powered Neutron Detector (SPND) are used to measure various characteristics of the nuclear fuel and irradiated material. This paper describes not only the status of HANARO and the status and perspective of irradiation devices and instrumentation for carrying out nuclear fuel and material tests in HANARO but also some results from instrumentation during irradiation tests

  12. Experimental study associated to irradiation of FBR structural material, (4)

    International Nuclear Information System (INIS)

    1976-01-01

    The study presents one of the bases to evaluate the results of the post-irradiation tests to conduct the thermal control tests related to the second JMTR irradiation (70M-61P) of the demestic austenitic stainless steels for the structural material of the FBR performed by Power Reactor and Nuclear Fuel Development Corporation. The thermal control specimens were given the temperature history which simulated that of the irradiation temperature in vacuum by the electrical furnance, and then the tensile, fatigue and Charpy impact tests were performed. The changes of the material properties caused by the thermal history were investigated. (auth.)

  13. 10 CFR 36.69 - Irradiation of explosive or flammable materials.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Irradiation of explosive or flammable materials. 36.69... IRRADIATORS Operation of Irradiators § 36.69 Irradiation of explosive or flammable materials. (a) Irradiation... cause radiation overexposures of personnel. (b) Irradiation of more than small quantities of flammable...

  14. Scaling similarities of multiple fracturing of solid materials

    Directory of Open Access Journals (Sweden)

    P. G. Kapiris

    2004-01-01

    Full Text Available It has recently reported that electromagnetic flashes of low-energy -rays emitted during multi-fracturing on a neutron star, and electromagnetic pulses emitted in the laboratory by a disordered material subjected to an increasing external load, share distinctive statistical properties with earthquakes, such as power-law energy distributions (Cheng et al., 1996; Kossobokov et al., 2000; Rabinovitch et al., 2001; Sornette and Helmstetter, 2002. The neutron starquakes may release strain energies up to erg, while, the fractures in laboratory samples release strain energies approximately a fraction of an erg. An earthquake fault region can build up strain energy up to approximately erg for the strongest earthquakes. Clear sequences of kilohertz-megahertz electromagnetic avalanches have been detected from a few days up to a few hours prior to recent destructive earthquakes in Greece. A question that arises effortlessly is if the pre-seismic electromagnetic fluctuations also share the same statistical properties. Our study justifies a positive answer. Our analysis also reveals 'symptoms' of a transition to the main rupture common with earthquake sequences and acoustic emission pulses observed during laboratory experiments (Maes et al., 1998.

  15. Material property changes of stainless steels under PWR irradiation

    International Nuclear Information System (INIS)

    Fukuya, Koji; Nishioka, Hiromasa; Fujii, Katsuhiko; Kamaya, Masayuki; Miura, Terumitsu; Torimaru, Tadahiko

    2009-01-01

    Structural integrity of core structural materials is one of the key issues for long and safe operation of pressurized water reactors. The stainless steel components are exposed to neutron irradiation and high-temperature water, which cause significant property changes and irradiation assisted stress corrosion cracking (IASCC) in some cases. Understanding of irradiation induced material property changes is essential to predict integrity of core components. In the present study, microstructure and microchemistry, mechanical properties, and IASCC behavior were examined in 316 stainless steels irradiated to 1 - 73 dpa in a PWR. Dose-dependent changes of dislocation loops and cavities, grain boundary segregation, tensile properties and fracture mode, deformation behavior, and their interrelation were discussed. Tensile properties and deformation behavior were well coincident with microstructural changes. IASCC susceptibility under slow strain rate tensile tests, IASCC initiation under constant load tests in simulated PWR primary water, and their relationship to material changes were discussed. (author)

  16. Capsule Development and Utilization for Material Irradiation Tests

    International Nuclear Information System (INIS)

    Kang, Young Hwan; Kim, B. G.; Joo, K. N.

    2003-05-01

    The objective of this project was to establish basic capsule irradiation technology using the multi-purpose research reactor [HANARO] to eventually support national R and D projects of advanced fuel and materials related to domestic nuclear power plants and next generation reactors. There are several national nuclear projects in KAERI, which require several irradiation tests to investigate in-pile behavior of nuclear reactor fuel and materials for the R and D of several types of fuels such as advanced PWR and DUPIC fuels and for the R and D of structural materials such as RPV(reactor pressure vessel) steel, Inconel, zirconium alloy, and stainless steel. At the moment, internal and external researchers in institutes, industries and universities are interested in investigating the irradiation characteristics of materials using the irradiation facilities of HANARO. For these kinds of material irradiation tests, it is important to develop various capsules using our own techniques. The development of capsules requires several leading-edge technologies and our own experiences related to design and fabrication. In the second phase from April 1,2000 to March 31, 2003, the utilization technologies were developed using various sensors for the measurements of temperature, pressure and displacement, and instrumented capsule technologies for the required fuel irradiation tests were developed. In addition, the improvement of the existing capsule technologies and the development of an in-situ measurable creep capsule for specific purposes were done to meet the various requirements of users

  17. Microstructural examination of irradiated zircaloy-2 pressure tube material

    International Nuclear Information System (INIS)

    Srivastava, D.; Tewari, R.; Dey, G.K.; Sah, D.N.; Banerjee, S.

    2005-01-01

    Irradiation induced microstructural changes in Zr alloys strongly influence the creep, growth and mechanical properties of pressure tube material. Since dimensional changes and mechanical property degradation can limit the life of pressure tube, it is essential to study and develop an understanding of the microstructure produced by neutron irradiation, by examining samples taken from the irradiated components. In the present work, an effort has been made to examine, microstructure of the Zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water Reactor (PHWR). The present work is a first step towards a comprehensive program of characterization of microstructure of reactor materials after irradiation to different fluence levels in power reactors. In this study, samples from a Zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit 1, for a period for 6.77 effective full power years (EFPYs), have been prepared and examined. The samples selected from the tube are expected to have a cumulative radiation damage of about 3 dpa. Samples prepared from the off cuts of RAPS-1 pressure tubes were also studied for examining the unirradiated microstructure of the material. The samples were examined in a 200kV JEOL 2000 FX microscope. This paper presents the distinct features observed in irradiated sample and a comprehensive comparison of the microstructures of the unirradiated and irradiated material. The effect of annealing on the annihilation of the defects generated during irradiation has been also studied. The bright field micrographs revealed that microstructure of the irradiated samples was different in many respects from the microstructure of the unirradiated samples. The presence of defect structure in the form of loops etc could be seen in the irradiated sample. These loops were mostly c-type loops lying in the basal plane. The dissolution and redistribution of the precipitates were

  18. Needs of in-situ materials testing under neutron irradiation

    International Nuclear Information System (INIS)

    Noda, K.; Hishinuma, A.; Kiuchi, K.

    1989-01-01

    Under neutron irradiation, the component atoms of materials are displaced as primary knock-on atoms, and the energy of the primary knock-on atoms is consumed by electron excitation and nuclear collision. Elementary irradiation defects accumulate to form damage structure including voids and bubbles. In situ test under neutron irradiation is necessary for investigating into the effect of irradiation on creep behavior, the electric properties of ceramics, transport phenomena and so on. The in situ test is also important to investigate into the phenomena related to the chemical reaction with environment during irradiation. Accelerator type high energy neutron sources are preferable to fission reactors. In this paper, the needs and the research items of in situ test under neutron irradiation using a D-Li stripping type high energy neutron source on metallic and ceramic materials are described. Creep behavior is one of the most important mechanical properties, and depends strongly on irradiation environment, also it is closely related to microstructure. Irradiation affects the electric conductibity of ceramics and also their creep behavior. In this way, in situ test is necessary. (K.I.)

  19. Irradiation behavior of graphite shielding materials for FBR

    International Nuclear Information System (INIS)

    Maruyama, Tadashi; Kaito, Takeji; Onose, Shoji; Shibahara, Itaru

    1994-01-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor 'JOYO' to fluences from 2.11 to 2.86x10 26 n/m 2 (E>0.1 MeV) at temperatures from 549 to 597degC. Postirradiation examination was carried out on dimensional change, elastic modulus, and the thermal conductivity. The result of measurement of dimensional change indicated that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased to two to three times of unirradiated values. A large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependency on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, but the change in specific heat was negligibly small. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens. (author)

  20. How to improve the irradiation conditions for the International Fusion Materials Irradiation Facility

    CERN Document Server

    Daum, E

    2000-01-01

    The accelerator-based intense D-Li neutron source International Fusion Materials Irradiation Facility (IFMIF) provides very suitable irradiation conditions for fusion materials development with the attractive option of accelerated irradiations. Investigations show that a neutron moderator made of tungsten and placed in the IFMIF test cell can further improve the irradiation conditions. The moderator softens the IFMIF neutron spectrum by enhancing the fraction of low energy neutrons. For displacement damage, the ratio of point defects to cascades is more DEMO relevant and for tritium production in Li-based breeding ceramic materials it leads to a preferred production via the sup 6 Li(n,t) sup 4 He channel as it occurs in a DEMO breeding blanket.

  1. New JMTR irradiation test plan on fuels and materials

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Nishiyama, Yutaka; Chimi, Yasuhiro; Sasajima, Hideo; Ogiyanagi, Jin; Nakamura, Jinichi; Suzuki, Masahide; Kawamura, Hiroshi

    2009-01-01

    In order to maintain and enhance safety of light water reactors (LWRs) in long-term and up-graded operations, proper understanding of irradiation behavior of fuels and materials is essentially important. Japanese government and the Japan Atomic Energy Agency (JAEA) have decided to refurbish the Japan Materials Testing Reactor (JMTR) and to install new tests rigs, in order to play an active role for solving irradiation related issues on plant aging and high-duty uses of the current LWRs and on development of next-generation reactors. New tests on fuel integrity under simulated abnormal transients and high-duty irradiation conditions are planned in the JMTR. Power ramp tests of newdesign fuel rods will also be performed in the first stage of the program, which is expected to start in year 2011 after refurbishment of the JMTR. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor (NSRR) and loss of coolant accident tests in hot laboratories would serve as the integrated fuel safety research on the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients. For the materials irradiation, fracture toughness of reactor vessel steels and stress corrosion cracking behavior of stainless steels are being studied in addition to basic irradiation behavior of nuclear materials such as hafnium. The irradiation studies would contribute not only to solve the current problems but also to identify possible seeds of troubles and to make proactive responses. (author)

  2. Stored energy in fusion magnet materials irradiated at low temperatures

    International Nuclear Information System (INIS)

    Chaplin, R.L.; Kerchner, H.R.; Klabunde, C.E.; Coltman, R.R.

    1989-08-01

    During the power cycle of a fusion reactor, the radiation reaching the superconducting magnet system will produce an accumulation of immobile defects in the magnet materials. During a subsequent warm-up cycle of the magnet system, the defects will become mobile and interact to produce new defect configurations as well as some mutual defect annihilations which generate heat-the release of stored energy. This report presents a brief qualitative discussion of the mechanisms for the production and release of stored energy in irradiated materials, a theoretical analysis of the thermal response of irradiated materials, theoretical analysis of the thermal response of irradiated materials during warm-up, and a discussion of the possible impact of stored energy release on fusion magnet operation 20 refs

  3. Materials aging: first predictive modeling of iron under irradiation

    International Nuclear Information System (INIS)

    Anon.

    2005-01-01

    Researchers from the CEA-Bruyeres-le-Chatel have been able to quantitatively foresee for the very first time the evolution of irradiation defects inside a structural material. Their results, obtained with iron, will contribute to better understand the aging of the materials of today's nuclear power plants and of future nuclear systems. Short paper. (J.S.)

  4. TEM investigation of plant-irradiated NPP bolt material

    International Nuclear Information System (INIS)

    Pakarinen, J.; Ehrnsten, U.; Keinaenen, H.; Karlsen, W.; Karlsen, T.

    2015-01-01

    Analytical transmission electron microscopy (ATEM) was used to examine irradiation-induced damage in material removed from two different bolts from two different nuclear power plants. One section came from a French PWR, was made of CW AISI 316, and included a section of the bolt that had accumulated a dose of approximately 15 dpa during 19 operation cycles at 350 - 390 C. degrees. Another section came from a VVER bolt that was removed from the plant due to indications found in non-destructive examinations (NDE). The VVER bolt was made of solution annealed titanium stabilized 0X18H10T (corresponding to Type AISI 321) and had accumulated a fluence of 2.9 dpa. During the removal of that bolt, it was found that the bolt washer had been inappropriately spot welded to the shielding plate during assembly. Destructive investigations showed that the bolt had two large intergranular cracks, and the TEM samples were prepared from the material adjacent to those cracks. The PWR bolt had not failed, although cracks in the bolts with a similar history had been found previously. The fluence for the cold-worked AISI 316 PWR bolt was estimated to be about 15 dpa. Both the examined bolts showed a clear radiation induced segregation of alloying elements at the grain boundaries (GB-RIS), the presence of dislocation loops, the formation of precipitates, and linear deformation microstructures. Additionally, voids were found from the PWR bolt and the VVER bolt had a high density of dislocations. (authors)

  5. In-service irradiated and aged material evaluations

    International Nuclear Information System (INIS)

    Haggag, F.M.; Nanstad, R.K.; Alexander, D.J.

    1995-01-01

    The objective of this task is to provide a direct assessment of actual material properties in irradiated components of nuclear reactors, including the effects of irradiation and aging. Four activities are currently in progress: (1) establishing a machining capability for contaminated or activated materials by completing procurement and installation of a computer-based milling machine in a hot cell; (2) machining and testing specimens from cladding materials removed from the Gundremmingen reactor to establish their fracture properties; (3) preparing an interpretive report on the effects of neutron irradiation on cladding; and (4) continuing the evaluation of long-term aging of austenitic structural stainless steel weld metal by metallurgically examining and testing specimens aged at 288 and 343 degrees C and reporting the results, as well as by continuing the aging of the stainless steel cladding toward a total time of 50,000 h

  6. Accelerated irradiation test of gundremmingen reactor vessel trepan material

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1992-08-01

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

  7. Accelerated irradiation test of Gundremmingen reactor vessel trepan material

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-08-01

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

  8. The international fusion materials irradiation facility

    International Nuclear Information System (INIS)

    Shannon, T.E.; Cozzani, F.; Crandall, D.H.; Wiffen, F.W.; Katsuta, H.; Kondo, T.; Teplyakov, V.; Zavialsky, L.

    1994-01-01

    It is widely agreed that the development of materials for fusion systems requires a high flux, 14 MeV neutron source. The European Union, Japan, Russia and the US have initiated the conceptual design of such a facility. This activity, under the International Energy Agency (IEA) Fusion Materials Agreement, will develop the design for an accelerator-based D-Li system. The first organizational meeting was held in June 1994. This paper describes the system to be studied and the approach to be followed to complete the conceptual design by early 1997

  9. Modelling irradiation effects in fusion materials

    DEFF Research Database (Denmark)

    Victoria, M.; Dudarev, S.; Boutard, J.L.

    2007-01-01

    We review the current status of the European fusion materials modelling programme. We describe recent findings and outline potential areas for future development. Large-scale density functional theory (DFT) calculations reveal the structure of the point defects in α-Fe, and highlight the crucial...

  10. 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.

  11. The feasibility of welding irradiated materials

    Science.gov (United States)

    Lin, H. T.; Chin, B. A.

    1991-03-01

    Helium was implanted into solution-annealed (SA) 316 stainless steel, 20% cold-worked (CW) 316 stainless steel and titanium-modified Primary Candidate Alloy (PCA) through tritium decay to levels ranging from 0.18 to 256 appm. Full penetration welds were then made on helium-doped materials using gas tungsten arc welding (GTAW) under fully constrained conditions. Intergranular heat-affected zone (HAZ) cracking was observed in all of the materials containing greater than 1 appm He. Electron microscopy showed that the HAZ cracking originated from the growth and coalescence of grain boundary (GB) helium bubbles. Bubble growth kinetics in the HAZ is explained by stress-enhanced diffusive cavity growth. Results suggest that the propensity for HAZ cracking can be reduced by the pre-existing cold-worked structure and by finely-distributed MC precipitates that refine the distribution of helium bubbles and minimize the flow of vacancies in grain boundaries.

  12. The feasibility of welding of irradiated materials

    International Nuclear Information System (INIS)

    Lin, H.T.; Chin, B.A.; Auburn Univ., AL

    1989-01-01

    Helium was implanted into solution-annealed (SA) 316 stainless steel, 20% cold-worked (CW) 316 stainless steel and titanium-modified Primary Candidate Alloy (PCA) through tritium decay to levels ranging from 0.18 to 256 appm. Full penetration welds were then made on helium-doped materials using gas tungsten arc welding (GTAW) under fully constrained conditions. Intergranular heat-affected zone (HAZ) cracking was observed in all of the materials containing greater than 1 appm He. Electron microscopy showed that the HAZ cracking originated from the growth and coalescence of grain boundary (GB) helium bubbles. Bubble growth kinetics in the HAZ is explained by stress-enhanced diffusive cavity growth. Results suggest that the propensity for HAZ cracking can be reduced by the preexisting cold-worked structure and by finely-distributed MC precipitates that refine the distribution of helium bubbles and minimize the flow of vacancies in grain boundaries. 16 refs., 3 figs

  13. Irradiation devices for fusion reactor materials results obtained from irradiated lithium aluminate at the OSIRIS reactor

    International Nuclear Information System (INIS)

    Lefevre, F.; Thevenot, G.; Rasneur, B.; Botter, F.

    1986-06-01

    Studies about controlled fusion reactor of the Tokamak type require the examination of the radiation effects on the behaviour of various potential materials. Thus, in the first part of this paper, are presented the devices adapted to these materials studies and used in the OSIRIS reactor. In a second part, is described an experiment of irradiation ceramics used as candidates for breeding material and are given the first results

  14. 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

  15. Repair-welding technology of irradiated materials - WIM project

    International Nuclear Information System (INIS)

    Nakata, K.; Oishi, M.

    1998-01-01

    A new project on the development of repair-welding technology for core internals and reactor (pressure) vessel, consigned by the Ministry of International Trade and Industry (MITI), has been started from October 1997. The objective of the project is classified into three points as follows: (1) to develop repair-welding techniques for neutron irradiated materials, (2) to prove the availability of the techniques for core internals and reactor (pressure) vessel, and (3) to recommend the updated repair-welding for the Technical Rules and Standards. Total planning, neutron irradiation, preparation of welding equipment are now in progress. The materials are austenitic stainless steels and a low alloy steel. Neutron irradiation is performed using test reactors. In order to suppress the helium aggregation along grain boundaries, low heat input welding techniques, such as laser, low heat input TIG and friction weldings, will be applied. (author)

  16. Irradiation effects of hydrogen and helium plasma on different grade tungsten materials

    Directory of Open Access Journals (Sweden)

    X. Liu

    2017-08-01

    Full Text Available Fine-grain tungsten alloys could be one of the solutions for the plasma facing materials of future DEMO reactors. In order to evaluate the service performances of the newly developed W alloys under edge plasma irradiation and the synergetic effect of fusion plasma together with high heat flux, both low energy He ions and high energy H, H/He mixed neutral beam irradiation on W-ZrC, W-K, W-Y2O3, W-La2O3 and CVD-W coating were performed respectively at a liner plasma facility (Dalian Nationality University, China and the neutral beam facility GLADIS (IPP, Germany. Surface damages were characterized, and the crack formation and extension behaviors under ELM-like transient loading after H and H/He mixed beam irradiation were also investigated in the 60kW EMS-60 facility (Electron beam Materials testing Scenario at SWIP (Southwestern Institute of Physics, China. The experimental results indicated that surface damages induced by low or high energy H/He ion/neutral beam didn't closely correlate with the type of tungsten materials. However, H/He (6at% He concentration neutral beam induced more significant surface damages of the tested W materials than only H neutral beam irradiation under the similar irradiation conditions. Similarly, the mixed H/He pre-exposure remarkably reduced the critical power of crack initiation compared with the un-irradiated samples under 100 repetitive loads of 1ms pulse, while no significant degeneration for the case of only H beam irradiation was observed.

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

    International Nuclear Information System (INIS)

    Perlado, J. M.; Victoria, M.

    2007-01-01

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

  18. Electron irradiation experiments in support of fusion materials development

    International Nuclear Information System (INIS)

    Gelles, D.S.; Ohnuki, S.; Takahashi, H.; Matsui, H.; Kohno, Y.

    1991-11-01

    Microstructural evolution in response to 1 MeV irradiation has been investigated for three simple ferritic alloys, pure beryllium, pure vanadium, and two simple vanadium alloys over a range of temperatures and doses. Microstructural evolution in Fe-3, -9, and -18Cr ferritic alloys is found to consist of crenulated, faulted a loops and circular, unfaulted a/2 loops at low temperatures, but with only unfaulted loops at high temperatures. The complex dislocation evolution is attributed to sigma phase precipifaults arising from chromium segregation to point defect sinks. Beryllium is found to be resistant to electron damage; the only effect observed was enhanced dislocation mobility. Pure vanadium, V-5Fe, and V-1Ni microstructural response was complicated by precipitation on heating to 400 degrees C and above, but dislocation evolution was investigated in the range of room temperature to 300 degrees C and at 600 degrees C. The three materials behaved similarly, except that pure vanadium showed more rapid dislocation evolution. This difference does not explain the enhanced swelling observed in vanadium alloys

  19. Sampling by electro-erosion on irradiated materials

    International Nuclear Information System (INIS)

    Riviere, M.; Pizzanelli, J.P.

    1986-05-01

    Sampling on irradiated materials, in particular for mechanical property study of steels in the FAST NEUTRON program needed the set in a hot cell of a machining device by electroerosion. This device allows sampling of tenacity, traction, resilience test pieces [fr

  20. Fusion materials irradiation test facility: description and status

    International Nuclear Information System (INIS)

    Trego, A.L.; Parker, E.F.; Hagan, J.W.

    1982-01-01

    The Fusion Materials Irradiation Test (FMIT) Facility will generate a high-flux, high-energy neutron source that will provide a fusion-like radiation environment for fusion reactor materials development. The neutrons will be produced in a nuclear stripping reaction by impinging a 35 MeV beam of deuterons from an Alvarez-type linear accelerator on a flowing lithium target. The target will be located in a test cell which will provide an irradiation volume of over 750l within which 10 cm 3 will have an average neutron flux of greater than 1.4 x 10 15 n/cm 2 -s and 500 cm 3 an average flux of greater than 2.2 by 10 14 n/cm 2- s with an expected availability factor greater than 65%. The projected fluence within the 10 cm 3 high flux region of FMIT will effect damage upon the materials test specimens to 30 dpa (displacements per atom) for each 90 day irradiation period. This irradiation flux volume will be at least 500 times larger than that of any other facility with comparable neutron energy and will fully meet the fusion materials damage research objective of 100 dpa within three years for the first round of tests

  1. Irradiation-Accelerated Corrosion of Reactor Core Materials. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, Zhujie [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Bartels, David [Univ. of Notre Dame, IN (United States)

    2015-04-02

    This project aims to understand how radiation accelerates corrosion of reactor core materials. The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, as well as the success of most all GenIV concepts. Of these four drivers, the combination of radiation and corrosion places the most severe demands on materials, for which an understanding of the fundamental science is simply absent. Only a few experiments have been conducted to understand how corrosion occurs under irradiation, yet the limited data indicates that the effect is large; irradiation causes order of magnitude increases in corrosion rates. Without a firm understanding of the mechanisms by which radiation and corrosion interact in film formation, growth, breakdown and repair, the extension of the current LWR fleet beyond 60 years and the success of advanced nuclear energy systems are questionable. The proposed work will address the process of irradiation-accelerated corrosion that is important to all current and advanced reactor designs, but remains very poorly understood. An improved understanding of the role of irradiation in the corrosion process will provide the community with the tools to develop predictive models for in-reactor corrosion, and to address specific, important forms of corrosion such as irradiation assisted stress corrosion cracking.

  2. Growth kinetics of dislocation loops in irradiated ceramic materials

    International Nuclear Information System (INIS)

    Ryazanov, A.I.; Kinoshita, C.

    2002-01-01

    Ceramic materials are expected to be applied in the future fusion reactor as radio frequency (RF) windows, toroidal insulating breaks and diagnostic probes. The radiation resistance of ceramic materials, degradation of the electrical properties and radiation induced conductivity of these materials under neutron irradiation are determined by the kinetics of the accumulation of point defects in the matrix and point defect cluster formation (dislocation loops, voids, etc.). Under irradiation, due to the ionization process, excitation of electronic subsystem and covalent type of interaction between atoms the point defects in ceramic materials are characterized by the charge state (e.g. an F + center, an oxygen vacancy with a single trapped electron) and the effective charge. For the investigation of radiation resistance of ceramic materials for future fusion applications it is very important to understand the physical mechanisms of formation and growth of dislocation loops and voids under irradiation taking into account in this system the effective charge of point defects. In the present paper the physical mechanisms of dislocation loop growth in ceramic material are investigated. For this aim a theoretical model is suggested for the description of the kinetics of point defect accumulation in the matrix taking into account the charge state of the point defects and the effect of an electric field on diffusion migration process of charged point defects. A self-consistent system of kinetic equations describing the generation of electrical fields near dislocation loops and diffusion migration of charged point defects in elastic and electrical fields is formulated. The solution of the kinetic equations allows to find the growth rate of dislocation loops in ceramic materials under irradiation taking into account the charge state of the point defects and the effect of electric and elastic stress fields near dislocation loop on the diffusion processes

  3. Swedish studies on irradiation effect in structural materials

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M; Myers, H P

    1962-12-15

    A brief description of work in hand at AB Atomenergi concerning the effects of neutron irradiation on structural materials is given. Some recent data is listed for the following pressure vessel steels 2103/R3 as used in the Aagesta reactor, SIS 142103, NO345, Fortiweld and weld metal OK 54 P. Zircaloy-2 has been studied regarding the combined effects of neutron irradiation and hydrogen content on tensile properties. The difficulties associated with determination of neutron dose and the correlation of damage with dose and neutron energy spectrum are discussed.

  4. Effects of gamma irradiation on raw materials and perfumes

    International Nuclear Information System (INIS)

    Guillot, M.; Pelpel, A.

    1983-01-01

    In order to enlight the strange problem of apparent perfume stability observed in manufactured talc powders sterilized by gamma rays, investigations were made on samples of odorant substances (raw materials, essential oils, or elaborated mixtures). As a rule, no immediate adulteration of olfactive caracteristics resulted at once from gamma irradiation. In several cases, a stabilizing effect appeared immediately and remained effective after long storage in various conditions (of temperature, or light, or oxygen exposure). This unexpected effect seems to be in accordance with previous experiments on gamma or electron irradiations of mixtures of organic molecules, reported in litterature: a mutual inhibition was observed to take place [fr

  5. Swedish studies on irradiation effect in structural materials

    International Nuclear Information System (INIS)

    Grounes, M.; Myers, H.P.

    1962-12-01

    A brief description of work in hand at AB Atomenergi concerning the effects of neutron irradiation on structural materials is given. Some recent data is listed for the following pressure vessel steels 2103/R3 as used in the Aagesta reactor, SIS 142103, NO345, Fortiweld and weld metal OK 54 P. Zircaloy-2 has been studied regarding the combined effects of neutron irradiation and hydrogen content on tensile properties. The difficulties associated with determination of neutron dose and the correlation of damage with dose and neutron energy spectrum are discussed

  6. Direct or indirect UV-Irradiation in insect killers and similar equipment

    International Nuclear Information System (INIS)

    Heinz, G.

    1978-01-01

    UV sterilisation equipment is used in the refrigerating and storage rooms of food and meat processing factories. The UV radiation used has a wavelength of 254 nm. Due to possible side effects on the food, the packing is directly irradiated but the food itself only indirectly. This method makes high irradiation energies necessary for the inhibition of bacterial growth on the foodstuffs. (AJ) 891 AJ [de

  7. Botryllus schlosseri (Tunicata) whole colony irradiation: Do senescent zooid resorption and immunological resorption involve similar recognition events

    International Nuclear Information System (INIS)

    Rinkevich, B.; Weissman, I.L.

    1990-01-01

    The colonial tunicate Botryllus schlosseri undergoes cyclic blastogenesis where feeding zooids are senescened and resorbed and a new generation of zooids takes over the colony. When non-identical colonies come into direct contact, they either reject each other or fuse. Fusion is usually followed by the resorption of one of the partners in the chimera (immunological resorption). The striking morphological similarities between the two resorption phenomena suggest that both may involve tissue destruction following self-nonself recognition events. Here we attempt to modify these two events by whole colony gamma irradiation assays. Three sets of experiments were performed: (1) different doses of whole colony irradiation for determination of irradiation effects (110 colonies); (2) pairs of irradiated-nonirradiated isografts of clonal replicates for the potential of reconstruction of the irradiated partners (23 pairs); (3) chimeras of irradiated-nonirradiated partners for analysis of resorption hierarchy. Mortality increased with the irradiation dose. All colonies exposed to more than 5,000 rads died within 19 days, while no colony died below 2,000 rads. The average mortality periods, in days, for doses of 6,000-8,000, 5,000, and 2,500-4,000 rads were 14.4 +/- 3.1 (n = 24), 19.8 +/- 6.0 (n = 15), and 19.6 + 5.1 (n = 22), respectively. Younger colonies (3-6 months old) may survive radiation better than older ones (more than 13 months). Many morphological alterations were recorded in irradiated colonies: ampullar contraction and/or dilation, accumulation of pigment cells within ampullae, abnormal bleeding from blood vessels, sluggish blood circulation, necrotic zones, reduction in bud number, and irregularities in zooid and system structures. With doses of 3,000-4,000 rads and above, irradiation arrested the formation of new buds and interrupted normal takeover

  8. Application of electron irradiation to food containers and packaging materials

    International Nuclear Information System (INIS)

    Ueno, Koji

    2010-01-01

    Problems caused by microbial contamination and hazardous chemicals have attracted much attention in the food industry. The number of systems such as hygienic management systems and Hazard Analysis Critical Control Point (HACCP) systems adopted in the manufacturing process is increasing. As manufacturing process control has become stricter, stricter control is also required for microbial control for containers and packaging materials (from disinfection to sterilization). Since safe and reliable methods for sterilizing food containers and packaging materials that leave no residue are required, electron beam sterilization used for medical equipment has attracted attention from the food industry. This paper describes an electron irradiation facility, methods for applying electron beams to food containers and packaging materials, and products irradiated with electron beams. (author)

  9. Conceptual Design Report for the Irradiated Materials Characterization Laboratory (IMCL)

    Energy Technology Data Exchange (ETDEWEB)

    Stephanie Austad

    2010-06-01

    This document describes the design at a conceptual level for the Irradiated Materials Characterization Laboratory (IMCL) to be located at the Materials and Fuels Complex (MFC) at the Idaho National Laboratory (INL). The IMCL is an 11,000-ft2, Hazard Category-2 nuclear facility that is designed for use as a state of the-art nuclear facility for the purpose of hands-on and remote handling, characterization, and examination of irradiated and nonirradiated nuclear material samples. The IMCL will accommodate a series of future, modular, and reconfigurable instrument enclosures or caves. To provide a bounding design basis envelope for the facility-provided space and infrastructure, an instrument enclosure or cave configuration was developed and is described in some detail. However, the future instrument enclosures may be modular, integral with the instrument, or reconfigurable to enable various characterization environments to be configured as changes in demand occur. They are not provided as part of the facility.

  10. Segmented fuel irradiation program: investigation on advanced materials

    International Nuclear Information System (INIS)

    Uchida, H.; Goto, K.; Sabate, R.; Abeta, S.; Baba, T.; Matias, E. de; Alonso, J.

    1999-01-01

    The Segmented Fuel Irradiation Program, started in 1991, is a collaboration between the Japanese organisations Nuclear Power Engineering Corporation (NUPEC), the Kansai Electric Power Co., Inc. (KEPCO) representing other Japanese utilities, and Mitsubishi Heavy Industries, Ltd. (MHI); and the Spanish Organisations Empresa Nacional de Electricidad, S.A. (ENDESA) representing A.N. Vandellos 2, and Empresa Nacional Uranio, S.A. (ENUSA); with the collaboration of Westinghouse. The objective of the Program is to make substantial contribution to the development of advanced cladding and fuel materials for better performance at high burn-up and under operational power transients. For this Program, segmented fuel rods were selected as the most appropriate vehicle to accomplish the aforementioned objective. Thus, a large number of fuel and cladding combinations are provided while minimising the total amount of new material, at the same time, facilitating an eventual irradiation extension in a test reactor. The Program consists of three major phases: phase I: design, licensing, fabrication and characterisation of the assemblies carrying the segmented rods (1991 - 1994); phase II: base irradiation of the assemblies at Vandellos 2 NPP, and on-site examination at the end of four cycles (1994-1999). Phase III: ramp testing at the Studsvik facilities and hot cell PIE (1996-2001). The main fuel design features whose effects on fuel behaviour are being analysed are: alloy composition (MDA and ZIRLO vs. Zircaloy-4); tubing texture; pellet grain size. The Program is progressing satisfactorily as planned. The base irradiation is completed in the first quarter of 1999, and so far, tests and inspections already carried out are providing useful information on the behaviour of the new materials. Also, the Program is delivering a well characterized fuel material, irradiated in a commercial reactor, which can be further used in other fuel behaviour experiments. The paper presents the main

  11. TEM examination of irradiated zircaloy-2 pressure tube material

    International Nuclear Information System (INIS)

    Srivastava, D.; Tewari, R.; Dey, G.K.; Sharma, B.P.; Sah, D.N.; Banerjee, Suparna; Sahoo, K.C.

    2005-09-01

    In the present work, microstructure of the zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water RAPP-1. Reactor (PHWR) has been examined for the first time using transmission electron microscope (TEM). The samples were obtained from a zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit -1, for a period for 6.77 effective full power years (EFPYs) and expected to have a cumulative radiation damage of about 3 dpa. In this study irradiated microstructure has been characterized and compared it with the microstructure of the unirradiated pressure tube samples. The effect of irradiation on the hydriding behaviour is also studied. (author)

  12. Irradiation of aluminium alloy materials with electron beam

    International Nuclear Information System (INIS)

    Konno, Osamu; Masumoto, Kazuyoshi

    1982-01-01

    It is a theme with a room for discussion to employ the stainless steel composed of longer half-life materials for the vacuum system of accelerators, from the viewpoint of radiation exposure. Therefore, it is desirable to use aluminium of shorter half-life in place of stainless steel. As a result of investigation on the above theme in the 1.2 GeV electron linac project in Tohoku University, it has been concluded that aluminium alloy vacuum chambers can reduce exposure dose by about one or two figures as compared with stainless steel ones. Of course, aluminium alloy contains trace amounts of Mg, Si, Ti, Cr, Mn, Fe, Zn, Cu and others. Therefore, four kinds of aluminium alloy considered to be usable have been examined for induced radioactivity by electron beam irradiation. Stainless steel SUS 304 has been also irradiated for comparison. Radiation energy has been 30 MeV and 200 MeV. When stainless steel and aluminium alloy were compared, aluminium alloy was very effective for reducing surface dose in low energy irradiation. In 200 MeV irradiation, the dose ratio of aluminium alloy to stainless steel became 1/30 to 1/100 after one week, though the dose difference between these two materials became smaller in 100 days or more after irradiation. If practical inspection and repair are implemented during the period from a few days to one week after shutdown, the aluminium alloy is preferable for exposure dose reduction even in high energy irradiation. (Wakatsuki, Y.)

  13. Project accent: graphite irradiated creep in a materials test reactor

    International Nuclear Information System (INIS)

    Brooking, M.

    2014-01-01

    Atkins manages a pioneering programme of irradiation experiments for EDF Energy. One of these projects is Project ACCENT, designed to obtain evidence of a beneficial physical property of the graphite, which may extend the life of the Advanced Gas-cooled Reactors (AGRs). The project team combines the in-house experience of EDF Energy with two supplier organisations (providing the material test reactors and testing facilities) and supporting consultancies (Atkins and an independent technical expert). This paper describes: - Brief summary of the Project; - Discussion of the challenges faced by the Project; and - Conclusion elaborating on the aims of the Project. These challenging experiments use bespoke technology and both un-irradiated (virgin) and irradiated AGR graphite. The results will help to better understand graphite irradiation-induced creep (or stress modified dimensional change) properties and therefore more accurately determine lifetime and safe operating envelopes of the AGRs. The first round of irradiation has been completed, with a second round about to commence. This is a key step to realising the full lifetime ambition for AGRs, demonstrating the relaxation of stresses within the graphite bricks. (authors)

  14. Thermal analysis of the APT materials irradiation samples

    International Nuclear Information System (INIS)

    Maloy, S.A.; Willcutt, G.J.; James, M.R.; Teague, J.; Diebe, D.A.; Sommer, W.F.; Ferguson, P.D.

    1998-01-01

    The accelerator production of tritium (APT) project proposes to use a 1.7 GeV, 100 mA proton beam to produce neutrons from an Inconel 718 clad tungsten target. The neutrons are multiplied and moderated in a lead/water blanket before being captured in He 3 to form tritium. In this process, the materials in the target and blanket region are exposed to a wide range of different fluxes comprised of protons and neutrons with energies into the GeV range. To investigate the effect of irradiation on the mechanical properties of candidate APT materials (Inconel 718, 316L stainless steel, Al 6061-T6, Mod 9Cr-1Mo, 304L stainless steel and Al5052-0), the APT Engineering Design and Development group fielded an extensive materials irradiation using the LANSCE (Los Alamos Neutron Science Center) accelerator, which operates at an energy of 800 MeV and a current of 1 mA. The test set-up was designed to place mechanical test specimens in locations in and near the proton beam where the environment of proton and neutron fluxes and temperatures are prototypic to those expected in the APT target/blanket (50--170 C). After irradiating for about 3,600 hours, the maximum achieved proton fluence was 4--5 x 10 21 p/cm 2 for the materials in the center of the beam. To obtain relevant data on the change in the mechanical properties with fluence, it is essential to know the temperature at which the materials were irradiated. This paper explains the method of determining the specimen temperature and reports some specific examples

  15. Calculations on neutron irradiation damage in reactor materials

    International Nuclear Information System (INIS)

    Sone, Kazuho; Shiraishi, Kensuke

    1976-01-01

    Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

  16. Contamination confinement system of irradiated materials handling laboratories

    International Nuclear Information System (INIS)

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

    1988-06-01

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

  17. Construction of irradiated material examination facility-basic design

    International Nuclear Information System (INIS)

    Ro, Seung Gy; Kim, Eun Ka; Hong, Gye Won; Herr, Young Hoi; Hong, Kwon Pyo; Lee, Myeong Han; Baik, Sang Youl; Choo, Yong Sun; Baik, Seung Je

    1989-02-01

    The basic design of the hot cell facility which has the main purpose of doing mechanical and physical property tests of irradiated materials, the examination process, and the annexed facility has been made. Also basic and detall designs for the underground excavation work have been performed. The project management and tasks required for the license application have been carried out in due course. The facility is expected to be completed by the end of 1992, if the budgetary support is sufficient. (Author)

  18. Containment system of contamination in irradiated materials handling laboratories

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  19. Microstructural evolution of CANDU spacer material Inconel X-750 under in situ ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, He Ken [Department of Mechanical and Materials Engineering, Queen’s University Kingston, Ontario K7L 3N6 (Canada); Yao, Zhongwen, E-mail: yaoz@me.queensu.ca [Department of Mechanical and Materials Engineering, Queen’s University Kingston, Ontario K7L 3N6 (Canada); Judge, Colin; Griffiths, Malcolm [Deformation Technology Branch, AECL, Chalk River Laboratories Chalk River, Ontario K0J 1J0 (Canada)

    2013-11-15

    Highlights: •γ′ Disordered at low dose. •Cascade induced SFTs were observed in alloy X-750. •No cavities were found from mono heavy ions irradiated samples. -- Abstract: Work on Inconel® X-750 spacers removed from CANDU® reactors has shown that they become embrittled and there is development of many small cavities within the metal matrix and along grain boundaries. In order to emulate the neutron irradiation induced microstructural changes, heavy ion irradiations (1 MeV Kr{sup 2+} ions) were performed while observing the damage evolution using an intermediate voltage electron microscope (IVEM) operating at 200 kV. The irradiations were carried out at various temperatures 60–400 °C. The principal strengthening phase, γ′, was disordered at low doses (∼0.06 dpa) during the irradiation. M{sub 23}C{sub 6} carbides were found to be stable up to 5.4 dpa. Lattice defects consisted mostly of stacking fault tetrahedras (SFTs), 1/2<1 1 0> perfect loops and small 1/3<1 1 1> faulted Frank loops. The ratio of SFT number density to loop number density for each irradiation condition was found to be neither temperature nor dose dependent. Under the operation of the ion beam the SFT production was very rapid, with no evidence for further growth once formed, indicating that they probably formed as a result of cascade collapse in a single cascade. The number density of the defects was found to saturate at low dose (∼0.68 dpa). No cavities were observed regardless of the irradiation temperature between 60 °C and 400 °C for doses up to 5.4 dpa. In contrast, cavities have been observed after neutron irradiation in the same material at similar doses and temperatures indicating that helium, produce during neutron irradiation, may be essential for the nucleation and growth of cavities.

  20. Irradiation effects on the ductility of fusion reactor structural materials

    International Nuclear Information System (INIS)

    Boudamous, F.

    1986-10-01

    Austenitic and ferritic-martensitic stainless steels have been proposed as first wall structural materials for the next generation of fusion devices. In order to study the effect of high temperature irradiation on their tensile properties, specimens of the steel AISI 316 L (CEC reference), of the martensitic steel W. Nr 1.4914 and of the duplex ferritic-martensitic steel EM12 have been irradiated in the BR2 reactor in Mol. The austenitic steel was irradiated at 470 0 C to about 1.1 10 22 n/cm 2 ( E>0.1 MeV) while the ferritic-martensitic steels were irradiated at 590 0 C to about 7.7 10 22 n/cm 2 (E>0.1 MeV). The tensile tests of the 316 L steel have been performed between 250 and 750 0 C. Below around 550 0 C, the yield stress after irradiation increased from about 160 to 270 MPa and the total elongation decreased from 42 to about 26%. At 750 0 C, the yield stress increase was small but the total elongation decreased from 60 to only 10%. At this temperature, the rupture of the irradiated specimen was intergranular while all the other specimens presented a transgranular rupture. At 650 0 C the variations were intermediate. The change of the ultimate tensile strength was small at all test temperatures. The EM12 and W. Nr 1.4914 steels tested only at 550 0 C, showed a decrease of the yield and tensile strength as well as an increase of the total elongation. The same tests performed on specimens which have been heat treated in parallel showed that the observed changes were due, in a large part, if not completely, to the maintenance of steels at high temperature

  1. Nanostructured Solar Irradiation Control Materials for Solar Energy Conversion

    Science.gov (United States)

    Kang, Jinho; Marshall, I. A.; Torrico, M. N.; Taylor, C. R.; Ely, Jeffry; Henderson, Angel Z.; Kim, J.-W.; Sauti, G.; Gibbons, L. J.; Park, C.; hide

    2012-01-01

    Tailoring the solar absorptivity (alpha(sub s)) and thermal emissivity (epsilon(sub T)) of materials constitutes an innovative approach to solar energy control and energy conversion. Numerous ceramic and metallic materials are currently available for solar absorbance/thermal emittance control. However, conventional metal oxides and dielectric/metal/dielectric multi-coatings have limited utility due to residual shear stresses resulting from the different coefficient of thermal expansion of the layered materials. This research presents an alternate approach based on nanoparticle-filled polymers to afford mechanically durable solar-absorptive and thermally-emissive polymer nanocomposites. The alpha(sub s) and epsilon(sub T) were measured with various nano inclusions, such as carbon nanophase particles (CNPs), at different concentrations. Research has shown that adding only 5 wt% CNPs increased the alpha(sub s) and epsilon(sub T) by a factor of about 47 and 2, respectively, compared to the pristine polymer. The effect of solar irradiation control of the nanocomposite on solar energy conversion was studied. The solar irradiation control coatings increased the power generation of solar thermoelectric cells by more than 380% compared to that of a control power cell without solar irradiation control coatings.

  2. Preparation of silica-based hybrid materials by gamma irradiation

    International Nuclear Information System (INIS)

    Gomes, S.R.; Margaca, F.M.A.; Miranda Salvado, I.M.; Ferreira, L.M.; Falcao, A.N.

    2006-01-01

    Gamma-ray irradiation is well known to promote the crosslinking of polymer chains. The method is now used by the authors to prepare hybrid materials from a mixture of polymer and metallic alkoxides of silicium and zirconium that are usually obtained via the sol-gel process. Macroscopically homogeneous and transparent hybrid materials have been obtained by γ-irradiation of polydimethylsiloxane (PDMS), tetraethylorthosilicate (TEOS) and zirconium propoxide (PrZr). The influence of several parameters has been studied. The dose rate was found to have no significant impact in the prepared material. The polymer molecular weight was also observed not to play any special role. It was found that all irradiated samples consist of a polymer gel matrix. In the case where both alkoxides are present there are inorganic oxide regions linked to the PDMS network. However when one of the alkoxides is absent there is no formation of inorganic oxide regions linked to the polymer matrix, there being only a few individual derived molecules of the other alkoxide linked to the polymer

  3. Development of Environment and Irradiation Effects of High Temperature Materials

    International Nuclear Information System (INIS)

    Ryu, Woo Seog; Kim, D. W.; Kim, S. H.

    2009-11-01

    Proposed materials, Mod.9Cr-1Mo steel (32 mm thickness) and 9Cr-1Mo-1W (100 mm thickness), for the reactor vessel were procured, and welded by the qualified welding technologies. Welding soundness was conformed by NDT, and mechanical testings were done along to weld depth. Two new irradiation capsules for use in the OR test hole of HANARO were designed and fabricated. specimens was irradiated in the OR5 test hole of HANARO with a 30MW thermal power at 390±10 .deg. C up to a fast neutron fluence of 4.4x10 19 (n/cm 2 ) (E>1.0 MeV). The dpa was evaluated to be 0.034∼0.07. Base metals and weldments of both Mod.9Cr-1Mo and 9Cr-1Mo-1W steels were tested tensile and impact properties in order to evaluate the irradiation hardening effects due to neutron irradiation. DBTT of base metal and weldment of Mod.9Cr-1Mo steel were -16 .deg. C and 1 .deg. C, respectively. After neutron irradiation, DBTT of weldment of Mod.9Cr-1Mo steel increased to 25 . deg. C. Alloy 617 and several nickel-base superalloys were studied to evaluate high temperature degradation mechanisms. Helium loop was developed to evaluate the oxidation behaviors of materials in the VHTR environments. In addition, creep behaviors in air and He environments were compared, and oxidation layers formed outer surfaces were measured as a function of applied stress and these results were investigated to the creep life

  4. Effects of irradiation on four solid breeder materials

    International Nuclear Information System (INIS)

    Hollenberg, G.W.

    1984-01-01

    The tritium breeding material with the highest lithium atom density, Li 2 O has been observed to incur significant swelling (>4%) under fast reactor irradiation. Such swelling, if unrestrained leads to either unacceptable, induced-strains in adjacent structural material or undesirable design compromises. Fortunately, however, Li 2 O deforms at low temperatures so that swelling strains may be internally accommodated. Laboratory dilational creep experiments were conducted on unirraciated Li 2 O between 500 and 700 0 C in order to provide data for structural analysis of in-reactor experiments and blanket design studies. A densification model agreed with most of the available data. 15 refs

  5. Optical and electrical phenomena in dielectric materials under irradiation

    CERN Document Server

    Plaksin, O A; Stepanov, P A; Demenkov, P V; Chernov, V M; Krutskikh, A O

    2002-01-01

    Optical and acoustic properties of the materials based on Al sub 2 O sub 3 , SiO sub 2 and BN under 8 MeV proton irradiation (<10 sup 4 Gy/s) have been measured. Electric charge partitioning has been shown to result in charging the microscopic regions in the bulk of the dielectrics under irradiation, which is due to different mobility of free electrons and holes (sapphire), concentration inhomogeneity in the system of charge carrier traps (alumina), or thermodynamic instability of the homogeneous distribution of the filled traps (silica glasses). Prevalent charge carrier recombination in the grain boundaries causes re-crystallization of pyrolytic boron nitride under irradiation, which shows up as simultaneous decrease of the intensity of radiation-induced luminescence (RIL) of the centres in the grain boundaries and the BN. The local charging results in optical inhomogeneity of the silica glasses which is sustained by the optical loss spectra of the irradiated glasses, features of kinetics of bleaching, RI...

  6. Gamma irradiation induced effects of butyl rubber based damping material

    Science.gov (United States)

    Chen, Hong-Bing; Wang, Pu-Cheng; Liu, Bo; Zhang, Feng-Shun; Ao, Yin-Yong

    2018-04-01

    The effects of gamma irradiation on the butyl rubber based damping material (BRP) at various doses in nitrogen were investigated in this study. The results show that irradiation leads to radiolysis of BRP, with extractives increasing from 14.9 ± 0.8% of control to 37.2 ± 1.2% of sample irradiated at 350 kGy, while the swelling ratio increasing from 294 ± 3% to 766 ± 4%. The further investigation of the extractives with FTIR shows that the newly generated extractives are organic compounds containing C-H and C˭C bonds, with molecular weight ranging from 26,500 to 46,300. SEM characterization shows smoother surface with holes disappearing with increasing absorbed doses, consistent with "softer" material because of radiolysis. Dynamic mechanical study of BRP show that tan δ first slightly then obviously increases with increasing absorbed dose, while storage modulus slightly decreases. The tensile testing shows that the tensile strength decreases while the elongation at break increases with increasing dose. The positron annihilation lifetime spectroscopy show no obvious relations between free volume parameters and the damping properties, indicating the complicated influencing factors of damping properties.

  7. Small Punch Test Techniques for Irradiated Materials in Hot Cell

    International Nuclear Information System (INIS)

    Kim, Do Sik; Ahn, S. B.; Oh, W. H.; Yoo, B. O.; Choo, Y. S.

    2006-06-01

    Detailed procedures of the small punch test including the apparatus, the definition of small punch-related parameters, and the interpretation of results were presented. The testing machine should have a capability of the compressive loading and unloading at a given deflection level. The small punch specimen holder consists of an upper and lower die and clamping screws. The clamped specimen is deformed by using ball or spherical head punch. Two type of specimens with a circular and a square shape were used. The irradiated small punch specimen is made from the undamaged portion of the broken CVN bars or prepared by the irradiation of the specimen fabricated from the fresh materials. The heating and cooling devices should have the capability of the temperature control within ±2 .deg. C for the target value during the test. Based on the load-deflection data obtained from the small punch test. the empirical correlation between the small punch related parameters and a tensile properties such as 0.2% yield strength and ultimate tensile strength, fracture toughness, ductile-brittle transition temperature and creep properties determined from the standard test method is established and used to evaluate the mechanical properties of an irradiated materials. In addition, from the quantitative fractographic assessment of small punch test specimens, the relationship between the small punch energy and the quantity of ductile crack growth is obtained. Analytical formulations demonstrated good agreement with experimental load-deflection curves

  8. Irradiation-accelerated corrosion of reactor core materials

    International Nuclear Information System (INIS)

    Bartels, David; Was, Gary; Jiao, Zhijie

    2012-09-01

    The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, but also applies to most all other GenIV concepts. Of these four drivers, the combination of radiation and corrosion presents a unique and extremely challenging environment for materials, for which an understanding of the fundamental science is essentially absent. Irradiation can affect corrosion or oxidation in at least three different ways. Radiation interaction with water results in the decomposition of water into radicals and oxidizing species that will increase the electrochemical corrosion potential and lead to greater corrosion rates. Irradiation of the solid surface can produce excited states that can alter corrosion, such as in the case of photo-induced corrosion. Lastly, displacement damage in the solid will result in a high flux of defects to the solid-solution interface that can alter and perhaps, accelerate interface reactions. While there exists reasonable understanding of how corrosion is affected by irradiation of the aqueous environment, there is little understanding of how irradiation affects corrosion through its impact on the solid, whether metal or oxide. The reason is largely due to the difficulty of conducting experiments that can measure this effect separately. We have undertaken a project specifically to separate the several effects of irradiation on the mechanisms of corrosion. We seek to answer the question: How does radiation damage to the solution-oxide couple affect the oxidation process differently from radiation damage to either component alone? The approach taken in this work is to closely compare corrosion accelerated by (1) proton irradiation, (2) electron irradiation, and (3) chemical corrosion potential effects alone, under typical PWR operating conditions at 300 deg. C. Both 316 stainless steel and zirconium are to be studied. The proton

  9. Staged deployment of the International Fusion Materials Irradiation Facility

    International Nuclear Information System (INIS)

    Takeuchi, H.; Sugimoto, M.; Nakamura, H.

    2001-01-01

    The International Fusion Materials Irradiation Facility (IFMIF) employs an accelerator based D-Li intense neutron source as defined in the 1995-96 Conceptual Design Activity (CDA) study. In 1999, IEA mandated a review of the CDA IFMIF design for cost reduction without change to its original mission. This objective was accomplished by eliminating the previously assumed possibility of potential upgrade of IFMIF beyond the user requirements. The total estimated cost was reduced from $797.2 M to $487.8 M. An option of deployment in 3 stages was also examined to reduce the initial investment and annual expenditures during construction. In this scenario, full performance is achieved gradually with each interim stage as follows. 1st Stage: 20% operation for material selection for ITER breeding blanket, 2nd Stage: 50% operation to demonstrate materials performance of a reference alloy for DEMO, 3rd Stage: full performance operation ( 2MW/m 2 at 500cm 3 ) to obtain engineering data for potential DEMO materials under irradiation up to 100-200 dpa. In summary, the new, reduced cost IFMIF design and staged deployment still satisfies the original mission. The estimated cost of the 1st Stage facility is only $303.6 M making it financially much more attractive. Currently, IFMIF Key Element Technology Phase (KEP) is underway to reduce the key technology risk factors. (author)

  10. Neutron-Irradiated Samples as Test Materials for MPEX

    International Nuclear Information System (INIS)

    Ellis, Ronald James; Rapp, Juergen

    2015-01-01

    Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility

  11. Assessment of repair welding technologies of irradiated materials

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Damages on reactor internals of stainless steels caused by stress corrosion cracking and fatigue were identified in aged BWR plants. Repair-welding is one of the practical countermeasure candidates to restore the soundness of components and structures. The project of 'Assessment of Repair welding Technologies of Irradiated Materials' has been carried out to develop the technical guideline regarding the repair-welding of reactor internals. In FY 2011, we investigated the fatigue strength of stainless steel SUS316L irradiated by YAG laser welding. Furthermore, revision of the technical guideline regarding the repair-welding of reactor internals was discussed. Diagram of tungsten inert gas (TIG) weld cracking caused by entrapped Helium was modified. Helium concentration for evaluation-free of TIG weld cracking caused by entrapped Helium was revised to 0.007appm from 0.01appm. (author)

  12. Irradiation tests on bitumen and bitumen coated materials

    International Nuclear Information System (INIS)

    Tabardel-Brian, R.; Rodier, J.; Lefillatre, G.

    1969-01-01

    The use of bitumen as a material for coating high-activity products calls for prior study of the resistance of bitumen to irradiation. After giving briefly the methods of preparation of bitumen- coated products, this report lists the equipment which has been used for carrying out the β and γ irradiations of these products, and gives the analytical results obtained as a function of the dose rates chosen and of the total integrated dose. Finally, some conclusions have been drawn concerning the best types of bitumen. It should be stressed that some bitumens apparently underwent no degradation whatsoever nor any volume increase, for a total integrated dose of 1.8 x 10 10 rads. (authors) [fr

  13. Process for the irradiation of a film-like material

    International Nuclear Information System (INIS)

    Takimoto, Kazuo; Inoue, Takashi.

    1969-01-01

    Herein provided is a process for curing a polymerizable coating applied to a strip-like material by irradiating the film with high energy radiation. A plurality of rollers are arranged on both sides of the radiation path in a rectangular configuration such that only the underside of the film contacts the rollers as it is unwound in spiral fashion from a feed bobbin and rewound by a take-up bobbin located within the rectangle. The rollers are further positioned to feed the film in a direction perpendicular to the radiation beam path and to assure that successive levels of the strip superimposed while being inwardly wound are mutually parallel, uniformly spaced and adjusted to precisely intercept the radiation beam. Such an arrangement prevents a polymerizable liquid coating applied to the surface of the strip from contacting the rollers and allows effective repetitive irradiation of the strip as it passes through successive levels of the spiral before being rewound. (Owens, K. J.)

  14. Dielectric changes in neutron-irradiated rf window materials

    International Nuclear Information System (INIS)

    Frost, H.M.; Clinard, F.W. Jr.

    1987-01-01

    Ceramics used for windows in ECRH heating systems for magnetically-confined fusion reactors must retain adequate properties during and after intense neutron irradiation. Of particular concern is a decrease in transmissivity, a parameter inversely related to the product of dielectric constant K and loss tangent tanδ. Samples of polycrystalline Al 2 O 3 and BeO were irradiated to 1 x 10 26 n/m 2 at 660K in the EBR-II fission reactor, and the above properties subsequently measured at 95 GHz. It was found that ktanδ for both materials doubled, implying a doubling of thermal stresses and a consequent reduction of time-to-failure from an assumed one year to 20 min for beryllia and 2 s for alumina. In the case of BeO, a large increase in reflectance of the incident millimeter-wave power results from dielectrically uncompensated swelling. This phenomenon could significantly degrade source performance

  15. Testing capabilities of Los Alamos National Laboratory for irradiated materials

    International Nuclear Information System (INIS)

    Maloy, S.A.; James, M.R.; Sommer, W.F.

    1999-01-01

    Spallation neutron sources expose materials to high energy (>100 MeV) proton and neutron spectra. Although numerous studies have investigated the effects of radiation damage in a lower energy neutron flux from fission or fusion reactors on the mechanical properties of materials, very little work has been performed on the effects that exposure to a spallation neutron spectrum has on the mechanical properties of materials. These effects can be significantly different than those observed in a fission or fusion reactor spectrum because exposure to high energy protons and neutrons produces more He and H along with the atomic displacement damage. Los Alamos National Laboratory has unique facilities to study the effects of spallation radiation damage on the mechanical properties of materials. The Los Alamos Neutron Science Center (LANSCE) has a pulsed linear accelerator which operates at 800 MeV and 1 mA. The Los Alamos Spallation Radiation Effect Facility (LASREF) located at the end of this accelerator is designed to allow the irradiation of components in a proton beam while water cooling these components and measuring their temperature. After irradiation, specimens can be investigated at hot cells located at the Chemical Metallurgy Research Building. Wing 9 of this facility contains 16 hot cells set up in two groups of eight, each having a corridor in the center to allow easy transfer of radioactive shipments into and out of the hot cells. These corridors have been used to prepare specimens for shipment to collaborating laboratories such as PNNL, ORNL, BNL, and the Paul Scherrer Institute to perform specialized testing at their hot cells. The LANL hot cells contain capabilities for opening radioactive components and testing their mechanical properties as well as preparing specimens from irradiated components

  16. Radiation damage and materials performance in irradiation environment

    International Nuclear Information System (INIS)

    Singh, B.N.

    2009-01-01

    Collisions of energetic projectile particles with host atoms produce atomic displacements in the target materials. Subsequently, some of these displacements are transformed into lattice defects and survive in the form of single defects and of defect clusters. Depending on the ambient temperature, these defects and their clusters diffuse, interact, annihilate, segregate and accumulate in various forms and are responsible for the evolution of the irradiation-induced microstructure. Naturally, both physical and mechanical properties and thereby the performance and lifetime of target materials are likely to be determined by the nature and the magnitude of the accumulated defects and their spatial dispositions. The defect accumulation, microstructural evolution and the resulting materials response gets very complicated particularly under the reactor operational conditions. The complication arises from the fact that the materials used in the structural components will experience concurrently generation of defects produced by the flux of neutrons and generation of dislocations due to plastic deformation. In other words, the defect accumulation will have to be considered under the conditions of two interactive reaction kinetics operating simultaneously. Both materials and experimental variables are likely to affect the damage accumulation and thereby the materials performance. Experimental and theoretical results pertaining to effects of major materials and experimental variables on materials performance will be briefly examined. (au)

  17. Study of PDMS conformation in PDMS-based hybrid materials prepared by gamma irradiation

    International Nuclear Information System (INIS)

    Lancastre, J.J.H.; Fernandes, N.; Margaça, F.M.A.; Miranda Salvado, I.M.; Ferreira, L.M.; Falcão, A.N.; Casimiro, M.H.

    2012-01-01

    Polydimethylsiloxane-silicate based hybrid materials have recognized properties (high flexibility, low elastic modulus or high mechanical strength) for which there are a large number of applications in development, such as for the bioapplications field. The hybrids addressed in the present study were prepared by gamma irradiation of a mixture of polydimethylsiloxane (PDMS) with tetraethylorthosilicate (TEOS) and zirconium propoxide (PrZr) without addition of any solvent or other product. The materials are homogeneous, transparent, monolithic and flexible. The structure dependence on the PrZr content is addressed. A combination of X-ray diffraction (XRD) and Infrared Spectroscopy (IR) was used. The results reveal that the polymer in the hybrids prepared with PrZr, in a content≤5 wt%, shows a structure similar to that in the irradiated pure polymer sample. In these samples the presence of ordered polymer regions is clearly found. For samples prepared with higher content of Zr almost no ordered polymer regions are observed. The addition of PrZr plays an important role on polymer conformation in these hybrid materials. - Highlights: ► PDMS-based hybrid materials were prepared by γ-irradiation. ► FTIR, ATR/FT-IR and XRD techniques were used to characterize the materials. ► Changes in FTIR bands reflect growth of crosslinking network. ► Above certain Zr concentration regions of Zr-silicate oxide are formed. ► Zr content determines conformation of the polymer chain network.

  18. The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions

    Energy Technology Data Exchange (ETDEWEB)

    Garrison, L. M., E-mail: garrisonlm@ornl.gov; Egle, B. J. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, Tennessee 37831 (United States); Fusion Technology Institute, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, Wisconsin 53706 (United States); Zenobia, S. J.; Kulcinski, G. L.; Santarius, J. F. [Fusion Technology Institute, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, Wisconsin 53706 (United States)

    2016-08-15

    The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000 °C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ion gun can irradiate the samples with ion currents of 20 μA–500 μA; the typical current used is 72 μA, which is an average flux of 9 × 10{sup 14} ions/(cm{sup 2} s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. The MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.

  19. Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 2000C

    International Nuclear Information System (INIS)

    Conlin, J.A.

    1985-08-01

    A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 200 0 C. The design description and results of the prototype capsule performance are presented

  20. Neutron irradiation facilities for fission and fusion reactor materials studies

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1985-01-01

    The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

  1. IFMIF [International Fusion Materials Irradiation Facility], an accelerator-based neutron source for fusion components irradiation testing: Materials testing capabilities

    International Nuclear Information System (INIS)

    Mann, F.M.

    1988-08-01

    The International Fusion Materials Irradiation Facility (IFMIF) is proposed as an advanced accelerator-based neutron source for high-flux irradiation testing of large-sized fusion reactor components. The facility would require only small extensions to existing accelerator and target technology originally developed for the Fusion Materials Irradiation Test (FMIT) facility. At the extended facility, neutrons would be produced by a 0.1-A beam of 35-MeV deuterons incident upon a liquid lithium target. The volume available for high-flux (>10/sup 15/ n/cm/sup 2/-s) testing in IFMITF would be over a liter, a factor of about three larger than in the FMIT facility. This is because the effective beam current of 35-MeV deuterons on target can be increased by a factor of ten to 1A or more. Such an increase can be accomplished by funneling beams of deuterium ions from the radio-frequency quadruple into a linear accelerator and by taking advantage of recent developments in accelerator technology. Multiple beams and large total current allow great variety in available testing. For example, multiple simultaneous experiments, and great flexibility in tailoring spatial distributions of flux and spectra can be achieved. 5 refs., 2 figs., 1 tab

  2. Dose requirements for microbial decontamination of botanical materials by irradiation

    International Nuclear Information System (INIS)

    Razem, D.; Katusin-Razem, Branka

    2002-01-01

    Microbial contamination levels and corresponding resistivities to irradiation (expressed as dose required for the first 90% reduction, D first 9 0% r ed ) were analyzed in a number of various botanical materials. The following generalizations could be made: total aerobic plate count is the most informative measure of contamination; the probability of contamination depends on available surface of the material and processing history: flowers and leaves usually contain more contamination than fruits and seeds, while crude herbs contain more than extracts; liquid extracts are more contaminated than dry ones. At the same time, resistivity to irradiation increases approximately in the reverse order of contamination level on going from flowers and leaves, to fruits and seeds, to liquid and dry extracts. The two quantities, probability of contamination and D first 9 0% r ed being inversely related, the treatment dose needed to reduce initial contamination to tolerable level amounts to between 4 and 30 kGy under a typical scenario, and between 8 and 40 kGy under the worst-case scenario for the whole range of raw materials and botanical products

  3. Determination of material irradiation parameters. Required accuracies and available methods

    International Nuclear Information System (INIS)

    Cerles, J.M.; Mas, P.

    1978-01-01

    In this paper, the author reports some main methods to determine the nuclear parameters of material irradiation in testing reactor (nuclear power, burn-up, fluxes, fluences, ...). The different methods (theoretical or experimental) are reviewed: neutronics measurements and calculations, gamma scanning, thermal balance, ... The required accuracies are reviewed: they are of 3-5% on flux, fluences, nuclear power, burn-up, conversion factor, ... These required accuracies are compared with the real accuracies available which are at the present time of order of 5-20% on these parameters

  4. Precise measurement of fuel content of irradiated and nonirradiated materials

    International Nuclear Information System (INIS)

    Harker, Y.D.; Napper, P.R.; Proctor, A.E.

    1984-01-01

    This paper discusses the application of precise reactivity measurements in the Advanced Reactivity Measurement Facility at Idaho National Engineering Laboratory (INEL) to determine th fuel content in irradiated and nonirradiated materials. Different methods of reactivity measurements and examples of how they have been are presented, which provides an insight in capabilities available to analyze samples with different geometrical sizes from small volumes approx. 100 cc to 12 ft long fuel pins and also samples with different fuel content ranges from approx. 2 mg to approx. 600 g. The overall accuracy of these measurements is approx. 0.5% (1sigma)

  5. Interfacial degradation of organic composite material by irradiation in reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nishijima, Shigehiro; Nishiura, Tetsuya; Okada, Toichi [Osaka Univ., Ibaraki (Japan). Inst. of Scientific and Industrial Research

    1996-04-01

    Glass fiber reinforced plastics (GFRP) with many kinds of matrix resins were made of E glass treated with silane as the reinforced material. Degradation of shearing strength of GFRP irradiated at low temperature was determined. It was clear from the results of comparing the degradation process with the fractured surface that the degradation was very affected by the radiation resistance of the bonded part between resin and coupling agents. It means that we had to be careful in the choice of interfacial treatments and epoxy matrices corresponded to it. (S.Y.)

  6. Shield design for the Fusion Materials Irradiation Test facility

    International Nuclear Information System (INIS)

    Carter, L.L.; Mann, F.M.; Morford, R.J.; Wilcox, A.D.; Johnson, D.L.; Huang, S.T.

    1983-03-01

    The shield design for the Fusion Materials Irradiation Test facility is based upon one-, two- and three-dimensional transport calculations with experimental measurements utilized to refine the nuclear data including the neutron cross sections from 20 to 50 MeV and the gamma ray and neutron source terms. The high energy neutrons and deuterons produce activation products from the numerous reactions that are kinematically allowed. The analyses for both beam-on and beam-off (from the activation products) conditions have required extensive nuclear data libraries and the utilization of Monte Carlo, discrete ordinates, point kernel and auxiliary computer codes

  7. Simulation of tensile stress-strain properties of irradiated type 316 SS by heavily cold-worked material

    International Nuclear Information System (INIS)

    Muto, Yasushi; Jitsukawa, Shiro; Hishinuma, Akimichi

    1995-07-01

    Type 316 stainless steel is one of the most promising candidate materials to be used for the structural parts of plasma facing components in the nuclear fusion reactor. The neutron irradiation make the material brittle and reduces its uniform elongation to almost zero at heavy doses. In order to apply such a material of reduced ductility to structural components, the structural integrity should be examined and assured by the fracture mechanics. The procedure requires a formulated stress-strain relationship. However, the available irradiated tensile test data are very limited at present, so that the cold-worked material was used as a simulated material in this study. Property changes of 316 SS, that is, a reduction of uniform elongation and an enhancement of yield stress are seemingly very similar for both the irradiated 316 SS and the cold-worked one. The specimens made of annealed 316 SS, 20% (or 15%) cold worked one and 40% cold worked one were prepared. After the formulation of stress strain behavior, the equation for the cold-worked 316 SS was fitted to the data on irradiated material under the assumption that the yield stress is the same for both materials. In addition, the upper limit for the plastic strain was introduced using the data on the irradiated material. (author)

  8. Irradiation data analysis and thermal analysis of the 02M-02K capsule for material irradiation test

    International Nuclear Information System (INIS)

    Choi, Myoung Hwan; Choo, K. N.; Kang, Y. H.; Kim, B. G.; Cho, M. S.; Sohn, J. M.; Shin, Y. T.; Park, S. J.; Kim, Y. J.

    2004-11-01

    In order to evaluate the fracture toughness of RPV materials, the material irradiation test using the instrumented capsule (02M-02K) were carried out in the HANARO in August 2003. Based on the user's requirements the thermal design analysis of the capsule 02M-02K was performed, and the specimens were suitably arranged in each step of the capsule main body. In this report, both the temperature data of specimens measured during irradiation test and the calculated data from the thermal analysis are compared and evaluated. Also, the temperature profile in each step with the HANARO reactor power and helium pressure is reviewed and evaluated. The effects of the gap size such as theoretically calculated from thermal expansion during irradiation test and measured one in the manufacturing of the capsule on the specimen temperature were reviewed. The thermal analysis was performed by using a Finite Element (FE) analysis program, ANSYS. Two-dimensional model for the 1/4 section of the capsule is generated, and the γ-heating rate of the materials used in the capsule at the control rod position of 430 mm is used as input data. The thermal analysis using a 3-dimensional model, which is quite similar to the actual shape of the capsule, is also conducted to obtain the temperature distribution in the axial direction. The analysis results show that the temperature difference between the top and bottom positions of a specimen is found to be smaller than 13.2 .deg. C. The maximum measured and calculated temperature in the step 3 of the capsule is 256 .deg. C and 264 .deg. C, respectively. The measured temperature data are obtained at the reactor power of 24 MW, the heater power of 0 W and the helium pressure of 760 torr. Generally, the temperature data obtained by the FE analysis are slightly lower than those of the measured except the step 1 of the capsule. However, the temperature difference between the measured and the calculated shows a good agreement within 9 percent. It is

  9. Spectral Classification of Similar Materials using the Tetracorder Algorithm: The Calcite-Epidote-Chlorite Problem

    Science.gov (United States)

    Dalton, J. Brad; Bove, Dana; Mladinich, Carol; Clark, Roger; Rockwell, Barnaby; Swayze, Gregg; King, Trude; Church, Stanley

    2001-01-01

    Recent work on automated spectral classification algorithms has sought to distinguish ever-more similar materials. From modest beginnings separating shade, soil, rock and vegetation to ambitious attempts to discriminate mineral types and specific plant species, the trend seems to be toward using increasingly subtle spectral differences to perform the classification. Rule-based expert systems exploiting the underlying physics of spectroscopy such as the US Geological Society Tetracorder system are now taking advantage of the high spectral resolution and dimensionality of current imaging spectrometer designs to discriminate spectrally similar materials. The current paper details recent efforts to discriminate three minerals having absorptions centered at the same wavelength, with encouraging results.

  10. Evidence of different red emissions in irradiated germanosilicate materials

    Energy Technology Data Exchange (ETDEWEB)

    Alessi, A., E-mail: antonino.alessi@univ-st-etienne.fr [Univ-Lyon, Laboratoire H. Curien, UMR CNRS 5516, Université Jean Monnet, 18 rue du Pr. Benoît Lauras, 42000 Saint-Etienne (France); Di Francesca, D. [Univ-Lyon, Laboratoire H. Curien, UMR CNRS 5516, Université Jean Monnet, 18 rue du Pr. Benoît Lauras, 42000 Saint-Etienne (France); Agnello, S. [Dipartimento di Fisica e Chimica, Università di Palermo, I-90123 Palermo (Italy); Girard, S. [Univ-Lyon, Laboratoire H. Curien, UMR CNRS 5516, Université Jean Monnet, 18 rue du Pr. Benoît Lauras, 42000 Saint-Etienne (France); Cannas, M. [Dipartimento di Fisica e Chimica, Università di Palermo, I-90123 Palermo (Italy); Richard, N. [CEA, DAM, DIF, F91297 Arpajon (France); Boukenter, A.; Ouerdane, Y. [Univ-Lyon, Laboratoire H. Curien, UMR CNRS 5516, Université Jean Monnet, 18 rue du Pr. Benoît Lauras, 42000 Saint-Etienne (France)

    2016-09-15

    This experimental investigation is focused on a radiation induced red emission in Ge doped silica materials, elaborated with different methods and processes. The differently irradiated samples as well as the pristine ones were analyzed with various spectroscopic techniques, such as confocal microscopy luminescence (CML), time resolved luminescence (TRL), photoluminescence excitation (PLE) and electron paramagnetic resonance (EPR). Our data prove that irradiation induces a red luminescence related to the presence of the Ge atoms. Such emission features a photoexcitation spectrum in the UV-blue spectral range and, TRL measurements show that its decrease differs from a single exponential law with a lifetime of tens of nanoseconds. CML measurements under laser at 633 nm evidenced the lack of correlation of the emission here reported with that of the Ge- or Si- non bridging oxygen hole centers. Moreover, our EPR experiments highlighted the lack of correlation between the red emitting defect with other radiation induced paramagnetic centers such as the E′Ge and Ge(2). The relation of the investigated emission with the H(II) defects, previously considered as responsible for a red emission, can not be totally excluded. - Highlights: • Composite nature of the red emission in Ge-doped doped silica materials. • Experimental study with various spectroscopic techniques and on different samples. • Time resolved and stationary characterization of an new red emission. • Study of the spatial distributions of diverse red emissions in optical fibers.

  11. Genetic similarity among commercial oil palm materials based on microsatellite markers

    Directory of Open Access Journals (Sweden)

    Diana Arias

    2012-08-01

    Full Text Available Microsatellite markers are used to determine genetic similarities among individuals and might be used in various applications in breeding programs. For example, knowing the genetic similarity relationships of commercial planting materials helps to better understand their responses to environmental, agronomic and plant health factors. This study assessed 17 microsatellite markers in 9 crosses (D x P of Elaeis guineensis Jacq. from various commercial companies in Malaysia, France, Costa Rica and Colombia, in order to find possible genetic differences and/or similarities. Seventy-seven alleles were obtained, with an average of 4.5 alleles per primer and a range of 2-8 amplified alleles. The results show a significant reduction of alleles, compared to the number of alleles reported for wild oil palm populations. The obtained dendrogram shows the formation of two groups based on their genetic similarity. Group A, with ~76% similarity, contains the commercial material of 3 codes of Deli x La Mé crosses produced in France and Colombia, and group B, with ~66% genetic similarity, includes all the materials produced by commercial companies in Malaysia, France, Costa Rica and Colombia

  12. Experimental Irradiations of Materials and Fuels in the BR2 Reactor: An Overview of Current Programmes

    International Nuclear Information System (INIS)

    Van Dyck, S.; Koonen, E.; Verwerft, M.; Wéber, M.

    2013-01-01

    The BR2 material test reactor offers a variety of experimental irradiation possibilities for testing of materials, fuels and instruments. The current paper gives an overview of the recent and ongoing programmes in order to illustrate the experimental potential of the reactor. Three domains of applications are reviewed: Irradiation of materials and fuels for pressurised water reactors (PWR); irradiation of materials for accelerator driven systems (ADS), cooled by liquid lead alloys; and irradiation of fuel for Material Test Reactors (MTR). For PWR relevant tests, a dedicated loop is available, providing a full simulation of the thermo hydraulic conditions of a PWR. ADS related tests require particular control of the irradiation environment and the necessary safety precautions in order to avoid 210 Po contamination. In-core mechanical testing of materials is done in comparison and complimentarily to post-irradiation examinations in order to assess flux related effects on the deformation behaviour of materials. (author)

  13. Influence of Curing Humidity on the Compressive Strength of Gypsum-Cemented Similar Materials

    Directory of Open Access Journals (Sweden)

    Weiming Guan

    2016-01-01

    Full Text Available The analogous simulation experiment is widely used in geotechnical and mining engineering. However, systematic errors derived from unified standard curing procedure have been underestimated to some extent. In this study, 140 gypsum-cemented similar material specimens were chosen to study their curing procedure with different relative humidity, which is 10%–15%, 40%, 60%, and 80%, respectively. SEM microstructures and XRD spectra were adopted to detect the correlation between microstructures and macroscopic mechanical strength during curing. Our results indicated that the needle-like phases of similar materials began to develop in the early stage of the hydration process through intersecting with each other and eventually transformed into mat-like phases. Increase of humidity may inhibit the development of needle-like phases; thus the compressive strength changes more smoothly, and the time required for the material strength to reach the peak value will be prolonged. The peak strength decreases along with the increase of humidity while the humidity is higher than 40%; however, the reverse tendency was observed if the humidity was lower than 40%. Finally, we noticed that the material strength usually reaches the peak value when the water content continuously reduces and tends towards stability. Based on the above observation, a curing method determination model and experimental strength predication method for gypsum-cemented similar materials were proposed.

  14. Characterization of damaging in apatitic materials irradiated with heavy ions and thermally annealed

    International Nuclear Information System (INIS)

    Tisserand, R.

    2004-12-01

    Some minerals belonging to the family of apatite are seen to be potential candidates for use as conditioning matrices or transmutation targets for high level nuclear waste management. Indeed, studies of natural nuclear reactors (Oklo) highlighted the strong ability of these minerals to anneal irradiation damage. In order to determine the global behaviour of these materials, we performed a fundamental study on the evolution of irradiation damage induced by various heavy ions in two apatites: a natural phospho-calcic fluor-apatite from Durango and a synthetic sintered mono-silicated fluor-apatite, called britholite. The damage in these materials was measured by using channelling R.B.S. and X-ray diffraction respectively and by determining an amorphization effective radius Re. The results revealed a similar behaviour for both apatites according to the electronic energy deposit at the entrance of the material. In addition, the effect of an isothermal annealing at 300 C was quantified on a mono-silicated britholite previously irradiated with Kr ions. We highlighted in this case the return of the lattice parameters to their initial values, followed by a partial and slow rebuilding of the crystalline lattice versus the annealing time. Finally, we followed the changes in the morphology of etch pits in the Durango fluor-apatite after acid dissolution as a function of the energy deposit by the ions. We showed that the influence of crystallography leads quickly to opening angles close to 30 degrees. The calculation of etching velocities within the irradiated material highlighted that there is a range of deposit energy where the velocity ratio increases strongly before becoming constant. (author)

  15. In-situ high temperature irradiation setup for temperature dependent structural studies of materials under swift heavy ion irradiation

    International Nuclear Information System (INIS)

    Kulriya, P.K.; Kumari, Renu; Kumar, Rajesh; Grover, V.; Shukla, R.; Tyagi, A.K.; Avasthi, D.K.

    2015-01-01

    An in-situ high temperature (1000 K) setup is designed and installed in the materials science beam line of superconducting linear accelerator at the Inter-University Accelerator Centre (IUAC) for temperature dependent ion irradiation studies on the materials exposed with swift heavy ion (SHI) irradiation. The Gd 2 Ti 2 O 7 pyrochlore is irradiated using 120 MeV Au ion at 1000 K using the high temperature irradiation facility and characterized by ex-situ X-ray diffraction (XRD). Another set of Gd 2 Ti 2 O 7 samples are irradiated with the same ion beam parameter at 300 K and simultaneously characterized using in-situ XRD available in same beam line. The XRD studies along with the Raman spectroscopic investigations reveal that the structural modification induced by the ion irradiation is strongly dependent on the temperature of the sample. The Gd 2 Ti 2 O 7 is readily amorphized at an ion fluence 6 × 10 12 ions/cm 2 on irradiation at 300 K, whereas it is transformed to a radiation-resistant anion-deficient fluorite structure on high temperature irradiation, that amorphized at ion fluence higher than 1 × 10 13 ions/cm 2 . The temperature dependent ion irradiation studies showed that the ion fluence required to cause amorphization at 1000 K irradiation is significantly higher than that required at room temperature irradiation. In addition to testing the efficiency of the in-situ high temperature irradiation facility, the present study establishes that the radiation stability of the pyrochlore is enhanced at higher temperatures

  16. Proposed rf system for the fusion materials irradiation test facility

    International Nuclear Information System (INIS)

    Fazio, M.V.; Johnson, H.P.; Hoffert, W.J.; Boyd, T.J.

    1979-01-01

    Preliminary rf system design for the accelerator portion of the Fusion Materials Irradiation Test (FMIT) Facility is in progress. The 35-MeV, 100-mA, cw deuteron beam will require 6.3 MW rf power at 80 MHz. Initial testing indicates the EIMAC 8973 tetrode is the most suitable final amplifier tube for each of a series of 15 amplifier chains operating at 0.5-MW output. To satisfy the beam dynamics requirements for particle acceleration and to minimize beam spill, each amplifier output must be controlled to +-1 0 in phase and the field amplitude in the tanks must be held within a 1% tolerance. These tolerances put stringent demands on the rf phase and amplitude control system

  17. Material irradiation techniques used in corrosion and wear analysis

    International Nuclear Information System (INIS)

    Tenreiro, Claudio

    1996-01-01

    Full text: Nuclear physics methods, applied to material analysis are discussed and some application examples are given. Experiments have been performed to study corrosion du to the presence of humidity and sulfur compounds. The use of resonant reactors allows the determination of depth profiles of H and S from structures located in particularly contaminated areas. The method provides a non destructive and quick way of estimating the effect of such elements in different types of structures, such as the ones used in high voltage transmission lines. Also the wear out rates in mechanical engine components having a difficult direct access, have been evaluated by proton activation analysis. The evaluation of the advantages of this method is being done. The effect of irradiation damage on superconducting high temperature ceramics was analyzed by the interaction of energetic alpha particles with high T c YBaCuO samples

  18. Accelerator conceptual design of the international fusion materials irradiation facility

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, M.; Kinsho, M. [Japan Atomic Energy Res. Inst., Tokai, Ibaraki (Japan). Intense Neutron Source Lab.; Jameson, R.A.; Blind, B. [Los Alamos National Lab., NM (United States); Teplyakov, V. [Institute for High Energy Physics, Moscow (Russian Federation); Berwald, D.; Bruhwiler, D.; Peakock, M.; Rathke, J. [Northrop Grumman Corp., Bethpage, NY (United States); Deitinghoff, H.; Klein, H.; Pozimski, Y.; Volk, K. [Johann Wolfgang Goethe Univ., Frankfurt (Germany). Inst. fur Angewandte Phys.; Ferdinand, R.; Lagniel, J.-M. [CEA Saclay LNS, Gif-sur-Yvette (France); Miyahara, A. [Teikyo Univ., Tokyo (Japan); Olivier, M. [CEA DSM, Saclay, Gif-sur-Yvette (France); Piechowiak, E. [Northrop Grumman Corp., Baltimore, MD (United States); Tanabe, Y. [Toshiba Corp., Tsurumi-ku, Yokohama (Japan)

    1998-10-01

    The accelerator system of the international fusion materials irradiation facility (IFMIF) provides the 250-mA, 40-MeV continuous-wave deuteron beam at one of the two lithium target stations. It consists of two identical linear accelerator modules, each of which independently delivers a 125-mA beam to the common footprint of 20 cm x 5 cm at the target surface. The accelerator module consists of an ion injector, a 175 MHz RFQ and eight DTL tanks, and rf power supply system. The requirements for the accelerator system and the design concept are described. The interface issues and operational considerations to attain the proposed availability are also discussed. (orig.) 8 refs.

  19. Accelerator conceptual design of the international fusion materials irradiation facility

    International Nuclear Information System (INIS)

    Sugimoto, M.; Kinsho, M.; Teplyakov, V.; Berwald, D.; Bruhwiler, D.; Peakock, M.; Rathke, J.; Deitinghoff, H.; Klein, H.; Pozimski, Y.; Volk, K.; Miyahara, A.; Olivier, M.; Piechowiak, E.; Tanabe, Y.

    1998-01-01

    The accelerator system of the international fusion materials irradiation facility (IFMIF) provides the 250-mA, 40-MeV continuous-wave deuteron beam at one of the two lithium target stations. It consists of two identical linear accelerator modules, each of which independently delivers a 125-mA beam to the common footprint of 20 cm x 5 cm at the target surface. The accelerator module consists of an ion injector, a 175 MHz RFQ and eight DTL tanks, and rf power supply system. The requirements for the accelerator system and the design concept are described. The interface issues and operational considerations to attain the proposed availability are also discussed. (orig.)

  20. RF structure design of the China Material Irradiation Facility RFQ

    Science.gov (United States)

    Li, Chenxing; He, Yuan; Xu, Xianbo; Zhang, Zhouli; Wang, Fengfeng; Dou, Weiping; Wang, Zhijun; Wang, Tieshan

    2017-10-01

    The radio frequency structure design of the radio frequency quadrupole (RFQ) for the front end of China Material Irradiation Facility (CMIF), which is an accelerator based neutron irradiation facility for fusion reactor material qualification, has been completed. The RFQ is specified to accelerate 10 mA continuous deuteron beams from the energies of 20 keV/u to 1.5 MeV/u within the vane length of 5250 mm. The working frequency of the RFQ is selected to 162.5 MHz and the inter-vane voltage is set to 65 kV. Four-vane cavity type is selected and the cavity structure is designed drawing on the experience of China Initiative Accelerator Driven System (CIADS) Injector II RFQ. In order to reduce the azimuthal asymmetry of the field caused from errors in fabrication and assembly, a frequency separation between the working mode and its nearest dipole mode is reached to 17.66 MHz by utilizing 20 pairs of π-mode stabilizing loops (PISLs) distributed along the longitudinal direction with equal intervals. For the purpose of tuning, 100 slug tuners were introduced to compensate the errors caused by machining and assembly. In order to obtain a homogeneous electrical field distribution along cavity, vane cutbacks are introduced and output endplate is modified. Multi-physics study of the cavity with radio frequency power and water cooling is performed to obtain the water temperature tuning coefficients. Through comparing to the worldwide CW RFQs, it is indicated that the power density of the designed structure is moderate for operation under continuous wave (CW) mode.

  1. Microwave irradiation of lignocellulosic materials, 4: Enhancement of enzymatic susceptibility of microwave-irradiated softwoods

    International Nuclear Information System (INIS)

    Azuma, J.; Higashino, J.; Isaka, M.; Koshijima, T.

    1985-01-01

    Effect of microwave irradiation on the enzymatic susceptibility of various softwoods was investigated. The pH values of the reaction liquor dropped with increasing temperature to 2.9-3.3 at 230°C, consistent with increase in acidity (0.5-0.85 meq at 230-239° C). Above approximately 180°C, hemicellulose underwent acid-mediated autohydrolysis and became water-soluble yielding a mixture of oligosaccharides and monosaccharides. The composition of water-soluble portion was similar for all wood species tested. The maximum extents of saccharification below 240°C ranged between 36-62% for softwoods, while those for hardwoods were between 88-93%. The present investigation confirmed that microwave pretreatment enhanced the enzymatic susceptibility of various softwoods. However, further attempt should be needed to give higher values equal to those for hardwoods. (author)

  2. Experimental data base for assessment of irradiation induced ageing effects in pre-irradiated RPV materials of German PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hein, H.; Gundermann, A.; Keim, E.; Schnabel, H. [AREVA NP GmbH (Germany); Ganswind, J. [VGB PowerTech e.V (Germany)

    2011-07-01

    The 5 year research program CARISMA which ended in 2008 has produced a data base to characterize the fracture toughness of pre-irradiated original RPV (Reactor Pressure Vessel) materials being representative for all four German PWR construction lines of former Siemens/KWU company. For this purpose tensile, Charpy-V impact, crack initiation and crack arrest tests have been performed for three base materials and four weld metals irradiated to neutron fluences beyond the designed EoL range. RPV steels with optimized chemical composition and with high copper as well as high nickel content were examined in this study. The RTNDT concept and the Master Curve approach were applied for the assessment of the generated data in order to compare both approaches. A further objective was to clarify in which extent crack arrest curves can be generated for irradiated materials and how crack arrest can be integrated into the Master Curve approach. By the ongoing follow-up project CARINA the experimental data base will be extended by additional representative materials irradiated under different conditions and with respect to the accumulated neutron fluences and specific impact parameters such as neutron flux and manufacturing effects. The irradiation data cover also the long term irradiation behavior of the RPV steels concerned. Moreover, most of the irradiated materials were and will be used for microstructural examinations to get a deeper insight in the irradiation embrittlement mechanisms and their causal relationship to the material property changes. By evaluation of the data base the applicability of the Master Curve approach for both crack initiation and arrest was confirmed to a large extent. Moreover, within both research programs progress was made in the development of crack arrest test techniques and in specific issues of RPV integrity assessment. (authors)

  3. Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

    International Nuclear Information System (INIS)

    Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

    2011-01-01

    The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

  4. Similar kinetics of chromatid aberrations in X-irradiated xrs 5 and wild-type Chinese hamster ovary cells

    International Nuclear Information System (INIS)

    MacLeod, R.A.F.; Bryant, P.E.

    1990-01-01

    We have studied the kinetics of chromatid aberrations in cells of the Chinese hamster ovary (CHO-K1) derived, X-ray sensitive cell line xrs 5 irradiated in the G 2 phase at 37 0 C, as well as during a cell cycle extended by transient hypothermia at 33 0 C. While a given X-ray dose was estimated to produce about 4 times as many chromatid break and twice the frequency of exchanges in xrs 5 cells as in the parent line, there was no difference between the lines in the rates of disappearance of chromatid breaks during G 2 at either temperature; and similar patterns of chromatid exchange kinetics were observed in the two lines. Both the frequencies and distributions of chromatid breaks at different times after irradiation are consistent with the view that the disappearance of these during incubation represents a repair process. These results imply that the G 2 chromosomal radiosensitivity of the xrs 5 mutant resides at the level of initial chromatid damage. (author)

  5. The reprocessing of irradiated MTR fuel and the nuclear material accountancy - Dounreay, UKAEA

    Energy Technology Data Exchange (ETDEWEB)

    Barrett, T.R.; Harrison, R. [UKAEA, Nuclear Materials Control Dep., Dounreay (United Kingdom)

    1997-07-01

    The reprocessing of irradiated HEU MTR fuel is a sensible part of a safeguards regime. It brings together fuel otherwise scattered around the world into a concerted accountancy and protection arrangement. From a nuclear material accountants view the overall accountancy performance has been excellent. While investigations have been required for a few individual MUFs or trends, very little effort has required to be expended by the Nuclear Materials Control Department. That is a definition of a 'good plant'; it operates, measures and records input and output streams, and then the accountancy falls into place. As identified in this paper, the accountancy of the nuclear material processed in the plant is well founded and sound. The accountancy results over several decades confirm the adequacy of the safeguards arrangements at Dounreay. The processing makes good commercial sense and meets the current philosophy of recycling valuable resource materials. The risk of operating the full fuel cycle are less than those of extended storage of irradiated fuel at disparate diverse locations. The reprocessing at Dounreay accords with all of these philosophies. The assessed risk is at a very low level, well within published UK HSE 'tolerability of risk' regulatory guidelines. The impact of the operations are similarly low within the guidelines, for the operators and for the general public. (author)

  6. The reprocessing of irradiated MTR fuel and the nuclear material accountancy - Dounreay, UKAEA

    International Nuclear Information System (INIS)

    Barrett, T.R.; Harrison, R.

    1997-01-01

    The reprocessing of irradiated HEU MTR fuel is a sensible part of a safeguards regime. It brings together fuel otherwise scattered around the world into a concerted accountancy and protection arrangement. From a nuclear material accountants view the overall accountancy performance has been excellent. While investigations have been required for a few individual MUFs or trends, very little effort has required to be expended by the Nuclear Materials Control Department. That is a definition of a 'good plant'; it operates, measures and records input and output streams, and then the accountancy falls into place. As identified in this paper, the accountancy of the nuclear material processed in the plant is well founded and sound. The accountancy results over several decades confirm the adequacy of the safeguards arrangements at Dounreay. The processing makes good commercial sense and meets the current philosophy of recycling valuable resource materials. The risk of operating the full fuel cycle are less than those of extended storage of irradiated fuel at disparate diverse locations. The reprocessing at Dounreay accords with all of these philosophies. The assessed risk is at a very low level, well within published UK HSE 'tolerability of risk' regulatory guidelines. The impact of the operations are similarly low within the guidelines, for the operators and for the general public. (author)

  7. A Study on the Thermal Neutron Filter for the Irradiation of Electronic Materials at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Kim, Sung Ryul; Park, Seung Jae; Shin, Yoon Taeg; Cho, Man Soon; Cho, Kee Nam [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The representative example is a technique of making the semiconductor with the transmutation using the pure Si. This NTD (Neutron Transmutation Doping) Si is used as a high-quality semiconductor because it has a uniform resistance. Likewise, the electronic materials are being investigated to improve the performance of material using the neutron irradiation method. The mechanism for reaction between the electronic materials and the neutrons depends on the energy of the neutron. Capturing reaction by thermal neutrons causes the transmutation and a lot of defects are made by fast neutrons. The study for the effect by such neutron energy is necessary to understand the performance improvement of the irradiated electronic materials. The thermal neutron filter was investigated to be used for the irradiation of electronic materials at HANARO. IP irradiation hole was selected and the irradiation device was designed. The analysis was conducted considering four candidate materials.

  8. Status of the material capsule irradiation and the development of the new capsule technology in HANARO

    International Nuclear Information System (INIS)

    Choo, Kee-Nam; Kang, Young-Hwan; Choi, Myoung-Hwan; Cho, Man-Soon; Kim, Bong-Goo

    2006-01-01

    A material capsule system including a main capsule, fixing, control, cutting, and transport systems was developed for an irradiation test of non-fissile materials in HANARO. 14 irradiation capsules (12 instrumented and 2 non-instrumented capsules) have been designed, fabricated and successfully irradiated in the HANARO CT and IR test holes since 1995. The capsules were mainly designed for an irradiation of the RPV (Reactor Pressure Vessel), reactor core materials, and Zr-based alloys. Most capsules were made for KAERI material research projects, but 5 capsules were made as a part of national projects for the promotion of the HANARO utilization for universities. Based on the accumulated irradiation experience and the user's sophisticated requirements, development of new instrumented capsule technologies for a more precise control of the irradiation temperature and fluence of a specimen irrespective of the reactor operation has been performed in HANARO. (author)

  9. Skin changes in `screen dermatitis` versus classical UV- and ionizing irradiation-related damage - similarities and differences

    Energy Technology Data Exchange (ETDEWEB)

    Gangi, S.; Johansson, O. [Karolinska Inst., Dept. of Neuroscience, Experimental Dermatology Unit, stockholm (Sweden)

    1997-12-01

    An increasing number of persons say that they get cutaneous problems as well as symptoms from certain internal organs, such as the central nervous system (CNS) and the heart, when being close to electric equipment. A major group of these patients are the users of video display terminals (VDTs), who claim to have subjective and objective skin- and mucosa-related symptoms, such as pain, itch, heat sensation, ery-therma, papules, and pustules. The CNS symptoms are, e.g. dizziness, tiredness, and headache. Erythema, itch, heat sensation, edema and pain are also common symptoms of sunburn (UV dermatitis). Alterations have been observed in cell populations of the skin of patients suffering from so-called `screen dermatitis` similar to those observed in the skin damaged due to ultraviolet (UV) light or ionizing radiation. In `screen dermatitis` patients a much higher number of mast cells have been observed. It is known that UVE irradiation induces mast cell degranulation and release of TNF-{alpha}. The high number of mast cells present in the `screen dermatitis` patients and the possible release of specific substances, such as histamine, may explain their clinical symptoms of itch, pain, edema and erythema. The most remarkable change among cutaneous cells, after exposure with the above-mentioned irradiation sources, is the disappearance of the Langerhans` cells. This change has also been observed in `screen dermatitis` patients, again pointing to a common cellular and molecular basis. The results of this literature study demonstrate that highly similar changes exist in the skin of `screen dermatitis` patients, as regard the clinical manifestations as well as alterations in the cell populations, and in skin damaged by UV light or ionizing radiation. (au) 93 refs.

  10. Skin changes in 'screen dermatitis' versus classical UV- and ionizing irradiation-related damage - similarities and differences

    International Nuclear Information System (INIS)

    Gangi, S.; Johansson, O.

    1997-01-01

    An increasing number of persons say that they get cutaneous problems as well as symptoms from certain internal organs, such as the central nervous system (CNS) and the heart, when being close to electric equipment. A major group of these patients are the users of video display terminals (VDTs), who claim to have subjective and objective skin- and mucosa-related symptoms, such as pain, itch, heat sensation, ery-therma, papules, and pustules. The CNS symptoms are, e.g. dizziness, tiredness, and headache. Erythema, itch, heat sensation, edema and pain are also common symptoms of sunburn (UV dermatitis). Alterations have been observed in cell populations of the skin of patients suffering from so-called 'screen dermatitis' similar to those observed in the skin damaged due to ultraviolet (UV) light or ionizing radiation. In 'screen dermatitis' patients a much higher number of mast cells have been observed. It is known that UVE irradiation induces mast cell degranulation and release of TNF-α. The high number of mast cells present in the 'screen dermatitis' patients and the possible release of specific substances, such as histamine, may explain their clinical symptoms of itch, pain, edema and erythema. The most remarkable change among cutaneous cells, after exposure with the above-mentioned irradiation sources, is the disappearance of the Langerhans' cells. This change has also been observed in 'screen dermatitis' patients, again pointing to a common cellular and molecular basis. The results of this literature study demonstrate that highly similar changes exist in the skin of 'screen dermatitis' patients, as regard the clinical manifestations as well as alterations in the cell populations, and in skin damaged by UV light or ionizing radiation. (au)

  11. Comparison of swelling for structural materials on neutron and ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Loomis, B.A.

    1986-03-01

    The swelling of V-base alloys, Type 316 stainless steel, Fe-25Ni-15Cr alloys, ferritic steels, Cu, Ni, Nb-1% Zr, and Mo on neutron irradiation is compared with the swelling for these materials on ion irradiation. The results of this comparison show that utilization of the ion-irradiation technique provides for a discriminative assessment of the potential for swelling of candidate materials for fusion reactors.

  12. Analysis of the irradiation data for A302B and A533B correlation monitor materials

    International Nuclear Information System (INIS)

    Wang, J.A.

    1996-04-01

    The results of Charpy V-notch impact tests for A302B and A533B-1 Correlation Monitor Materials (CMM) listed in the surveillance power reactor data base (PR-EDB) and material test reactor data base (TR-EDB) are analyzed. The shift of the transition temperature at 30 ft-lb (T 30 ) is considered as the primary measure of radiation embrittlement in this report. The hyperbolic tangent fitting model and uncertainty of the fitting parameters for Charpy impact tests are presented in this report. For the surveillance CMM data, the transition temperature shifts at 30 ft-lb (ΔT 30 ) generally follow the predictions provided by Revision 2 of Regulatory Guide 1.99 (R.G. 1.99). Difference in capsule temperatures is a likely explanation for large deviations from R.G. 1.99 predictions. Deviations from the R.G. 1.99 predictions are correlated to similar deviations for the accompanying materials in the same capsules, but large random fluctuations prevent precise quantitative determination. Significant scatter is noted in the surveillance data, some of which may be attributed to variations from one specimen set to another, or inherent in Charpy V-notch testing. The major contributions to the uncertainty of the R.G. 1.99 prediction model, and the overall data scatter are from mechanical test results, chemical analysis, irradiation environments, fluence evaluation, and inhomogeneous material properties. Thus in order to improve the prediction model, control of the above-mentioned error sources needs to be improved. In general the embrittlement behavior of both the A302B and A533B-1 plate materials is similar. There is evidence for a fluence-rate effect in the CMM data irradiated in test reactors; thus its implication on power reactor surveillance programs deserves special attention

  13. Materials of 15. autumn school on irradiated food

    International Nuclear Information System (INIS)

    1994-01-01

    The ionizing radiation use for food preservation has been shown on the background of other methods. Several aspects connected with food irradiation have been discussed. Among them the legal aspects and recommendations have been performed. The healthy aspects from the view point of the radiolysis of main components of irradiated food have been presented. The broad review of physical, chemical and biological methods for identification of irradiated food products has been done. The accelerator pilot plant for food irradiation working at the Institute of Nuclear Chemistry and Technology, Warsaw, has been presented as well

  14. Anomaly Detection in Nanofibrous Materials by CNN-Based Self-Similarity

    Directory of Open Access Journals (Sweden)

    Paolo Napoletano

    2018-01-01

    Full Text Available Automatic detection and localization of anomalies in nanofibrous materials help to reduce the cost of the production process and the time of the post-production visual inspection process. Amongst all the monitoring methods, those exploiting Scanning Electron Microscope (SEM imaging are the most effective. In this paper, we propose a region-based method for the detection and localization of anomalies in SEM images, based on Convolutional Neural Networks (CNNs and self-similarity. The method evaluates the degree of abnormality of each subregion of an image under consideration by computing a CNN-based visual similarity with respect to a dictionary of anomaly-free subregions belonging to a training set. The proposed method outperforms the state of the art.

  15. Similarity and symmetry methods applications in elasticity and mechanics of materials

    CERN Document Server

    Mladenov, Ivaïlo

    2014-01-01

    The principle aim of the book is to present a self-contained, modern account of similarity and symmetry methods, which are important mathematical tools for both physicists, engineers and applied mathematicians. The idea is to provide a balanced presentation of the mathematical techniques and applications of symmetry methods in mathematics, physics and engineering. That is why it includes recent developments and many examples in finding systematically conservation laws, local and nonlocal symmetries for ordinary and partial differential equations. The role of continuous symmetries in classical and quantum field theories is exposed at a technical level accessible even for non specialists. The importance of symmetries in continuum mechanics and mechanics of materials is highlighted through recent developments, such as the construction of constitutive models for various materials combining Lie symmetries with experimental data. As a whole this book is a unique collection of contributions from experts in the field...

  16. Resistivity measurements on the neutron irradiated detector grade silicon materials

    Energy Technology Data Exchange (ETDEWEB)

    Li, Zheng

    1993-11-01

    Resistivity measurements under the condition of no or low electrical field (electrical neutral bulk or ENB condition) have been made on various device configurations on detector grade silicon materials after neutron irradiation. Results of the measurements have shown that the ENB resistivity increases with neutron fluence ({Phi}{sub n}) at low {phi}{sub n} (<10{sup 13} n/cm{sup 2}) and saturates at a value between 300 and 400 k{Omega}-cm at {phi}{sub n} {approximately}10{sup 13} n/cm{sup 2}. Meanwhile, the effective doping concentration N{sub eff} in the space charge region (SCR) obtained from the C-V measurements of fully depleted p{sup +}/n silicon junction detectors has been found to increase nearly linearly with {phi}{sub n} at high fluences ({phi}{sub n} > 10{sup 13} n/cm{sup 2}). The experimental results are explained by the deep levels crossing the Fermi level in the SCR and near perfect compensation in the ENB by all deep levels, resulting in N{sub eff} (SCR) {ne} n or p (free carrier concentrations in the ENB).

  17. Hazard evaluation of The International Fusion Materials Irradiation Facility

    Energy Technology Data Exchange (ETDEWEB)

    Burgazzi, Luciano [ENEA-Centro Ricerche ' Ezio Clementel' , Advanced Physics Technology Division, Via Martiri di Monte Sole, 4, 40129 Bologna (Italy)]. E-mail: burgazzi@bologna.enea.it

    2005-01-15

    The International Fusion Materials Irradiation Facility (IFMIF) is aimed to provide an intense neutron source by a high current deuteron linear accelerator and a high-speed lithium flow target, for testing candidate materials for fusion. Liquid lithium is being circulated through a loop and is kept at a temperature above its freezing point. In the frame of the design phase called Key Element technology Phase (KEP), jointly performed by an international team to verify the most important risk factors, safety assessment of the whole plant has been required in order to identify the hazards associated with the plant operation. This paper discusses the safety assessments that were performed and their outcome: Failure Mode and Effect Analysis (FMEA) approach has been adopted in order to accomplish the task. Main conclusions of the study is that, on account of the safety and preventive measures adopted, potential plant related hazards are confined within the IFMIF security boundaries and great care must be exercised to protect workers and site personnel from operating the plant. The analysis has provided as a result a set of Postulated Initiating Events (PIEs), that is off-normal events, that could result in hazardous consequences for the plant, together with the total frequency and the list of component failures which could induce the PIE: this assures the exhaustive list of major initiating events of accident sequences, helpful to the further accident sequence analysis phase. Finally, for each one of the individuated PIEs, the evaluation of the accident evolution, in terms of effects on the plant and relative countermeasures, has allowed to verify that adequate measures are being taken both to prevent the accident occurrence and to cope with the accident consequences, thus assuring the fulfilment of the safety requirements.

  18. Neutron irradiation effects on superconducting and stabilizing materials for fusion magnets

    International Nuclear Information System (INIS)

    Maurer, W.

    1984-05-01

    Available low-temperature neutron irradiation data for the superconductors NbTi and Nb 3 Sn and the stabilization materials Cu and Al are collected and maximum tolerable doses for these materials are defined. A neutron flux in a reactor of about 10 9 n/cm 2 s at the magnet position is expected. However, in fusion experiments the flux can be higher by an order of magnitude or more. The energy spectrum is similar to a fission reactor. A fluence of about 10 18 n/cm 2 results during the lifetime of a fusion magnet (about 20 full power years). At this fluence and energy spectrum no severe degradation of the superconducting properties of NbTi and Nb 3 Sn will occur. But the radiation-induced resistivity is for Cu about a twentieth of the room temperature resistivity and a tenth for Al. (orig.) [de

  19. Report on the program of 4 K irradiation of insulating materials for the Superconducting Super Collider

    International Nuclear Information System (INIS)

    Spindel, A.

    1993-07-01

    This report is intended to serve as an aid to material selection. The results reported herein are the product of a careful investigation and can be used with confidence in their validity. The selection of materials based on this data, however, is not the responsibility of the author. This report will not approve or disapprove any specific material for use in the Super Collider. The author of this report does not assume any design responsibility or responsibility for material selection for any application. It is, therefore, very important that those with design responsibility use this report wisely. For this reason, the following informational guide to the material selection process has been provided. There are several issues to take into account when evaluating a material for radiation resistance. It is very important that the design criteria and operating loads for the application be known. For many applications the actual loading, and therefore required properties, are unknown. Certain materials have empirically been used successfully in a similar application and those materials have often been selected on that basis. Both percent degradation and the magnitude of the actual properties after irradiation need to be considered. Consider the scenario where two materials are being compared that both have acceptable properties after exposure to 10 9 rads. It is preferable to choose the material with less degradation because degradation tends to be a threshold phenomena with properties declining rapidly with dose after a certain threshold dose. The properties of the initially strong material, therefore, will be extremely sensitive to dose in that dose range and slight magnet-to-magnet differences in dose may, depending on the application, lead to performance variations

  20. 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

  1. Crack growth and development of fracture zones in plain concrete and similar materials

    International Nuclear Information System (INIS)

    Petersson, P.-E.

    1981-12-01

    A calculation model (the Fictitious Crack Model), based on fracture mechanics and the finite element method, is presented. In the model the fracture zone in front of a crack is represented by a fictitious crack that is able to transfer stress. The stress transferring capability of the fictitious crack normally decreases when the crack width increases. The applicability of linear elastic fracture mechanics to concrete and similar materials is analysed by use of the Fictitious Crack Model. The complete tensile stress-strain curve is introduced as a fracture mechanical parameter. The curve can be approximately determined if the tensile strength, the Young's modulus and the fracture energy are known. Suitable test methods for determining these properties are presented and test results are reported for a number of concrete qualities. A new type of very stiff tensile testing machine is presented by which it is possible to carry out stable tensile tests on concrete. The complete tensile stress-strain curves have been determined for a number of concrete qualities. A complete system for analysing crack propagation in concrete is covered, as a realistic material model, a functional calculation model and methods for determining the material properties necessary for the calculations are included. (Auth.)

  2. Irradiation of structural materials in contact with lead bismuth eutectic in the high flux reactor

    Energy Technology Data Exchange (ETDEWEB)

    Magielsen, A.J., E-mail: magielsen@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Jong, M.; Bakker, T.; Luzginova, N.V.; Mutnuru, R.K.; Ketema, D.J.; Fedorov, A.V. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands)

    2011-08-31

    In the framework of the materials domain DEMETRA in the European Transmutation research and development project EUROTRANS, irradiation experiment IBIS has been performed in the High Flux Reactor in Petten. The objective was to investigate the synergystic effects of irradiation and lead bismuth eutectic exposure on the mechanical properties of structural materials and welds. In this experiment ferritic martensitic 9 Cr steel, austenitic 316L stainless steel and their welds have been irradiated for 250 Full Power Days up to a dose level of 2 dpa. Irradiation temperatures have been kept constant at 300 deg. C and 500 deg. C. During the post-irradiation test phase, tensile tests performed on the specimens irradiated at 300 deg. C have shown that the irradiation hardening of ferritic martensitic 9 Cr steel at 1.3 dpa is 254 MPa, which is in line with the irradiation hardening obtained for ferritic martensitic Eurofer97 steel investigated in the fusion program. This result indicates that no LBE interaction at this irradiation temperature is present. A visual inspection is performed on the specimens irradiated in contact with LBE at 500 deg. C and have shown blackening on the surface of the specimens and remains of LBE that makes a special cleaning procedure necessary before post-irradiation mechanical testing.

  3. Effects of material property changes on irradiation assisted stress corrosion cracking

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, Morihito; Fukuya, Koji; Fujii, Katsuhiko [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2002-09-01

    Irradiation assisted stress corrosion cracking (IASCC) susceptibility and radiation-induced material changes in microstructure and microchemistry under pressurized water reactor (PWR) environment were examined on irradiated stainless steels (SSs), post-irradiation annealed SSs and post-irradiation deformed SS. The yield stress and grain boundary segregation were considerably high in SSs highly irradiated to 1-8 x 10{sup 26}n/m{sup 2} (E > 0.1 MeV) in PWR at 290-320degC, resulting in a high IASCC susceptibility. Following post-irradiation annealing of highly irradiated SSs, IASCC susceptibility showed significant recovery from 89% (as-irradiated) to 8% (550degC) of %IGSCC, while the hardness recovered from Hv375 (400degC) to Hv315 (550degC). Apparent recovery of segregation at grain boundaries was not observed. The SSs irradiated to 5.3 x 10{sup 24}n/m{sup 2} (E>1MeV) in the Japan Materials Testing Reactor (JMTR) at < 400degC, which had grain boundary segregation and low hardness, showed no IASCC susceptibility. Due to post-irradiation deforming for JMTR irradiated SS, the hardness increased but IASCC did not occur. These results suggested that the hardening would be a key factor for IASCC initiation under PWR hydrogenated water and that a yield stress threshold for IASCC initiation under slow strain rate tensile (SSRT) testing would the about 600MPa. (author)

  4. Effects of material property changes on irradiation assisted stress corrosion cracking

    International Nuclear Information System (INIS)

    Nakano, Morihito; Fukuya, Koji; Fujii, Katsuhiko

    2002-01-01

    Irradiation assisted stress corrosion cracking (IASCC) susceptibility and radiation-induced material changes in microstructure and microchemistry under pressurized water reactor (PWR) environment were examined on irradiated stainless steels (SSs), post-irradiation annealed SSs and post-irradiation deformed SS. The yield stress and grain boundary segregation were considerably high in SSs highly irradiated to 1-8 x 10 26 n/m 2 (E > 0.1 MeV) in PWR at 290-320degC, resulting in a high IASCC susceptibility. Following post-irradiation annealing of highly irradiated SSs, IASCC susceptibility showed significant recovery from 89% (as-irradiated) to 8% (550degC) of %IGSCC, while the hardness recovered from Hv375 (400degC) to Hv315 (550degC). Apparent recovery of segregation at grain boundaries was not observed. The SSs irradiated to 5.3 x 10 24 n/m 2 (E>1MeV) in the Japan Materials Testing Reactor (JMTR) at < 400degC, which had grain boundary segregation and low hardness, showed no IASCC susceptibility. Due to post-irradiation deforming for JMTR irradiated SS, the hardness increased but IASCC did not occur. These results suggested that the hardening would be a key factor for IASCC initiation under PWR hydrogenated water and that a yield stress threshold for IASCC initiation under slow strain rate tensile (SSRT) testing would the about 600MPa. (author)

  5. Induction of materials for mutation breeding of strawberry (Fragaria × Ananassa) by gamma irradiation

    International Nuclear Information System (INIS)

    Le Ngoc Trieu; Nguyen Tuong Mien; Le Tien Thanh; Huynh Thi Trung; Pham Van Nhi; Vu Thi Trac

    2015-01-01

    From collected New Zealand strawberry runners, micropropagation was executed to establish 500 shoot clusters for investigation effect of Gamma ray irradiation doses on survival rate. LD_5_0 at 52 Gy was recorded 45 days after re-injection and used as base for choosing 5 irradiation doses of 20, 40, 60, 80, 100 Gy for creation potentially existent mutant materials. 30 shoot clusters were irradiated at each chosen dose. Irradiated material was propagated by in vitro techniques to achieve 300 plantlets/chosen dose. There was no recorded alteration in survival rate and other morphological characteristics of irradiated materials compared to the control in nursery period. These materials were transplanted to plastic greenhouse to screen the mutant. (author)

  6. Development of a Low Temperature Irradiation Capsule for Research Reactor Materials

    International Nuclear Information System (INIS)

    Choo, Kee Nam; Cho, Man Soon; Lee, Cheol Yong; Yang, Sung Woo; Shin, Yoon Taek; Park, Seng Jae; Kang, Suk Hoon; Kang, Young Hwan; Park, Sang Jun

    2013-01-01

    A new capsule design was prepared and tested at HANARO for a neutron irradiation of core materials of research reactors as a part of the research reactor development project. Irradiation testing of the materials including graphite, beryllium, and zircaloy-4 that are supposed to be used as core materials in research reactors was required for irradiation at up to 8 reactor operation cycles at low temperature (<100 .deg. C). Therefore, three instrumented capsules were designed and fabricated for an evaluation of the neutron irradiation properties of the core materials (Graphite, Be, Zircaloy-4) of research reactors. The capsules were first designed and fabricated to irradiate materials at low temperature (<100 .deg. C) for a long cycle of 8 irradiation cycles at HANARO. Therefore, the safety of the new designed capsule should be fully checked before irradiation testing. Out-pile performance and endurance testing before HANARO irradiation testing was performed using a capsule under a 110% condition of a reactor coolant flow amount. The structural integrity of the capsule was analyzed in terms of a vibration-induced fatigue cracking of a rod tip of the capsule that is suspected to be the most vulnerable part of a capsule. Another two capsules were irradiated at HANARO for 4 cycles, and one capsule was transferred to a hot cell to examine the integrity of the rod tip of the capsule. After confirming the soundness of the 4 cycle-irradiated capsule, the remaining capsule was irradiated at up to 8 cycles at HANARO. Based on the structural integrity analysis of the capsule, an improved capsule design will be suggested for a longer irradiation test at HANARO

  7. Irradiation studies of mallard duck eggs material containing Mirex

    International Nuclear Information System (INIS)

    Lane, R.H.; Grodner, R.M.; Graves, J.L.

    1976-01-01

    Eggs containing Mirex (dodecachloropentacyclo[5.3.0.0 2 , 6 .0 3 , 9 .0 4 , 8 ]decane) from mallard ducks (Anas platyrhychos l.), fed diets with the insecticide incorporated at levels of 1 and 100 ppM for 25 weeks, were subjected to ultraviolet (uv) and γ irradiation. Seven derivatives were obtained on photolysis and eight derivatives were obtained from γ irradiation. Irradiation products appeared to be mono and dihydro derivatives of Mirex. Structural assignments for two monohydro derivatives and three dihydro derivatives were made on the basis of retention time and mass spectral data

  8. Low temperature gamma-ray irradiation effects on polymer materials

    International Nuclear Information System (INIS)

    Kudoh, Hisaaki; Kasai, Noboru; Sasuga, Tsuneo; Seguchi, Tadao

    1995-01-01

    The gamma radiation induced degradation of glass fiber reinforced plastic (GFRP) and polymethylmethacrylate (PMMA) at 77K was examined by flexural test and gas analysis after irradiation and compared by the irradiation at room temperature. The decrease in flexural strength at break was much less at 77K than at RT. The evolution of CH 4 , CO and CO 2 was also depressed at 77K. The temperature dependence of the degradation closely relates to the local molecular motion of matrix resin during irradiation. Polytetrafluoroethylene (PTFE) was also studied by irradiation at RT, 77K and 4K in terms of tensile elongation and molecular weight. The degradation was much less at 77K and 4K than at RT, and the same between 77K and 4K. (author)

  9. Heavy-ion irradiation induced diamond formation in carbonaceous materials

    International Nuclear Information System (INIS)

    Daulton, T. L.

    1999-01-01

    The basic mechanisms of metastable phase formation produced under highly non-equilibrium thermodynamic conditions within high-energy particle tracks are investigated. In particular, the possible formation of diamond by heavy-ion irradiation of graphite at ambient temperature is examined. This work was motivated, in part, by earlier studies which discovered nanometer-grain polycrystalline diamond aggregates of submicron-size in uranium-rich carbonaceous mineral assemblages of Precambrian age. It was proposed that the radioactive decay of uranium formed diamond in the fission particle tracks produced in the carbonaceous minerals. To test the hypothesis that nanodiamonds can form by ion irradiation, fine-grain polycrystalline graphite sheets were irradiated with 400 MeV Kr ions. The ion irradiated graphite (and unirradiated graphite control) were then subjected to acid dissolution treatments to remove the graphite and isolate any diamonds that were produced. The acid residues were then characterized by analytical and high-resolution transmission electron microscopy. The acid residues of the ion-irradiated graphite were found to contain ppm concentrations of nanodiamonds, suggesting that ion irradiation of bulk graphite at ambient temperature can produce diamond

  10. Analysis of the radiolytic products on high-dose irradiated food and packing materials

    International Nuclear Information System (INIS)

    Kim, Kyong Su; Shim, Sung Lye; Chung, In Sun

    2010-04-01

    The aims of this study were to prepare the government approval for the extension of food irradiation item to food or its products, to promote the industrial application of radiation technology, and to apply basic data in policy for introduction of irradiation. The change of hydrocarbons by irradiation was evaluated for the detection of irradiated meat. The results showed that hydrocarbons were detected in all of irradiated samples, but these hydrocarbons were not detected in non-irradiated samples. There were no difference between vacuum and N 2 - packaging. According to fatty acid compounds and degradation pathway of beef and pork, it could be deliberated that a great amount of produced hydrocarbons such as 8-heptadenene and 1,7-hexadecadien were able to be used as identification factor of irradiated meat. Effects of γ-irradiation on the volatile organic compounds in agricultural products were determined by analyzing changes of volatile composition. The composition of volatile organic compounds were little changed, but few specific compounds induced by γ-irradiation were identified. The variations of concentration in irradiated samples identified in this study could be due to the radiation sensitivity of compounds with the dose used. Effects of γ-irradiation on the volatile compounds in packaging materials were determined by analyzing changes of volatile composition. In polyethylene and polypropylene, 1,3-DBB was identified only in irradiated samples. Levels of 1,3-DBB increased with increasing irradiation doses. These results suggest may be useful in evaluation of γ-irradiation effects on food packaging materials

  11. Analysis of the radiolytic products on high-dose irradiated food and packing materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyong Su; Shim, Sung Lye; Chung, In Sun [Chosun University, Gwangju (Korea, Republic of)

    2010-04-15

    The aims of this study were to prepare the government approval for the extension of food irradiation item to food or its products, to promote the industrial application of radiation technology, and to apply basic data in policy for introduction of irradiation. The change of hydrocarbons by irradiation was evaluated for the detection of irradiated meat. The results showed that hydrocarbons were detected in all of irradiated samples, but these hydrocarbons were not detected in non-irradiated samples. There were no difference between vacuum and N{sub 2}- packaging. According to fatty acid compounds and degradation pathway of beef and pork, it could be deliberated that a great amount of produced hydrocarbons such as 8-heptadenene and 1,7-hexadecadien were able to be used as identification factor of irradiated meat. Effects of {gamma}-irradiation on the volatile organic compounds in agricultural products were determined by analyzing changes of volatile composition. The composition of volatile organic compounds were little changed, but few specific compounds induced by {gamma}-irradiation were identified. The variations of concentration in irradiated samples identified in this study could be due to the radiation sensitivity of compounds with the dose used. Effects of {gamma}-irradiation on the volatile compounds in packaging materials were determined by analyzing changes of volatile composition. In polyethylene and polypropylene, 1,3-DBB was identified only in irradiated samples. Levels of 1,3-DBB increased with increasing irradiation doses. These results suggest may be useful in evaluation of {gamma}-irradiation effects on food packaging materials

  12. Chemical and physical change of packaging materials for food by γ-ray irradiation

    International Nuclear Information System (INIS)

    Kawamura, Yoko; Takeda, Yuiko; Yamada, Takashi

    1998-01-01

    Packaging materials for food made of polyethylene, polypropylene and polystyrene were irradiated with 60 Co γ-ray. Exposure was 10, 30 and 50 kGy at 5 kGy/h exposure rate. With irradiating, all packaging materials of polyethylene and polypropylene produced volatile substances, for example, aldehydes, ketones and alcohols, especially, large amount of acetic acid and acetone. These volatile compounds were not observed in the sample unirradiated and increased with increasing exposure. Accordingly, it is concluded that they were decomposition products depend on irradiation. Polypropylene products were much more easily decomposed than polyethylene one because much more kinds and amount of volatile products were formed. However, on polystyrene products, content of styrene and ethylbenzene, monomer of raw materials, were reduced by irradiation and small amount of volatile substances were formed. These results proved its resistance to irradiation. (S.Y.)

  13. A study on the proton irradiation effect of reactor materials using cyclotron

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Park, Jong Man; Park, Deuk Keun; Lee, Bong Sang; Oh, Jong Myung

    1993-02-01

    Understanding on radiation damage of important structural materials is important for safe operation and radiation damage evaluation of new reactor structural materials. This study was performed to simulate and evaluate 14 MeV neutron irradiation effects on mechanical properties of candidate structural materials (HT-9/SS316) of next generation reactors (FBR, Fusion) irradiated by Cyclotron(MC-50) using SP test technique. After qualification of SP test techniques from J IC and ε qf correlation, SP tests were performed to evaluate 16MeV proton irradiation effects on mechanical properties of irradiated and unirradiated HT-9/SS316 steels. Test results were evaluated for ε qf , energy and displacement up to failure and J IC change. In addition, damaged zone and dpa upon depth after irradiation were calculated using TRIM code and Doppler broadening line shapes were measured to evaluate defects for 15% cold worked HT-9 steel using PAS. (Author)

  14. Stock selection of high-dose-irradiation-resistant materials for filter press under high-dose irradiation operation

    International Nuclear Information System (INIS)

    Ishiyama, Shintaro; Minami, Mamoru; Hara, Kouji; Yamashita, Manabu

    2015-01-01

    In a volume reduction process for the decontamination of contained soil, the performance degradation of a filter press is expected owing to material deterioration under high-dose irradiation. Eleven-stock selection of candidate materials including polymers, fibers and rubbers for the filter press was conducted to achieve a high performance of volume reduction of contaminated soil and the following results were derived. Crude rubber and nylon were selected as prime candidates for packing, diaphragm and filter plate materials. Polyethylene was also selected as a prime candidate for the filter cloth material. (author)

  15. Joint research centre fusion materials irradiations in HFR: Present status and prospectives

    International Nuclear Information System (INIS)

    Casini, G.; Fenici, P.

    1989-01-01

    First a review is made of the Joint Research Centre experimental activity at HFR-Petten in the frame of the Fusion Technology and Safety Programme. The materials under investigation are: Cr-Ni Austenitic steels (316-L type) and Cr-Mn Austenitic steels (AMCR and FI type) as structural materials and Pb-17Li eutetic as tritium breeding material. The experiments on structural materials comprise: Sample irradiations with post-irradiation tensile tests (FRUST) Sample irradiations under constant load and post-irradiation strain measurement (TRIESTE) On-line creep tests (CRISP). The experiments on Pb-17Li breeder material regard sample irradiations to investigate tritium production and recovery as well as tritium permeation through blanket structures (LIBRETTO Experiment). Both irradiations on structural and breeding materials will be pursued up to the end of the current JRC-Multiannual Programme (1988-1991) and even further. In the last part of the paper expected developments of the testing programme at HFR are discussed. New areas of research should involve materials for divertor applications (NET/ITER) and advanced low activation composite materials for Commercial Power Reactors

  16. Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

    International Nuclear Information System (INIS)

    Poirier, J.; Dupouy, J.M.

    In this conference are presented papers dealing with swelling of metals and alloys, (and specially ferritic steels), structural evolution and stability under irradiation, modifications of mechanical properties, consequences on the behaviour of fuel elements and the optimization of materials selection, and irradiation creep [fr

  17. Neutron and gamma irradiation damage to organic materials.

    Energy Technology Data Exchange (ETDEWEB)

    White, Gregory Von, II; Bernstein, Robert

    2012-04-01

    This document discusses open literature reports which investigate the damage effects of neutron and gamma irradiation on polymers and/or epoxies - damage refers to reduced physical chemical, and electrical properties. Based on the literature, correlations are made for an SNL developed epoxy (Epon 828-1031/DDS) with an expected total fast-neutron fluence of {approx}10{sup 12} n/cm{sup 2} and a {gamma} dosage of {approx}500 Gy received over {approx}30 years at < 200 C. In short, there are no gamma and neutron irradiation concerns for Epon 828-1031/DDS. To enhance the fidelity of our hypotheses, in regards to radiation damage, we propose future work consisting of simultaneous thermal/irradiation (neutron and gamma) experiments that will help elucidate any damage concerns at these specified environmental conditions.

  18. Gamma irradiation of yellow and blue colorants in polystyrene packaging materials

    International Nuclear Information System (INIS)

    Komolprasert, V.; Diel, Todd; Sadler, G.

    2006-01-01

    The effect of 10- and 20-kGy gamma irradiation was studied on chromophtal yellow 2RLTS (Yellow 110-2, 3, 4, 5-tetrachloro-6-cyanobenzoic acid) and Irgalite Blue GBP (copper (II) phthalocyanine blue) colorants, which were added to polystyrene (PS) material used to package food prior to irradiation. Analytical results obtained suggest that irradiation did not generate any new chemicals in the PS polymer containing either yellow or blue colorant at a concentration of up to 1% (w/w). Both yellow and blue colorants are relatively stable to gamma irradiation

  19. Effect of ammonia and electron beam irradiation on lignocelulosic materials

    International Nuclear Information System (INIS)

    Mastro, N.L. del; Gennari, S.M.; Castagnet, A.C.G.

    1986-01-01

    Reports on some of the effects produced on sugarcane bagasse and eucaliptus wood saccharification by combining irradiation and NH 3 treatment. The samples irradiated at 10 5 Gy, 2x10 5 Gy and 5x10 5 Gy with an electron accelerator were treated with anhydrous gaseous ammonia. Cellulase complex from T. reesei was used for hydrolysis assays. Bromatological analysis and 'in vitro' digestibility tests were performed. The combination of EBI and ammonia treatments produced and increase in the saccharification yield, 'in vitro' digestibility and protein content for the two kinds of sample. (Author) [pt

  20. An investigation of high-temperature irradiation test program of new ceramic materials

    International Nuclear Information System (INIS)

    Ishino, Shiori; Terai, Takayuki; Oku, Tatsuo

    1999-08-01

    The Japan Atomic Energy Research Institute entrusted the Atomic Energy Society of Japan with an investigation into the trend of irradiation processing/damage research on new ceramic materials. The present report describes the result of the investigation, which was aimed at effective execution of irradiation programs using the High Temperature Engineering Test Reactor (HTTR) by examining preferential research subjects and their concrete research methods. Objects of the investigation were currently on-going preliminary tests of functional materials (high-temperature oxide superconductor and high-temperature semiconductor) and structural materials (carbon/carbon and SiC/SiC composite materials), together with newly proposed subjects of, e.g., radiation effects on ceramics-coated materials and super-plastic ceramic materials as well as microscopic computer simulation of deformation and fracture of ceramics. These works have revealed 1) the background of each research subject, 2) its objective and significance from viewpoints of science and engineering, 3) research methodology in stages from preliminary tests to real HTTR irradiation, and 4) concrete HTTR-irradiation methods which include main specifications of test specimens, irradiation facilities and post-irradiation examination facilities and apparatuses. The present efforts have constructed the important fundamentals in the new ceramic materials field for further planning and execution of the innovative basic research on high-temperature engineering. (author)

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

    International Nuclear Information System (INIS)

    Petrequin, P.

    1995-01-01

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

  2. Irradiation effect of the insulating materials for fusion superconducting magnets at cryogenic temperature

    Science.gov (United States)

    Kobayashi, Koji; Akiyama, Yoko; Nishijima, Shigehiro

    2017-09-01

    In ITER, superconducting magnets should be used in such severe environment as high fluence of fast neutron, cryogenic temperature and large electromagnetic forces. Insulating material is one of the most sensitive component to radiation. So radiation resistance on mechanical properties at cryogenic temperature are required for insulating material. The purpose of this study is to evaluate irradiation effect of insulating material at cryogenic temperature by gamma-ray irradiation. Firstly, glass fiber reinforced plastic (GFRP) and hybrid composite were prepared. After irradiation at room temperature (RT) or liquid nitrogen temperature (LNT, 77 K), interlaminar shear strength (ILSS) and glass-transition temperature (Tg) measurement were conducted. It was shown that insulating materials irradiated at room temperature were much degraded than those at cryogenic temperature.

  3. 16 CFR 1145.2 - Paint (and other similar surface-coating materials) containing lead; toys, children's articles...

    Science.gov (United States)

    2010-01-01

    ... materials) containing lead; toys, children's articles, and articles of furniture bearing such paint (or... materials) containing lead; toys, children's articles, and articles of furniture bearing such paint (or...) Paint and other similar surface-coating materials containing lead and toys, children's articles, and...

  4. Installation of remote-handling typed EBSD-OIM analyzer for heavy irradiated reactor materials

    International Nuclear Information System (INIS)

    Kato, Yoshiaki; Takada, Fumiki; Ohmi, Masao; Nakagawa, Tetsuya; Miwa, Yukio

    2008-06-01

    The remote-handling typed EBSD-OIM analyzer for heavy irradiated reactor materials was installed in the JMTR hot laboratory at the first time in the world. The analyzer is used to study on IASCC (irradiation assisted stress corrosion cracking) or IGSCC (inter granular stress corrosion cracking) in reactor materials. This report describes the measurement procedure, the measured results and the operating experiences on the analyzer in the JMTR hot laboratory. (author)

  5. High dose neutron irradiation damage in beryllium as blanket material

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V.P. E-mail: fae@niiar.ru; Kazakov, V.A.; Teykovtsev, A.A.; Pimenov, V.V.; Shimansky, G.A.; Ostrovsky, Z.E.; Suslov, D.N.; Latypov, R.N.; Belozerov, S.V.; Kupriyanov, I.B. E-mail: vniinm.400@g23.relkom.ru

    2001-11-01

    The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8x10{sup 22} and 8.0x10{sup 22} cm{sup -2} (E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70-150 deg. C and neutron fluence of 2.8x10{sup 22} cm{sup -2} (E>1 MeV) makes up 0.8-1.5%, at 400 deg. C and fluence of 8x10{sup 22} cm{sup -2} (E>1 MeV)-3.2-5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket.

  6. High dose neutron irradiation damage in beryllium as blanket material

    International Nuclear Information System (INIS)

    Chakin, V.P.; Kazakov, V.A.; Teykovtsev, A.A.; Pimenov, V.V.; Shimansky, G.A.; Ostrovsky, Z.E.; Suslov, D.N.; Latypov, R.N.; Belozerov, S.V.; Kupriyanov, I.B.

    2001-01-01

    The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8x10 22 and 8.0x10 22 cm -2 (E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70-150 deg. C and neutron fluence of 2.8x10 22 cm -2 (E>1 MeV) makes up 0.8-1.5%, at 400 deg. C and fluence of 8x10 22 cm -2 (E>1 MeV)-3.2-5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket

  7. Binary-collision-approximation-based simulation of noble gas irradiation to tungsten materials

    International Nuclear Information System (INIS)

    Saito, Seiki; Takayama, Arimichi; Ito, Atsushi M.; Nakamura, Hiroaki

    2013-01-01

    To reveal the possibility of fuzz formation of tungsten material under noble gas irradiation, helium, neon, and argon atom injections into tungsten materials are performed by binary-collision-approximation-based simulation. The penetration depth is strongly depends on the structure of the target material. Therefore, the penetration depth for amorphous and bcc crystalline structure is carefully investigated in this paper

  8. Investigation of high flux test module for the international fusion materials irradiation facilities (IFMIF)

    International Nuclear Information System (INIS)

    Miyashita, Makoto; Sugimoto, Masayoshi; Yutani, Toshiaki

    2007-03-01

    This report describes investigation on structure of a high neutron flux test module (HFTM) for the International Fusion Materials Irradiation Facilities (IFMIF). The HFTM is aimed for neutron irradiation of a specimen in a high neutron flux domain of the test cell for irradiation ground of IFMIF. We investigated the overall structure of the HFTM that was able to include specimens in a rig and thermocouple arrangement, an interface of control signal and support structure. Moreover, pressure and the amount of the bend in the module vessel (a rectangular section pressure vessel) were calculated. The module vessel did a rectangular section from limitation of a high neutron flux domain. Also, we investigated damage of thermocouples under neutron irradiation, which was a temperature sensor of irradiation materials temperature control demanded high precision. Based on these results, drawings on the HTFM structure. (author)

  9. Radiation-Induced Fluidity and Glass-Liquid Transition in Irradiated Amorphous Materials

    International Nuclear Information System (INIS)

    Ojovan, M.I.

    2009-01-01

    This paper describes the fluidity behaviour of continuously irradiated glasses using the Congruent Bond Lattice model in which broken bonds 'configurons' facilitate the flow. Irradiation breaks the bonds creating configurons which at high concentrations provide the transition of material from the glassy to liquid state. An explicit equation of viscosity has been derived which gives results in agreement with experimental data. This equation provides correct viscosity data for non-irradiated materials and shows a significant increase of fluidity in radiation fields. It demonstrates a decrease of activation energy of flow for irradiated glasses. A simple equation for glass-transition temperature was also obtained which shows that irradiated glasses have lower glass transition temperatures and are readily transformed from glassy to liquid state e.g. fluidized in strong radiation fields. (authors)

  10. Deformation behavior of irradiated Zr-2.5Nb pressure tube material

    International Nuclear Information System (INIS)

    Himbeault, D.D.; Chow, C.K.; Puls, M.P.

    1994-01-01

    A study of the deformation behavior of irradiated highly textured Zr-2.5Nb pressure tube material in the temperature range of 30 degree C to 300 degree C was undertaken to understand better the mechanism for the deterioration of the fracture toughness with neutron irradiation. Strain localization behavior, believed to be a main contributor to reduced toughness, was observed in irradiated transverse tensile specimens at temperature greater than 100 degree C. The strain localization behavior was found to occur by the cooperative twinning of the highly textured grains of the material, resulting in a local softening of the material, where the flow than localizes. It is believed that the effect of the irradiation is to favor twinning at the expense of slip in the early stages of deformation. This effect becomes more pronounced at higher temperature, thus leading to the high-temperature strain localization behavior of the material. A limited amount of dislocation channeling was also observed; however, it is not considered to have a major role in the strain localization behavior of the material. Contrary to previous reports on irradiated zirconium alloys, static strain aging is observed in the irradiated material in the temperature range of 150 degree C to 300 degree C

  11. Development of Multiscale Materials Modeling Techniques and Coarse- Graining Strategies for Predicting Materials Degradation in Extreme Irradiation Environments

    Energy Technology Data Exchange (ETDEWEB)

    Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States)

    2016-01-12

    Exposure of metallic structural materials to irradiation environments results in significant microstructural evolution, property changes and performance degradation, which limits the extended operation of current generation light water reactors and restricts the design of advanced fission and fusion reactors [1-8]. This effect of irradiation on materials microstructure and properties is a classic example of an inherently multiscale phenomenon, as schematically illustrated in Figure 1a. Pertinent processes range from the atomic nucleus to structural component length scales, spanning more than 15 orders of magnitude. Time scales bridge more than 22 orders of magnitude, with the shortest being less than a femtosecond [1,8]. Further, the mix of radiation-induced features formed and the corresponding property degradation depend on a wide range of material and irradiation variables. This emphasizes the importance of closely integrating models with high-resolution experimental characterization of the evolving radiation- damaged microstructure, including measurements performed in-situ during irradiation. In this article, we review some recent successes through the use of closely coordinated modeling and experimental studies of the defect cluster evolution in irradiated body-centered cubic materials, followed by a discussion of outstanding challenges still to be addressed, which are necessary for the development of comprehensive models of radiation effects in structural materials.

  12. An investigation of neutron irradiation test on superplastic zirconia-ceramic materials

    International Nuclear Information System (INIS)

    Shibata, Taiju; Ishihara, Masahiro; Baba, Shinichi; Hayashi, Kimio

    2000-05-01

    A neutron irradiation test on superplastic ceramic materials at high temperature has been proposed as an innovative basic research on high-temperature engineering using the High Temperature Engineering Test Reactor (HTTR). For the effective execution of the test, we reviewed the superplastic deformation mechanism of ceramic materials and discussed neutron irradiation effects on the superplastic deformation process of stabilized Tetragonal Zirconia Polycrystal (TZP), which is a representative superplastic ceramic material. As a result, we pointed out that the decrease in the activation energy for superplastic deformation is expected by the radiation-enhanced diffusion. We selected a fast neutron fluence of 5x10 20 n/cm 2 and an irradiation temperature of about 600degC as test conditions for the first irradiation test on TZP and decided to perform a preliminary irradiation test by the Japan Materials Testing Reactor (JMTR). Moreover, we estimated the radioactivity of irradiated TZP and indicated that it is in the order of 10 10 Bq/g (about 0.3 Ci/g) immediately after irradiation to a thermal neutron fluence of 3x10 20 n/cm 2 and that it decays to about 1/100 in a year. (author)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  14. The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

    International Nuclear Information System (INIS)

    Toivonen, A.; Moilanen, P.; Pyykkoenen, M.; Taehtinen, S.; Rintamaa, R.; Saario, T.

    1998-01-01

    Environmentally assisted cracking (EAC) of core materials has become an increasingly important issue of downtime and maintenance costs in nuclear power plants. Small size specimens are necessary in stress corrosion testing of irradiated materials because of difficulties in handling high dose rate materials and because of restricted availability of the materials. The drawback of using small size specimens is that in some cases they do not fulfil the requirements of the relevant testing standards. Recently VTT has developed J-R testing with irradiated and non-irradiated sub size 3 PB specimens, both in inert and in LWR environments. Also, a new materials testing system which will enable simultaneous multiple specimen testing both in laboratory conditions and in operating reactor core is under development. The new testing system will utilize Charpy and sub size 3 PB specimens. The feasibility study of the system has been carried out using different materials. Fracture resistance curves of a Cu-Zr-Cr alloy are shown to be independent of the specimen geometry and size, to some extent. Results gained from tests in simulated boiling water reactor (BWR) water are presented for sensitized SIS 2333 stainless steel. The experimental results indicate that the size of the plastic zone or stress triaxiality must be further studied although no significant effect on the environmentally assisted crack growth rate was observed. (orig.)

  15. The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Toivonen, A.; Moilanen, P.; Pyykkoenen, M.; Taehtinen, S.; Rintamaa, R.; Saario, T. [Valtion Teknillinen Tutkimuskeskus, Espoo (Finland)

    1998-11-01

    Environmentally assisted cracking (EAC) of core materials has become an increasingly important issue of downtime and maintenance costs in nuclear power plants. Small size specimens are necessary in stress corrosion testing of irradiated materials because of difficulties in handling high dose rate materials and because of restricted availability of the materials. The drawback of using small size specimens is that in some cases they do not fulfil the requirements of the relevant testing standards. Recently VTT has developed J-R testing with irradiated and non-irradiated sub size 3 PB specimens, both in inert and in LWR environments. Also, a new materials testing system which will enable simultaneous multiple specimen testing both in laboratory conditions and in operating reactor core is under development. The new testing system will utilize Charpy and sub size 3 PB specimens. The feasibility study of the system has been carried out using different materials. Fracture resistance curves of a Cu-Zr-Cr alloy are shown to be independent of the specimen geometry and size, to some extent. Results gained from tests in simulated boiling water reactor (BWR) water are presented for sensitized SIS 2333 stainless steel. The experimental results indicate that the size of the plastic zone or stress triaxiality must be further studied although no significant effect on the environmentally assisted crack growth rate was observed. (orig.)

  16. Verification of neutron irradiation on S/G tube materials

    International Nuclear Information System (INIS)

    Kang, Byoung Hwi; Lee, S. K.; Jang, D. Y.; Jo, K. H.

    2010-12-01

    The fluence monitors were fabricated with metal wires of the purity ≥ 99.9%, whose dimensions were 0.1mm diameter, about 3mm length, and around 150-200 μg mass range. Three wire samples (Fe, Ni, Ti) were prepared for one irradiation aluminum capsule. Five capsules were irradiated in the OR5 hole of the HANARO reactor at 30 MW power for about 25 days. The reaction rates were calculated by using the measured radiation activity data, and then neutron fluence were obtained from the reaction rates and the weighted neutron cross section with calculated neutron spectrum at the fluence monitor position. The measured neutron fluences were compared to the calculated ones. (Errors ≤ 35%)

  17. Materials irradiation subpanel report to BESAC neutron sources and research panel

    International Nuclear Information System (INIS)

    Birtcher, R.C.; Goland, A.N.; Lott, R.

    1992-01-01

    The future success of the nuclear power option in the US (fission and fusion) depends critically on the continued existence of a healthy national materials-irradiation program. Consideration of the requirements for acceptable materials-irradiation systems in a new neutron source has led the subcommittee to identify an advanced steady-state reactor (ANS) as a better choice than a spallation neutron source. However, the subcommittee also hastens to point out that the ANS cannot stand alone as the nation's sole high-flux mixed-spectrum neutron irradiation source in the next century. It must be incorporated in a broader program that includes other currently existing neutron irradiation facilities. Upgrading and continuing support for these facilities must be planned. In particular, serious consideration should be given to converting the HFIR into a dedicated materials test reactor, and long-term support for several university reactors should be established

  18. Transmission electron microscopy of oxide dispersion strengthened (ODS) molybdenum: effects of irradiation on material microstructure

    International Nuclear Information System (INIS)

    Baranwal, R.; Burke, M.G.

    2003-01-01

    Oxide dispersion strengthened (ODS) molybdenum has been characterized using transmission electron microscopy (TEM) to determine the effects of irradiation on material microstructure. This work describes the results-to-date from TEM characterization of unirradiated and irradiated ODS molybdenum. The general microstructure of the unirradiated material consists of fine molybdenum grains (< 5 (micro)m average grain size) with numerous low angle boundaries and isolated dislocation networks. 'Ribbon'-like lanthanum oxides are aligned along the working direction of the product form and are frequently associated with grain boundaries, serving to inhibit grain boundary and dislocation movement. In addition to the 'ribbons', discrete lanthanum oxide particles have also been detected. After irradiation, the material is characterized by the presence of nonuniformly distributed large (∼ 20 to 100 nm in diameter), multi-faceted voids, while the molybdenum grain size and oxide morphology appear to be unaffected by irradiation

  19. Effects of non-steady irradiation conditions on fusion materials performance

    International Nuclear Information System (INIS)

    Matsui, H.; Fukumoto, K.; Nagumo, T.; Nita, N.

    2001-01-01

    During startup of fusion reactors, materials are exposed to neutron irradiation under non-steady temperature condition. Since the temperature of irradiation has decisive effects on the microstructural evolution, the non-steady temperature will have important consequences in the performance of fusion reactor materials. In the present study, a series of vanadium based alloys have been irradiated with neutrons in a temperature cycling condition. It has been found from this study that cavity number density is much greater in temperature cycled specimens than in steady temperature irradiation. Keeping the upper temperature constant, cavity number density is greater for smaller difference between the upper and the lower temperature. It follows that relatively small temperature excursions may have rather significant effects on the fusion material performance in service. (author)

  20. Creep of fissile ceramic materials under neutron irradiation

    International Nuclear Information System (INIS)

    Brucklacher, D.

    1975-01-01

    Theoretical estimation of the irradiation-induced creep rate of U0 2 by a modification of the Nabarro-Herring model for diffusional creep resulted in a creep rate range between about 6 x 10 -6 to 8 x 10 -5 h -1 for a fission rate of 1 x 10 14 f/cm 3 s and a stress of 2 kgf/mm 2 . Accordingly, the creep rate is enhanced by irradiation at temperatures below 1000 0 to 1200 0 C. It is essentially due to the 'thermal rods' along the fission fragment tracks. Therefore, irradiation-induced creep rates should depend only slightly on temperature and must be markedly lower for carbide and nitride fuel. In-reactor creep experiments on UO 2 were performed at fuel temperatures between 250 0 to 850 0 C. At burnups between 0.3 to 3% the steady-state compressive creep rates are proportional to stress (0 to 4 kgf/mm 2 ) and to fission rate (1 x 10 13 to 2 x 10 14 f/cm 3 s), and are in the range estimated before. The increase in the creep rate with increasing temperature is low and corresponds to an apparent activation energy of only 5200 cal/mol. At burnups above 3 to 4% the stress exponent of the irradiation-induced creep rate increased from n = 1 to n = 1.5. Creep measurements on UO 2 to 15 wt-%Pu0 2 (mechanically mixed, sintered density 86% TD) showed the same temperature dependence as UO 2 below 700 0 C. However, the creep rates were higher by a factor of about 20 compared to fully dense UO 2 . This difference may be explained by assuming a high 'effective' porosity. In-pile creep tests on some UN samples resulted in creep rates that were lower by an order of magnitude than for UO 2 under comparable conditions. (author)

  1. Identification of spectrally similar materials using the USGS Tetracorder algorithm: The calcite-epidote-chlorite problem

    Science.gov (United States)

    Dalton, J.B.; Bove, D.J.; Mladinich, C.S.; Rockwell, B.W.

    2004-01-01

    A scheme to discriminate and identify materials having overlapping spectral absorption features has been developed and tested based on the U.S. Geological Survey (USGS) Tetracorder system. The scheme has been applied to remotely sensed imaging spectroscopy data acquired by the Airborne Visible and Infrared Imaging Spectrometer (AVIRIS) instrument. This approach was used to identify the minerals calcite, epidote, and chlorite in the upper Animas River watershed, Colorado. The study was motivated by the need to characterize the distribution of calcite in the watershed and assess its acid-neutralizing potential with regard to acidic mine drainage. Identification of these three minerals is difficult because their diagnostic spectral features are all centered at 2.3 ??m, and have similar shapes and widths. Previous studies overestimated calcite abundance as a result of these spectral overlaps. The use of a reference library containing synthetic mixtures of the three minerals in varying proportions was found to simplify the task of identifying these minerals when used in conjunction with a rule-based expert system. Some inaccuracies in the mineral distribution maps remain, however, due to the influence of a fourth spectral component, sericite, which exhibits spectral absorption features at 2.2 and 2.4 ??m that overlap the 2.3-??m absorption features of the other three minerals. Whereas the endmember minerals calcite, epidote, chlorite, and sericite can be identified by the method presented here, discrepancies occur in areas where all four occur together as intimate mixtures. It is expected that future work will be able to reduce these discrepancies by including reference mixtures containing sericite. ?? 2004 Elsevier Inc. All rights reserved.

  2. Investigation of structural materials of reactors using high-energy heavy-ion irradiations

    International Nuclear Information System (INIS)

    Wang Zhiguang

    2007-01-01

    Radiation damage in structural materials of fission/fusion reactors is mainly attributed to the evolution of intensive atom displacement damage induced by energetic particles (n, α and/or fission fragments) and high-rate helium doping by direct α particle bombardments and/or (n, α) reactions. It can cause severe degradation of reactor structural materials such as surface blistering, bulk void swelling, deformation, fatigue, embrittlement, stress erosion corrosion and so on that will significantly affect the operation safety of reactors. However, up to now, behavior of structural materials at the end of their service can hardly be fully tested in a real reactor. In the present work, damage process in reactor structural materials is briefly introduced, then the advantages of energetic ion implantation/irradiation especially high-energy heavy ion irradiation are discussed, and several typical examples on simulation of radiation effects in reactor candidate structural materials using high-energy heavy ion irradiations are pronounced. Experimental results and theoretical analysis suggested that irradiation with energetic particles especially high-energy heavy ions is very useful technique for simulating the evolution of microstructures and macro-properties of reactor structural materials. Furthermore, an on-going plan of material irradiation experiments using high energy H- and He-ions based on the Heavy Ion Research Facilities in Lanzhou (HIRFL) is also briefly interpreted. (authors)

  3. Diffusion and solubility of oxygen in γ-ray irradiated polymer insulation materials

    International Nuclear Information System (INIS)

    Seguchi, Tadao; Yamamoto, Yasuaki.

    1986-03-01

    The effects of 60 Co γ-rays irradiation on diffusion and solubility of oxygen in polymer materials for electric cable insulation materials were investigated. The polymers were polyethylene, ethylene-propylene rubber, chlorinated polyethylene, chlorosulphonated polyethylene, and chloroprene rubber. They were pure grade and several types of formulation grade. The sheets of these polymers were irradiated up to 5 - 200 Mrad under vacuum or in oxygen under pressure of 3 - 15 atm at room temperature or at 70 deg C. By a method of gas desorption, the diffusion coefficient (D) and solubility coefficient (S) of oxygen or argon in polymer materials were determined at various temperatures of 10 - 80 deg C. The D and S decreased with increase of dose, and the decrease by irradiation with oxidation was more remarkable than that by irradiation without oxidation. However, the decreases of D and S by irradiation were reduced by the formulation of polymers. The additives in formulated polymers would reduce the reactions of crosslinking or oxidation by γ-ray irradiation. The activation energy of D was scarcely changed by irradiations with and without oxidation. (author)

  4. Complete Report on the Development of Welding Parameters for Irradiated Materials

    Energy Technology Data Exchange (ETDEWEB)

    Frederick, Greg [Electric Power Research Inst. (EPRI), Knoxville, TN (United States); Sutton, Benjamin J. [Electric Power Research Inst. (EPRI), Knoxville, TN (United States); Tatman, Jonathan K. [Electric Power Research Inst. (EPRI), Knoxville, TN (United States); Vance, Mark Christopher [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Clark, Scarlett R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Feng, Zhili [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Roger G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chen, Jian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Tang, Wei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gibson, Brian T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-11-01

    The advanced welding facility at the Radiochemical Engineering Development Center of Oak Ridge National Laboratory, which was conceived to enable research and development of weld repair techniques for nuclear power plant life extension, is now operational. The development of the facility and its advanced welding capabilities, along with the model materials for initial welding trials, were funded jointly by the U.S. Department of Energy, Office of Nuclear Energy, Light Water Reactor Sustainability Program, the Electric Power Research Institute, Long Term Operations Program and the Welding and Repair Technology Center, with additional support from Oak Ridge National Laboratory. Welding of irradiated materials was initiated on November 17, 2017, which marked a significant step in the development of the facility and the beginning of extensive welding research and development campaigns on irradiated materials that will eventually produce validated techniques and guidelines for weld repair activities carried out to extend the operational lifetimes of nuclear power plants beyond 60 years. This report summarizes the final steps that were required to complete weld process development, initial irradiated materials welding activities, near-term plans for irradiated materials welding, and plans for post-weld analyses that will be carried out to assess the ability of the advanced welding processes to make repairs on irradiated materials.

  5. Neutronics analysis of International Fusion Material Irradiation Facility (IFMIF). Japanese contributions

    International Nuclear Information System (INIS)

    Oyama, Yukio; Noda, Kenji; Kosako, Kazuaki.

    1997-10-01

    In fusion reactor development for demonstration reactor, i.e., DEMO, materials tolerable for D-T neutron irradiation are absolutely required for both mechanical and safety point of views. For this requirement, several kinds of low activation materials were proposed. However, experimental data by actual D-T fusion neutron irradiation have not existed so far because of lack of fusion neutron irradiation facility, except fundamental radiation damage studies at very low neutron fluence. Therefore such a facility has been strongly requested. According to agreement of need for such a facility among the international parties, a conceptual design activity (CDA) of International Fusion Material Irradiation Facility (IFMIF) has been carried out under the frame work of the IEA-Implementing Agreement. In the activity, a neutronics analysis on irradiation field optimization in the IFMIF test cell was performed in three parties, Japan, US and EU. As the Japanese contribution, the present paper describes a neutron source term as well as incident deuteron beam angle optimization of two beam geometry, beam shape (foot print) optimization, and dpa, gas production and heating estimation inside various material loading Module, including a sensitivity analysis of source term uncertainty to the estimated irradiation parameters. (author)

  6. Learning material recommendation based on case-based reasoning similarity scores

    Science.gov (United States)

    Masood, Mona; Mokmin, Nur Azlina Mohamed

    2017-10-01

    A personalized learning material recommendation is important in any Intelligent Tutoring System (ITS). Case-based Reasoning (CBR) is an Artificial Intelligent Algorithm that has been widely used in the development of ITS applications. This study has developed an ITS application that applied the CBR algorithm in the development process. The application has the ability to recommend the most suitable learning material to the specific student based on information in the student profile. In order to test the ability of the application in recommending learning material, two versions of the application were created. The first version displayed the most suitable learning material and the second version displayed the least preferable learning material. The results show the application has successfully assigned the students to the most suitable learning material.

  7. 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

  8. Effect of Fast Neutron Irradiation on Current Transport Properties of HTS Materials

    CERN Document Server

    Ballarino, A; Kruglov, V S; Latushkin, S T; Lubimov, A N; Ryazanov, A I; Shavkin, S V; Taylor, T M; Volkov, P V

    2004-01-01

    The effect of fast neutron irradiation with energy up to 35 MeV and integrated fluence of up to 5 x 10**15 cm-2 on the current transport properties of HTS materials Bi-2212 and Bi-2223 has been studied, both at liquid nitrogen and at room temperatures. The samples irradiated were selected after verification of the stability of their superconducting properties after temperature cycling in the range of 77 K - 293 K. It has been found that the irradiation by fast neutrons up to the above dose does not produce a significant degradation of critical current. The effect of room temperature annealing on the recovery of transport properties of the irradiated samples is also reported, as is a preliminary microstructure investigation of the effect of irradiation on the soldered contacts.

  9. Irradiation creep lifetime analysis on first wall structure materials for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Bing; Peng, Lei, E-mail: penglei@ustc.edu.cn; Zhang, Xiansheng; Shi, Jingyi; Zhan, Jie

    2017-05-15

    Fusion reactor first wall services on the conditions of high surface heat flux and intense neutron irradiation. For China Fusion Engineering Test Reactor (CFETR) with high duty time factor, it is important to analyze the irradiation effect on the creep lifetime of the main candidate structure materials for first wall, i.e. ferritic/martensitic steel, austenite steel and oxide dispersion strengthened steel. The allowable irradiation creep lifetime was evaluated with Larson-Miller Parameter (LMP) model and finite element method. The results show that the allowable irradiation creep lifetime decreases with increasing of surface heat flux, first wall thickness and inlet coolant temperature. For the current CFETR conceptual design, the lifetime is not limited by thermal creep or irradiation creep, which indicated the room for design parameters optimization.

  10. Visual and Phonological Similarity Effects in Verbal Immediate Serial Recall: A Test with Kanji Materials

    Science.gov (United States)

    Saito, Satoru; Logie, Robert H.; Morita, Aiko; Law, Anna

    2008-01-01

    In a series of three experiments, native speakers of Japanese performed serial ordered written recall of visually presented Japanese kanji characters that varied systematically in visual and phonological similarity. Overall effects of phonological similarity were observed for retention of serial order under silent reading in Experiments 1 and 3…

  11. A sharp interface model for void growth in irradiated materials

    Science.gov (United States)

    Hochrainer, Thomas; El-Azab, Anter

    2015-03-01

    A thermodynamic formalism for the interaction of point defects with free surfaces in single-component solids has been developed and applied to the problem of void growth by absorption of point defects in irradiated metals. This formalism consists of two parts, a detailed description of the dynamics of defects within the non-equilibrium thermodynamic frame, and the application of the second law of thermodynamics to provide closure relations for all kinetic equations. Enforcing the principle of non-negative entropy production showed that the description of the problem of void evolution under irradiation must include a relationship between the normal fluxes of defects into the void surface and the driving thermodynamic forces for the void surface motion; these thermodynamic forces are identified for both vacancies and interstitials and the relationships between these forces and the normal point defect fluxes are established using the concepts of transition state theory. The latter theory implies that the defect accommodation into the surface is a thermally activated process. Numerical examples are given to illustrate void growth dynamics in this new formalism and to investigate the effect of the surface energy barriers on void growth. Consequences for phase field models of void growth are discussed.

  12. Effects of irradiation temperature on polarisation and relaxation characteristics of polymeric materials

    Energy Technology Data Exchange (ETDEWEB)

    Bornstein, Marcel; Dutz, Hartmut; Goertz, Stefan; Reeve, Scott; Runkel, Stefan [Physikalisches Institut, Bonn Univ. (Germany)

    2016-07-01

    To achieve significant enhancement of polarisation of solid target materials one must use the principles of dynamic nuclear polarisation and utilise the coupling of the nuclear and electron spins. The unpaired electrons needed can be created as paramagnetic structural defects by irradiation of the material. Polyethylene and polypropylene materials were irradiated at various temperatures and subsequently polarised with microwaves of approximately 70 GHz at temperatures around 1 K. Additionally the samples were investigated with respect to the nature of the created paramagnetic defects using a X-band EPR spectrometer. It was found that the irradiation temperature has a significant effect on the polarisation values achieved and also on the relaxation times of the materials in the 2.5 T magnetic field. The EPR line shape is clearly dominated by the well known alkyl radical structure.

  13. Review Paper: Review of Instrumentation for Irradiation Testing of Nuclear Fuels and Materials

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Rempe, Joy L.; Villard, Jean-Francois; Solstadd, Steinar

    2011-01-01

    Over 50 years of nuclear fuels and materials irradiation testing has led to many countries developing significant improvements in instrumentation to monitor physical parameters and to control the test conditions in material test reactors (MTRs). Recently, there is increased interest to irradiate new materials and reactor fuels for advanced pressurized water reactors and Gen-IV reactor systems, such as sodium-cooled fast reactors, very high temperature reactors, supercritical water-cooled reactors, and gas-cooled fast reactors. This review paper documents the current state of instrumentation technologies in MTRs in the world and summarizes ongoing research efforts to deploy new sensors. As described in this paper, a wide range of sensors is available to measure key parameters of interest during fuels and materials irradiations in MTRs. Ongoing development efforts focus on providing MTR users a wider range of parameter measurements with smaller, higher accuracy sensors.

  14. Chemical and physical change of packaging materials for food by gamma-ray irradiation

    International Nuclear Information System (INIS)

    Kawamura, Yoko; Yamada, Takashi

    1999-01-01

    Recently, foods are often exposed to radiation under packed states with various wrapping materials. In this study, the effects of γ-ray radiation were investigated on the additives in wrapping materials on the market. 10 - 50 kGy γ radiation was irradiated to samples under sealed condition in a glass-ware. Polyethylene bag and wrapping film, polypropylene wrapping film, cup and sheet, and polystyrene cup were used as samples. And the additives in these materials were analyzed by GC/MS to evaluate the radiation effects on them. The irradiation was found to induce rapid degradation of antioxidants, especially, Irgafos 168. Some fatty acid amides used as a lubricant and a plasticizer, DBP were also reduced, but not aliphatic hydrocarbons. However, all polystyrene products used in this study included no additives. The contents of styrene dimer and trimer in those wrapping materials were not changed by γ-irradiation. (M.N.)

  15. Effect of γ-ray irradiation on properties of castor oil-polyurethane potting materials

    International Nuclear Information System (INIS)

    Guan Jian; Luo Xianglin; Yue Yilun

    2001-01-01

    After γ-ray sterilization, the amounts of 4,4'-methylenedianiline (MDA) in the five kinds of synthesized medical castor oil-polyurethane potting materials were detected by HPLC. The influences of γ-ray irradiation on the mechanical performance of the potting materials were also discussed quantitatively.The experimental results show that the amounts of produced MDA increases with γ-ray irradiation dosage. After 25 kGy γ-ray sterilization, the accumulated amounts of MDA in the five kinds of potting materials were 10.33, 10.37, 10.52, 10.59, 10.91 ? μg/g respectively. Those amounts are below the level of harm amount to human body. At the same time, the mechanical properties of the potting materials such as tensile strength, tear strength and hardness are improved because cross-linking happens under irradiation

  16. High dose radiation damage in nuclear energy structural materials investigated by heavy ion irradiation simulation

    International Nuclear Information System (INIS)

    Zheng Yongnan; Xu Yongjun; Yuan Daqing

    2014-01-01

    Structural materials in ITER, ADS and fast reactor suffer high dose irradiations of neutrons and/or protons, that leads to severe displacement damage up to lOO dpa per year. Investigation of radiation damage induced by such a high dose irradiation has attracted great attention along with the development of nuclear energy facilities of new generation. However, it is deeply hampered for the lacking of high dose neutron and proton sources. Irradiation simulation of heavy ions produced by accelerators opens up an effective way for laboratory investigation of high dose irradiation induced radiation damage encountered in the ITER, ADS, etc. Radiation damage is caused mainly by atomic displacement in materials. The displacement rate of heavy ions is about lO 3 ∼10 7 orders higher than those of neutrons and protons. High displacement rate of heavy ions significantly reduces the irradiation time. The heavy ion irradiation simulation technique (HIIS) technique has been developed at China Institute of Atomic Energy and a series of the HIIS experiments have been performed to investigate radiation damage in stainless steels, tungsten and tantalum at irradiation temperatures from room temperature to 800 ℃ and in the irradiation dose region up to 100 dpa. The experimental results show that he radiation swelling peak for the modified stainless steel appears in the temperature region around 580 ℃ and the radiation damage is more sensitive to the temperature, the size of the radiation induced vacancy cluster or void increase with the increasing of the irradiation dose, and among the three materials the home-made modified stainless steel has the best radiation resistant property. (authors)

  17. Temperature response of biological materials to pulsed non-ablative CO2 laser irradiation

    NARCIS (Netherlands)

    Brugmans, M. J.; Kemper, J.; Gijsbers, G. H.; van der Meulen, F. W.; van Gemert, M. J.

    1991-01-01

    This paper presents surface temperature responses of various tissue phantoms and in vitro and in vivo biological materials in air to non-ablative pulsed CO2 laser irradiation, measured with a thermocamera. We studied cooling off behavior of the materials after a laser pulse, to come to an

  18. Merton and Ziman's modes of science: the case of biological and similar material transfer agreements

    NARCIS (Netherlands)

    Rodriguez, V.F.

    2007-01-01

    This paper makes a connection between recent studies on research materials exchange and its effect on the progress of science. Academia fears that scientific development could be hampered by the privatised practices of research material exchange. Since post-academic science represents a sufficient

  19. A study on the irradiation effect of reactor materials using a cyclotron

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Joon Hwa; Ji, Se Hwan; Kang, Yung Hwan; Park, Duk Keun; Park, Jong Man; Lee, Bong Sang; Oh, Jong Myung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-12-01

    The objectives of the present study are to develop the simulation techniques of neutron irradiation through ion irradiation using a Cyclotron and small specimen techniques and to evaluate radiation effects of reactor materials. Effects of proton or neutron irradiation on domestic 12Cr-1MoV and SA508-3 steels were evaluated by small scale specimen test techniques, i.e, small punch and miniaturized tensile test. In order to study the radiation damage mechanism, irradiation effects of the steels were investigated by means of property change tests such as microstructure, physical and thermal properties. Feasibility study on application of a magnetic non-destructive methods to evaluate radiation effects on RPV materials was performed. 109 figs, 12 tabs, 102 refs. (Author).

  20. A study on the irradiation effect of reactor materials using a cyclotron

    International Nuclear Information System (INIS)

    Hong, Joon Hwa; Ji, Se Hwan; Kang, Yung Hwan; Park, Duk Keun; Park, Jong Man; Lee, Bong Sang; Oh, Jong Myung

    1994-12-01

    The objectives of the present study are to develop the simulation techniques of neutron irradiation through ion irradiation using a Cyclotron and small specimen techniques and to evaluate radiation effects of reactor materials. Effects of proton or neutron irradiation on domestic 12Cr-1MoV and SA508-3 steels were evaluated by small scale specimen test techniques, i.e, small punch and miniaturized tensile test. In order to study the radiation damage mechanism, irradiation effects of the steels were investigated by means of property change tests such as microstructure, physical and thermal properties. Feasibility study on application of a magnetic non-destructive methods to evaluate radiation effects on RPV materials was performed. 109 figs, 12 tabs, 102 refs. (Author)

  1. Inorganic-organic hybrid polymer for preparation of affiliating material using electron beam irradiation

    International Nuclear Information System (INIS)

    Chung, Jaeseung; Kim, Seongeun; Kim, Byounggak; Lee, Jongchan; Park, Jihyun; Lee, Byeongcheol

    2011-01-01

    Recently, silver nano materials have gained a lot of attentions in a variety of applications due to the unique biological, optical, and electrical properties. Especially, the antifouling property of these material is considered to be an important character for biomedical field, marine coatings industry, biosensor, and drug delivery. In this study, we design and synthesize the inorganic-organic hybrid polymer for preparation of affiliating materials. Silver nano materials having antifouling property with different shapes are prepared by control the electron beam irradiation conditions. Inorganic-organic hybrid polymer was synthesized and characterized. → Morphology and size controlled nano materials are prepared using electron beam irradiation. → Silver nano materials having various shapes can be used for antifouling material

  2. Effects of gamma-rays irradiation on tracking resistance of organic insulating materials

    Energy Technology Data Exchange (ETDEWEB)

    Du, Boxue; Suzuki, Akio; Kobayashi, Shigeo [Tokyo Univ. of Agriculture and Technology, Koganei (Japan). Faculty of Technology

    1996-04-01

    This paper describes the influence of gamma-rays irradiation on tracking failure of organic insulating materials by use of the IEC Publ.112 method. Tracking resistance of organic insulating materials under wet polluted condition has been studied by many investigators with a test method of the IEC Publ.112. The investigations on irradiation effects on tracking resistance should be enhanced due to the increasing usage of organic insulating materials in the radiation environments. The tracking resistance seems to be affected by gamma-irradiation, but the knowledge on the influence of gamma-irradiation is quite a few and systematic studies are needed. In this paper, modified polyphenylene oxide, polybutylene naphthalate, modified polycarbonate and polybutylene terephthalate which were irradiated in air until 1x10{sup 7}R and 1x10{sup 8}R with dose rate of 10{sup 6}R/hr using {sup 60}Co gamma-source have been employed. The total dose effects on the number of drops to tracking failure, contact angle and charges of scintillation have been studied. As the total doses are increased, the number of drops to tracking failure decreases with polybutylene terephthalate. On the other hand, the number of drops to tracking failure increases with polybutylene naphthalate and modified polycarbonate when the total doses are increased. The effects of gamma-rays irradiation on tracking failure are due to radiation-induced degradation or cross-linking of organic insulating materials. When the organic insulating materials are degraded by gamma-irradiation, the tracking resistance decreases, but for cross-linking type materials, the tracking resistance increases. (author)

  3. Irradiation as an alternative environmentally friendly method for microbiological decontamination of herbal raw material

    International Nuclear Information System (INIS)

    Dragusin, M.; Rotaru, R.

    2000-01-01

    Microbiological contamination of herbal raw materials is a serious problem in the production of therapeutical preparations. A good quality of the product, according to the pharmaceutical requirements may be achieved by applying suitable methods of decontamination. The decontamination treatments should be fast and effective against all microorganisms. It should ensure the decontamination of both packaging and the microorganisms present and must not reduce the sensory and technological qualities of the commodities. Decontamination of herbal raw materials by irradiation is a method by choice. It is because chemical methods are recognized recently as not safe to the consumer. Irradiation, in turn, is technically feasible, very effective and friendly enough to environment process. Under the prevailing production and handling conditions, most herbs contain a large number of microorganisms what is a serious problem in the production of therapeutical preparations. For several years the most widely used methods for decontamination of herbs was fumigation with ethylene oxide or methyl bromide. Both methods today banned in most countries. Irradiation is an alternative and safe method for effective reducing the microbial contamination of herbal raw materials. The following raw materials have been examined: Folium Cynara, Folium Plantago, Flos Chamomillae, Semen Sylibum Marianum and Folium Farfara. The content of biologically active compounds before and after irradiation of the raw materials did not change in a significant degree after irradiation. The dose of radiation for herbals raw materials was 10 kGy. There are two groups of raw materials: - The raw materials designed for preparing granulates, tablets, dragees, capsules, aqueous extracts, infusions, macerations and preparations for external use; - The raw materials assigned for preparing alcoholic preparations, isolated compounds, oil preparations and essential oils. The medical herbs and herbal raw materials before their

  4. Effect of packaging material on nitrate nitrogen content of irradiated potatoes

    International Nuclear Information System (INIS)

    Mondy, N.I.; Koushik, S.R.

    1990-01-01

    The effect of packaging materials on nitrate nitrogen content of irradiated potatoes was investigated. Tubers were irradiated at 10, 30 and 100 Krads and stored for 12 wk at 5 degrees C in paper or plastic bags. Nitrate nitrogen content was significantly (p 0.01) higher in tubers packaged in plastic as compared to those in paper bags. Irradiation significantly (p 0.01) increased nitrate nitrogen content between the lowest and highest levels of treatment in tubers stored in both paper and plastic bags

  5. Method and equipment to lead a cable-like material under an irradiation source

    International Nuclear Information System (INIS)

    Riesselmann, F.J.

    1975-01-01

    When irradiating cable-like material (cable jacketed with polyethylene) which is led through an irradiation source and is thus turned and twisted, no uniform irradiation and twist changes have so far been obtained. It is suggested to twist the cable before the first circuit by about 45 0 in one direction, after turning and the second circuit, to twist by about 90 0 in the other direction and to follow with a further two circuits with twisting. A suitable cable twisting device which works with discrete clamping jaw is described in detail. (UWI) [de

  6. The effect of neutron irradiation on the trapping of tritium in carbon-based materials

    International Nuclear Information System (INIS)

    Kwast, H.; Werle, H.; Glugla, M.; Wu, C.H.; Federici, G.

    1993-11-01

    Carbon-based materials are considered for protection of plasma facing components in the next step fusion device. To investigate the effects of neutron damage on the tritium behaviour an experimental study on the tritium retention of various neutron irradiated graphites and carbon/carbon fibre composites was started. The irradiation dose of the specimens ranges from 10 -3 to 3.5 dpa.g and the irradiation temperature from 390 C to 1500 C. A comparison of tritium retention in pre- and post-irradiated carbon-based materials as a function of the sample temperature is reported in this paper and the results are discussed. The first results indicate that the retention of tritium is higher in irradiated graphite than in unirradiated graphite and depends largely on the density and microstructure. The retention is also influenced by the tritium-loading temperature. Graphite of type S 1611, irradiated at 400 C and 600 C up to a damage of 0.1 dpa.g, retained about two times more tritium than the unirradiated material. (orig.)

  7. Design and fabrication of irradiation testing capsule for research reactor materials

    International Nuclear Information System (INIS)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu

    2012-01-01

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed

  8. Design and fabrication of irradiation testing capsule for research reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed.

  9. Effects of CTR irradiation on the mechanical properties of structural materials

    International Nuclear Information System (INIS)

    Wiffen, F.W.

    1976-11-01

    Mechanical properties of CTR structural materials are important in determining the reliability and economics of fusion power. Furthermore, these properties are significantly affected by the high neutron flux experienced by components in the regions near the plasma of the fusion reactor. In general, irradiation hardens the material and leads to a reduction in ductility. An exception to this is in some complex engineering alloys where either hardening or softening can be observed depending on the alloy and the irradiation conditions. Regardless of this restriction, irradiation usually leads to a reduction in ductility. Available tensile data examined in this paper show that significant ductility reduction can be found for irradiation conditions typical of CTR operation. Consideration of these effects show that extensive work will be needed to fully establish the in-service properties of CTR structures. This information will be used by designers to develop conditions and design philosophies adapted to avoid the most deleterious conditions and minimize stresses on structures on reactor design. The information will also be used as input to alloy development programs with goals of producing materials more resistant to property degradation during irradiation. It is clear that a great deal of additional work will be required both to understand the effect of CTR irradiation on properties and to develop optimal alloys for this application

  10. Residual stress improvement mechanism on metal material by underwater laser irradiation

    International Nuclear Information System (INIS)

    Sano, Yuji; Yoda, Masaki; Mukai, Naruhiko; Obata, Minoru; Kanno, Masanori

    2000-01-01

    Residual stress improvement technology for component surface by underwater pulsed laser irradiation has been developed as a method of preventing stress corrosion cracking (SCC) of core components in nuclear reactors. In order to optimize the laser irradiation conditions based on a complete understanding of the mechanism, the propagation of a shock wave induced by the impulse of laser irradiation and the dynamic response of the irradiated material were analyzed through time-dependent elasto-plastic calculations with a finite element program. The calculated results are compared with the measured results obtained by experiments in which laser pulses with an energy of 200 mJ are focused to a diameter of 0.8 mm on a water-immersed test piece of 20% cold-worked Type 304 austenitic stainless steel to simulate neutron irradiation hardening. A residual compressive stress, which is nearly equivalent to the yield stress of the processed material, remains on the material surface after passage of the shock wave with enough amplitude to induce a permanent strain. Multiple irradiation of laser pulses extends the stress-improved depth to about 1 mm, which would be the limit corresponding to the three-dimensional dispersion effect of the shock wave. (author)

  11. Effect of γ-irradiation on commercial polypropylene based mono and multi-layered retortable food packaging materials

    Science.gov (United States)

    George, Johnsy; Kumar, R.; Sajeevkumar, V. A.; Sabapathy, S. N.; Vaijapurkar, S. G.; Kumar, D.; Kchawahha, A.; Bawa, A. S.

    2007-07-01

    Irradiation processing of food in the prepackaged form may affect chemical and physical properties of the plastic packaging materials. The effect of γ-irradiation doses (2.5-10.0 kGy) on polypropylene (PP)-based retortable food packaging materials, were investigated using Fourier transform infrared (FTIR) spectroscopic analysis, which revealed the changes happening to these materials after irradiation. The mechanical properties decreased with irradiation while oxygen transmission rate (OTR) was not affected significantly. Colour measurement indicated that Nylon 6 containing multilayer films became yellowish after irradiation. Thermal characterization revealed the changes in percentage crystallinity.

  12. Effect of γ-irradiation on commercial polypropylene based mono and multi-layered retortable food packaging materials

    International Nuclear Information System (INIS)

    George, Johnsy; Kumar, R.; Sajeevkumar, V.A.; Sabapathy, S.N.; Vaijapurkar, S.G.; Kumar, D.; Kchawahha, A.; Bawa, A.S.

    2007-01-01

    Irradiation processing of food in the prepackaged form may affect chemical and physical properties of the plastic packaging materials. The effect of γ-irradiation doses (2.5-10.0 kGy) on polypropylene (PP)-based retortable food packaging materials, were investigated using Fourier transform infrared (FTIR) spectroscopic analysis, which revealed the changes happening to these materials after irradiation. The mechanical properties decreased with irradiation while oxygen transmission rate (OTR) was not affected significantly. Colour measurement indicated that Nylon 6 containing multilayer films became yellowish after irradiation. Thermal characterization revealed the changes in percentage crystallinity

  13. Effect of {gamma}-irradiation on commercial polypropylene based mono and multi-layered retortable food packaging materials

    Energy Technology Data Exchange (ETDEWEB)

    George, Johnsy [Defence Food Research Laboratory, Siddarthanagar, Mysore, Karnataka 570011 (India)]. E-mail: g.johnsy@gmail.com; Kumar, R. [Defence Food Research Laboratory, Siddarthanagar, Mysore, Karnataka 570011 (India); Sajeevkumar, V.A. [Defence Food Research Laboratory, Siddarthanagar, Mysore, Karnataka 570011 (India); Sabapathy, S.N. [Defence Food Research Laboratory, Siddarthanagar, Mysore, Karnataka 570011 (India); Vaijapurkar, S.G. [Defence Laboratory, Ratanada Palace, Jodhpur, Rajastan 342011 (India); Kumar, D. [Defence Laboratory, Ratanada Palace, Jodhpur, Rajastan 342011 (India); Kchawahha, A. [Defence Laboratory, Ratanada Palace, Jodhpur, Rajastan 342011 (India); Bawa, A.S. [Defence Food Research Laboratory, Siddarthanagar, Mysore, Karnataka 570011 (India)

    2007-07-15

    Irradiation processing of food in the prepackaged form may affect chemical and physical properties of the plastic packaging materials. The effect of {gamma}-irradiation doses (2.5-10.0 kGy) on polypropylene (PP)-based retortable food packaging materials, were investigated using Fourier transform infrared (FTIR) spectroscopic analysis, which revealed the changes happening to these materials after irradiation. The mechanical properties decreased with irradiation while oxygen transmission rate (OTR) was not affected significantly. Colour measurement indicated that Nylon 6 containing multilayer films became yellowish after irradiation. Thermal characterization revealed the changes in percentage crystallinity.

  14. Researches, development and characterization of dosimetric materials for monitoring in irradiation processes with high doses

    International Nuclear Information System (INIS)

    Galante, Ana Maria Sisti

    2003-01-01

    Dosimetric materials that can be produced in Brazil with material acquired in the national market to replace the imported dosimeters used in radiation processing were developed in this work. Mixtures of potassium nitrate and sensitizers compounds as manganese dioxide, barium nitrate and potassium bromide were prepared in the pellet form. Dosimetric characteristics such as dose-response useful range, sensitivity, environmental conditions and dose rate influences were evaluated in 60 Co gamma radiation fields. Dyed polymethylmethacrylate detectors were also produced and its dosimetric characteristics were evaluated. The main characteristics evaluated in this case were: dose response useful range sensitivity, environmental conditions, dose rate influences and radiation energy dependence in gamma radiation fields and accelerated electrons beam of 0.8 to 1.5 MeV. The applied analytic technique was spectrophotometry. The calibration was performed in the irradiation facilities belonging to IPEN and certified by the International Atomic Energy Agency by means of the program IDAS (International Dose Assurance Service ) using the Fricke dosimeter. The mixture of potassium nitrate and manganese dioxide presented the best results and a wide dose range between 200 and 600 kGy. The response of the developed polymethylmethacrylate detectors are similar to the imported detectors and the dose range is characteristic to each detector and depends on the dye added in its formulation. (author)

  15. Mimicking lizard-like surface structures upon ultrashort laser pulse irradiation of inorganic materials

    Science.gov (United States)

    Hermens, U.; Kirner, S. V.; Emonts, C.; Comanns, P.; Skoulas, E.; Mimidis, A.; Mescheder, H.; Winands, K.; Krüger, J.; Stratakis, E.; Bonse, J.

    2017-10-01

    Inorganic materials, such as steel, were functionalized by ultrashort laser pulse irradiation (fs- to ps-range) to modify the surface's wetting behavior. The laser processing was performed by scanning the laser beam across the surface of initially polished flat sample material. A systematic experimental study of the laser processing parameters (peak fluence, scan velocity, line overlap) allowed the identification of different regimes associated with characteristic surface morphologies (laser-induced periodic surface structures, grooves, spikes, etc.). Analyses of the surface using optical as well as scanning electron microscopy revealed morphologies providing the optimum similarity to the natural skin of lizards. For mimicking skin structures of moisture-harvesting lizards towards an optimization of the surface wetting behavior, additionally a two-step laser processing strategy was established for realizing hierarchical microstructures. In this approach, micrometer-scaled capillaries (step 1) were superimposed by a laser-generated regular array of small dimples (step 2). Optical focus variation imaging measurements finally disclosed the three dimensional topography of the laser processed surfaces derived from lizard skin structures. The functionality of these surfaces was analyzed in view of wetting properties.

  16. Use of the SPIRAL 2 facility for material irradiations with 14 MeV energy neutrons

    International Nuclear Information System (INIS)

    Mosnier, A.; Ridikas, D.; Ledoux, X.; Pellemoine, F.; Anne, R.; Huguet, Y.; Lipa, M.; Magaud, P.; Marbach, G.; Saint-Laurent, M.G.; Villari, A.C.C.

    2005-01-01

    The primary goal of an irradiation facility for fusion applications will be to generate a material irradiation database for the design, construction, licensing and safe operation of a fusion demonstration power station (e.g., DEMO). This will be achieved through testing and qualifying material performance under neutron irradiation that simulates service up to the full lifetime anticipated in the power plant. Preliminary investigations of 14 MeV neutron effects on different kinds of fusion material could be assessed by the SPIRAL 2 Project at GANIL (Caen, France), aiming at rare isotope beams production for nuclear physics research with first beams expected by 2009. In SPIRAL 2, a deuteron beam of 5 mA and 40 MeV interacts with a rotating carbon disk producing high-energy neutrons (in the range between 1 and 40 MeV) via C (d, xn) reactions. Then, the facility could be used for 3-4 months y -1 for material irradiation purposes. This would correspond to damage rates in the order of 1-2 dpa y -1 (in Fe) in a volume of ∼10 cm 3 . Therefore, the use of miniaturized specimens will be essential in order to effectively utilize the available irradiation volume in SPIRAL 2. Sample package irradiation temperature would be in the range of 250-1000 deg. C. The irradiation level of 1-2 dpa y -1 with 14 MeV neutrons (average energy) may be interesting for micro-structural and metallurgical investigations (e.g., mini-traction, small punch tests, etc.) and possibly for the understanding of specimen size/geometric effects of critical material properties. Due to the small test cell volume, sample in situ experiments are not foreseen. However, sample packages would be, if required, available each month after transfer in a special hot cell on-site

  17. Irradiation capability of Japanese materials test reactor for water chemistry experiments

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Hata, Kuniki; Chimi, Yasuhiro; Nishiyama, Yutaka; Nakamura, Takehiko

    2012-09-01

    Appropriate understanding of water chemistry in the core of LWRs is essential as chemical species generated due to water radiolysis by neutron and gamma-ray irradiation govern corrosive environment of structural materials in the core and its periphery, causing material degradation such as stress corrosion cracking. Theoretical model calculation such as water radiolysis calculation gives comprehensive understanding of water chemistry at irradiation field where we cannot directly monitor. For enhancement of the technology, accuracy verification of theoretical models under wide range of irradiation conditions, i.e. dose rate, temperature etc., with well quantified in-pile measurement data is essential. Japan Atomic Energy Agency (JAEA) has decided to launch water chemistry experiments for obtaining data that applicable to model verification as well as model benchmarking, by using an in-pile loop which will be installed in the Japan Materials Testing Reactor (JMTR). In order to clarify the irradiation capability of the JMTR for water chemistry experiments, preliminary investigations by water radiolysis / ECP model calculations were performed. One of the important irradiation conditions for the experiments, i.e. dose rate by neutron and gamma-ray, can be controlled by selecting irradiation position in the core. In this preliminary study, several representative irradiation positions that cover from highest to low absorption dose rate were chosen and absorption dose rate at the irradiation positions were evaluated by MCNP calculations. As a result of the calculations, it became clear that the JMTR could provide the irradiation conditions close to the BWR. The calculated absorption dose rate at each irradiation position was provided to water radiolysis calculations. The radiolysis calculations were performed under various conditions by changing absorption dose rate, water chemistry of feeding water etc. parametrically. Qualitatively, the concentration of H 2 O 2 , O 2 and

  18. Neutron irradiation test of copper alloy/stainless steel joint materials

    International Nuclear Information System (INIS)

    Yamada, Hirokazu; Kawamura, Hiroshi

    2006-01-01

    As a study about the joint technology of copper alloy and stainless steel for utilization as cooling piping in International Thermonuclear Experimental Reactor (ITER), Al 2 O 3 -dispersed strengthened copper or CuCrZr was jointed to stainless steel by three kinds of joint methods (casting joint, brazing joint and friction welding method) for the evaluation of the neutron irradiation effect on joints. A neutron irradiation test was performed to three types of joints and each copper alloy. The average value of fast neutron fluence in this irradiation test was about 2 x 10 24 n/m 2 (E>1 MeV), and the irradiation temperature was about 130degC. As post-irradiation examinations, tensile tests, hardness tests and observation of fracture surface after the tensile tests were performed. All type joints changed to be brittle by the neutron irradiation effect like each copper alloy material, and no particular neutron irradiation effect due to the effect of joint process was observed. On the casting and friction welding, hardness of copper alloy near the joint boundary changed to be lower than that of each copper alloy by the effect of joint procedure. However, tensile strength of joints was almost the same as that of each copper alloy before/after neutron irradiation. On the other hand, tensile strength of joints by brazing changed to be much lower than CuAl-25 base material by the effect of joint process before/after neutron irradiation. Results in this study showed that the friction welding method and the casting would be able to apply to the joint method of piping in ITER. This report is based on the final report of the ITER Engineering Design Activities (EDA). (author)

  19. The irradiation induced creep in fuel compact materials for H.T.R. applications

    International Nuclear Information System (INIS)

    Veringa, H.; Blackstone, R.; Loelgen, R.

    1976-01-01

    Restrained shrinkage experiments up to 3 x 10 21 ncm -2 (DNE) in the temperature range of 600-1,200 0 C on three different dummy coated particle fuel compact materials were performed in the High Flux Reactor at Petten, the Netherlands. The data were evaluated to obtain the steady state irradiation creep coefficient of the compacts. It was found that for the materials investigated, the creep coefficient is temperature-dependent, but no clear relationship to the Young's modulus could be established. Under certain conditions, this irradiation-induced plasticity influences the elastic properties, while also the creep coefficient increases. This effect coincides with the formation and further opening of cracks due to stresses caused by irradiation shrinkage of the matrix material. (orig.) [de

  20. Irradiation as an alternative environment friendly method for microbiological decontamination of herbal raw material

    International Nuclear Information System (INIS)

    Gorecki, P.; Kedzia, B.; Migdal, W.; Owczarczyk, H.B.

    1998-01-01

    Microbiological contamination of herbal raw materials is a serious problem in the production of therapeutical preparations. A good quality of the product, according to the pharmaceutical requirements may be achieved by applying suitable methods of decontamination. The decontamination treatments should be fast and effective against all microorganisms. It should ensure the decontamination of both packaging and the product in order to act effectively against all the microorganisms present and must not reduce the sensory and technological qualities of the commodities. In the paper, the results of comparative investigations on the microbiological decontamination of herbal raw materials by chemical (ethylene oxide, methyl bromide) and physical method (irradiation) are presented. Decontamination of herbal raw materials by irradiation is a method by choice. It is because chemical methods have been recognized recently as not safe to the consumer. Irradiation, in turn, is technically feasible, very effective and friendly enough to environment process

  1. Remote-handling demonstration tests for the Fusion Materials Irradiation Test (FMIT) Facility

    International Nuclear Information System (INIS)

    Shen, E.J.; Hussey, M.W.; Kelly, V.P.; Yount, J.A.

    1982-01-01

    The mission of the Fusion Materials Irradiation Test (FMIT) Facility is to create a fusion-like environment for fusion materials development. Crucial to the success of FMIT is the development and testing of remote handling systems required to handle materials specimens and maintenance of the facility. The use of full scale mock-ups for demonstration tests provides the means for proving these systems

  2. International Fusion Materials Irradiation Facility conceptual design activity. Present status and perspective

    International Nuclear Information System (INIS)

    Kondo, Tatsuo; Noda, Kenji; Oyama, Yukio

    1998-01-01

    For developing the materials for nuclear fusion reactors, it is indispensable to study on the neutron irradiation behavior under fusion reactor conditions, but there is not any high energy neutron irradiation facility that can simulate fusion reactor conditions at present. Therefore, the investigation of the IFMIF was begun jointly by Japan, USA, Europe and Russia following the initiative of IEA. The conceptual design activities were completed in 1997. As to the background and the course, the present status of the research on heavy irradiation and the testing means for fusion materials, the requirement and the technical basis of high energy neutron irradiation, and the international joint design activities are reported. The materials for fusion reactors are exposed to the neutron irradiation with the energy spectra up to 14 MeV. The requirements from the users that the IFMIF should satisfy, the demand of the tests for the materials of prototype and demonstration fusion reactors and the evaluation of the neutron field characteristics of the IFMIF are discussed. As to the conceptual design of the IFMIF, the whole constitution, the operational mode, accelerator system and target system are described. (K.I.)

  3. The Influence of Irradiation Regimes on Retention Hydrogen Isotopes in Structural Materials

    International Nuclear Information System (INIS)

    Zaluzhnyi, A.

    2007-01-01

    Full text of publication follows: In the present work was investigated the influence of irradiation regimes on retention hydrogen isotopes in samples of austenitic steel during heating. The samples of studied materials were irradiated both in the reactor and by hydrogen isotopes ions of different energies and fluencies bombardment in an accelerator. Kinetic of hydrogen release from the samples worked with deuterium plasma was investigated. The following results were obtained. Heating the irradiate d samples of steel (irradiated in the reactor or by hydrogen isotopes ions bombardment), which have been kept in normal temperature during quite a long period after the irradiation, a shift of the diffusion peak of hydrogen release to higher temperatures, comparing to no irradiated samples, was observed. It means that atoms of hydrogen in the irradiated sample were caught by radiation defects, which are very effective as traps for hydrogen atoms till quite high temperatures (700 K). The worked out analysis of the received results supposes that vacancy complexes. On thermodesorption curves of hydrogen release from irradiated samples of austenitic steels a high temperature peak (900-1000 K) was observed because of dissociation of hydrogen containing compounds in micro pores. During investigations of hydrogen release from irradiated samples of austenitic steel, after it had been saturated with hydrogen plasma, abnormally big blisters were registered with cover thickness of about 1 mkm. Three peaks were observed on the thermodesorption curves of hydrogen release from irradiated samples, contained blisters. The low temperature spike (∼500 K) was showed to correspond to hydrogen release because of its resolution from blisters, where it was in molecular form. The high temperature peak (∼900 K) corresponds to hydrogen release from dissociating blisters, which contain hydrocarbons. The mechanism of abnormal blisters generation is offered. Inasmuch methane is not soluble in

  4. The effect of neutron irradiation on the structure and properties of carbon-carbon composite materials

    International Nuclear Information System (INIS)

    Burchell, T.D.; Eatherly, W.P.; Robbins, J.M.; Strizak, J.P.

    1991-01-01

    Carbon-based materials are an attractive choice for fusion reactor plasma facing components (PFCs) because of their low atomic number, superior thermal shock resistance, and low neutron activation. Next generation plasma fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER), will require advanced carbon-carbon composite materials possessing extremely high thermal conductivity to manage the anticipated severe heat loads. Moreover, ignition machines such as ITER will produce high neutron fluxes. Consequently, the influence of neutron damage on the structure and properties of carbon-carbon composite materials must be evaluated. Data from an irradiation experiment are reported and discussed here. Fusion relevant graphite and carbon-carbon composites were irradiated in a target capsule in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). A peak damage dose of 1.59 dpa at 600 degrees C was attained. The carbon materials irradiated included nuclear graphite grade H-451 and one-, two-, and three-directional carbon-carbon composite materials. Dimensional changes, thermal conductivity and strength are reported for the materials examined. The influence of fiber type, architecture, and heat treatment temperature on properties and irradiation behavior are reported. Carbon-Carbon composite dimensional changes are interpreted in terms of simple microstructural models

  5. Femtosecond Laser Irradiation of Plasmonic Nanoparticles in Polymer Matrix: Implications for Photothermal and Photochemical Material Alteration

    Directory of Open Access Journals (Sweden)

    Anton A. Smirnov

    2014-11-01

    Full Text Available We analyze the opportunities provided by the plasmonic nanoparticles inserted into the bulk of a transparent medium to modify the material by laser light irradiation. This study is provoked by the advent of photo-induced nano-composites consisting of a typical polymer matrix and metal nanoparticles located in the light-irradiated domains of the initially homogeneous material. The subsequent irradiation of these domains by femtosecond laser pulses promotes a further alteration of the material properties. We separately consider two different mechanisms of material alteration. First, we analyze a photochemical reaction initiated by the two-photon absorption of light near the plasmonic nanoparticle within the matrix. We show that the spatial distribution of the products of such a reaction changes the symmetry of the material, resulting in the appearance of anisotropy in the initially isotropic material or even in the loss of the center of symmetry. Second, we analyze the efficiency of a thermally-activated chemical reaction at the surface of a plasmonic particle and the distribution of the product of such a reaction just near the metal nanoparticle irradiated by an ultrashort laser pulse.

  6. Capsule development and utilization for material irradiation tests

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Hwan; Kim, B. G.; Joo, K. N. [and others

    2000-05-01

    The development program of advanced nuclear structural and fuel materials includes the in-pile tests using the instrumented capsule at HANARO. The tests were performed in the in-core test holes of CT, IR 1 and 2 and OR 4 and 5 of HANARO. Extensive efforts have also been made to establish design and manufacturing technology for the instrumented capsule and its related system, which should be compatible with the HANARO's characteristics. Since the first instrumented capsule(97M-01K) had been designed and successfully fabricated, five tests were done to support the users and provided the economic benefits to user by generating the essential in-pile information on the performance and structural integrity of materials. This paper describes the present status and future plans of these R and D activities for the development of the instrumented capsule including in-situ material property measurement capsules and nuclear fuel test capsules.

  7. Capsule development and utilization for material irradiation tests

    International Nuclear Information System (INIS)

    Kang, Young Hwan; Kim, B. G.; Joo, K. N.

    2000-05-01

    The development program of advanced nuclear structural and fuel materials includes the in-pile tests using the instrumented capsule at HANARO. The tests were performed in the in-core test holes of CT, IR 1 and 2 and OR 4 and 5 of HANARO. Extensive efforts have also been made to establish design and manufacturing technology for the instrumented capsule and its related system, which should be compatible with the HANARO's characteristics. Since the first instrumented capsule(97M-01K) had been designed and successfully fabricated, five tests were done to support the users and provided the economic benefits to user by generating the essential in-pile information on the performance and structural integrity of materials. This paper describes the present status and future plans of these R and D activities for the development of the instrumented capsule including in-situ material property measurement capsules and nuclear fuel test capsules

  8. Capsule development and utilization for material irradiation tests

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Hwan; Kim, B G; Joo, K N [and others

    2000-05-01

    The development program of advanced nuclear structural and fuel materials includes the in-pile tests using the instrumented capsule at HANARO. The tests were performed in the in-core test holes of CT, IR 1 and 2 and OR 4 and 5 of HANARO. Extensive efforts have also been made to establish design and manufacturing technology for the instrumented capsule and its related system, which should be compatible with the HANARO's characteristics. Since the first instrumented capsule(97M-01K) had been designed and successfully fabricated, five tests were done to support the users and provided the economic benefits to user by generating the essential in-pile information on the performance and structural integrity of materials. This paper describes the present status and future plans of these R and D activities for the development of the instrumented capsule including in-situ material property measurement capsules and nuclear fuel test capsules.

  9. Instrumentation Technologies for Improving an Irradiation Testing of Nuclear Fuels and Materials at the HANARO

    International Nuclear Information System (INIS)

    Kim, Bong Goo; Park, Sung Jae; Choo, Ki Nam

    2011-01-01

    Over 50 years of nuclear fuels and materials irradiation testing has led to many countries developing significant improvements in instrumentation to monitor physical parameters and to control the test conditions in Materials Test Reactors (MTRs) or research reactors. Recent effort to deploy new fuels and materials in existing and advanced reactors has increased the demand for well-instrumented irradiation tests. Specifically, demand has increased for tests with sensors capable of providing real-time measurement of key parameters, such as temperature, geometry changes, thermal conductivity, fission gas release, cracking, coating buildup, thermal and fast flux, etc. This review paper documents the current state of instrumentation technologies in MTRs in the world and summarizes on-going research efforts to deploy new sensors. There is increased interest to irradiate new materials and reactor fuels for advanced PWRs and the Gen-IV reactor systems, such as SFRs (Sodium-cooled Fast Reactors), VHTRs (Very-High-Temperature Reactors), SCWRs (Supercritical-Water-cooled Reactors) and GFRs (Gas-cooled Fast Reactor). This review documents the current state of instrumentation technologies in MTRs in the world, identifies challenges faced by previous testing methods and how these challenges were overcome. A wide range of sensors are available to measure key parameters of interest during fuels and materials irradiations in MTRs. Such sensors must be reliable, small size, highly accurate, and able to withstand harsh conditions. On-going development efforts are focusing on providing MTR users a wider range of parameter measurements with increased accuracy. In addition, development efforts are focusing on reducing the impact of sensor on measurements by reducing sensor size. This report includes not only status of instrumentation using research reactors in the world to irradiate nuclear fuels and materials but also future directions relating to instrumentation technologies for

  10. Irradiation effects on material properties of steels used in nuclear reactors: a literature review

    International Nuclear Information System (INIS)

    Gerceker, N.; Dara, I. H.

    2001-01-01

    The structural materials of a nuclear power plant are of vital importance as they provide mechanical strength, structural support and physical containment for the primary reactor components as well as the nuclear power plant itself. These structural materials comprise mainly of metals and their alloys, ceramics and cermets. However, metals and their alloys are the most widely used materials and the irradiation effects are more pronounced on metallic materials as of their high temperature properties are more sensitive (with respect to ceramics and cermets) to any kind of external effects. The wholesale creation of effects on material properties has been studied for over four decades and it is not realistic to attempt to represent even a small part of the field in single poster paper. In the present contribution, a literature review of the irradiation effects on the material properties of different types of steel alloys will be given because steels are widely used as structural materials in reactors and therefore the irradiation effects on steels may be of paramount importance for reactor design, operation and safety concepts which will be discussed about radiation effects on material properties of steels will provide highlights to better understanding of the origins and development of radiation effects in materials

  11. Determination of rheological properties of fresh concrete and similar materials in a vibration rheometer

    Directory of Open Access Journals (Sweden)

    Sandra Juradin

    2012-02-01

    Full Text Available A vibration rheometer has been developed for the purpose of determining the viscosity coefficient and the yield value of fresh concrete under vibration. The main parts of the apparatus, a test specimen and a vibration source form a unitary oscillatory system whose parameters can be measured with sufficient precision. Two types of fine grained reference material have been prepared and examined; one with a high coefficient of viscosity and the other with a high yield value. The rheological properties of reference materials have been determined in a capillary tube viscometer. Since there is no analytical solution to the flow in the vibration rheometer, the constants of the vibration rheometer have been determined by experiment, for each position of the apparatus piston within the measuring range. The parameters of the flow depend on the maximum acceleration of the vibration source. An increase in acceleration causes an increase in the yield value as well as a reduction in the plastic viscosity coefficient of the material specimen. A testing of fresh mortar has been carried out as well. The obtained results have been compared with the impact on reference materials, which makes the results of our research applicable to fresh vibrated concrete.

  12. Thickness measuring apparatus and method for tire ply and similar materials

    International Nuclear Information System (INIS)

    Cho, B.Y.; Utt, O.L.

    1978-01-01

    The invention relates to method and apparatus for measuring the amount of material on opposite sides of the reinforcing cord of a tire ply, and more particularly relates to a backscatter X-ray radiation gauging apparatus and method for obtaining the aforesaid measurements. (author)

  13. Semiconductor-diode-aided dosimetry of the irradiation of pourable bulk material

    International Nuclear Information System (INIS)

    Gruenewald, T.; Rudolf, M.

    1987-01-01

    The irradiation of unpackaged pourable bulk material requires the employment of a dosimeter which can be readily transported along with the material. Planar diffused silicon diodes have been found to be suitable for this purpose. To date these have been used solely for the purpose of dose rate measurements; however, it can be shown that the permanent change in reverse recover time at the p-n junction correlates with the absorbed irradiation dose in the range up to 10 kGy. Appropriate selection of the diode and thermal treatment lead to a linear dependence and enable the silicon dosimeter to be reused. (author). 16 refs, 4 figs

  14. Binary-collision-approximation simulation for noble gas irradiation onto plasma facing materials

    International Nuclear Information System (INIS)

    Saito, Seiki; Nakamura, Hiroaki; Takayama, Arimichi; Ito, Atsushi M

    2014-01-01

    A number of experiments show that helium plasma constructs filament (fuzz) structures whose diameter is in nanometer-scale on the tungsten material under the suitable experimental condition. In this paper, binary-collision-approximation-based simulation is performed to reveal the mechanism and the conditions of fuzz formation of tungsten material under plasma irradiation. The irradiation of the plasma of hydrogen, deuterium, and tritium, and also the plasma of noble gas such as helium, neon, and argon atoms are investigated. The possibility of fuzz formation is discussed on the simulation result of penetration depth of the incident atoms

  15. Study of characteristics of gamma-irradiated materials for calorimeters

    International Nuclear Information System (INIS)

    Britvich, G.I.; Vasil'chenko, V.G.; Peresypkin, A.I.

    1992-01-01

    The radiation resistance of some structural materials proposed for use in electromagnetic calorimeters is studied. Particular attention is given to the spectral, dose, and other postradiation characteristics of pure heavy fluorides and their solid solutions: The promise of the use of CdF 2 and CdI 2 crystals in calorimeters is noted. 19 refs., 5 figs

  16. Test plan for the irradiation of nonmetallic materials.

    Energy Technology Data Exchange (ETDEWEB)

    Brush, Laurence H.; Farnum, Cathy Ottinger; Gelbard, Fred; Dahl, M.; Joslyn, C. C.; Venetz, T. J.

    2013-03-01

    A comprehensive test program to evaluate nonmetallic materials use in the Hanford Tank Farms is described in detail. This test program determines the effects of simultaneous multiple stressors at reasonable conditions on in-service configuration components by engineering performance testing.

  17. Test plan for the irradiation of nonmetallic materials.

    Energy Technology Data Exchange (ETDEWEB)

    Brush, Laurence H.; Farnum, Cathy Ottinger; Dahl, M.; Joslyn, C. C.; Venetz, T. J.

    2013-05-01

    A comprehensive test program to evaluate nonmetallic materials use in the Hanford tank farms is described in detail. This test program determines the effects of simultaneous multiple stressors at reasonable conditions on in-service configuration components by engineering performance testing.

  18. Surface modification of ceramic materials induced by irradiation of high power pulsed ICP

    International Nuclear Information System (INIS)

    Ishigaki, Takamasa; Okada, Nobuhiro; Ohashi, Naoki; Haneda, Hajime

    2003-01-01

    Newly developed pulse-modulated high-power inductively coupled plasma [ICP] is expected to offer the unique physico-chemical condition, such as the increased concentration of chemically reactive species, as well as the appropriate heat flux for materials processing. Two kinds of oxide materials, titanium and zinc oxide, were placed at the downstream of Ar-H 2 ICP and irradiated in the plasma of continuous [CN] and pulse-modulated [PM] modes. The CN-ICP irradiation at the position close to the plasma tail gave rise to the thermal reduction of oxides. In the PM-ICP irradiation, the degree of thermal reduction depended on the lower power level during pulse-off time, as well as the total electric power. Irradiation in PM-ICP led to the increased formation of oxygen vacancies in titanium dioxide. In the case of zinc oxide, the UV emission efficiency was improved by PM-ICP irradiation, while the green emission became predominant by CN-ICP irradiation at the appropriate position. Induced effects in the two oxides by PM-ICP would be related to the high concentration of hydrogen radicals in the plasma. (author)

  19. Fracture toughness and strength change of neutron-irradiated ceramic materials

    International Nuclear Information System (INIS)

    Dienst, W.; Zimmermann, H.

    1994-01-01

    In order to analyse the results of bending strength measurements on neutron-irradiated samples of Al 2 O 3 , AlN and SiC, fracture toughness measurements were additionally conducted. The neutron fluences concerned were mostly in the range of 0.6 to 3.2x10 26 n/m 2 at irradiation temperatures of 400 to 550 C. A fracture toughness decrease was generally observed for polycrystalline materials which, however, was considerably smaller than the reduction of the fracture strength. Exceptional increase of the fracture toughness seems typical for the effect of rather coarse irradiation defects. The irradiation-induced change of the fracture toughness of single crystal Al 2 O 3 appeared dependent on the crystallographic orientation; both reduced and increased fracture toughness after irradiation was observed. Recent results of neutron irradiation to about 2x10 25 n/m 2 at 100 C showed, that the strength decrease of various Al 2 O 3 grades sets in at (3-5)x10 24 n/m 2 and seems to be little dependent on the irradiation temperature. ((orig.))

  20. Irradiation and testing of compact ignition tokamak toroidal field coil insulation materials

    International Nuclear Information System (INIS)

    Kanemoto, G.K.; Sherick, M.J.; Sparks, D.C.

    1990-05-01

    This report documents the results of an irradiation and testing program performed on behalf of Martin Marietta Energy Systems, Inc. in support of the Compact Ignition Tokamak Research and Development program. The purpose of the irradiation and testing program was to determine the effects of neutron and gamma irradiation on the mechanical and electrical properties of candidate toroidal field coil insulation materials. Insulation samples were irradiated in the Advanced Test Reactor (ATR) in a large I-hole. The insulation samples were irradiated within a lead shield to reduce exposure to gamma radiation to better approximate the desired ration of neutron to gamma exposure. Two different exposure levels were specified for the insulation samples. To accomplish this, the samples were encapsulated in two separate aluminum capsules; the capsules positioned at the ATR core mid-plane and at the top of the fueled region to take advantage of the axial cosine distribution of the neutron and gamma flux; and by varying the length of irradiation time of the two capsules. Disassembly of the irradiated capsules and testing of the insulation samples were performed at the Test Reactor Area (TRA) Hot Cell Facilities. Testing of the samples included shear compression static, shear compression fatigue, flexure static, and electrical resistance measurements

  1. Evaluation of thermal shock strengths for graphite materials using a laser irradiation method

    International Nuclear Information System (INIS)

    Kim, Jae Hoon; Lee, Young Shin; Kim, Duck Hoi; Park, No Seok; Suh, Jeong; Kim, Jeng O.; Il Moon, Soon

    2004-01-01

    Thermal shock is a physical phenomenon that occurs during the exposure to rapidly high temperature and pressure changes or during quenching of a material. The rocket nozzle throat is exposed to combustion gas of high temperature. Therefore, it is important to select suitable materials having the appropriate thermal shock resistance and to evaluate these materials for rocket nozzle design. The material of this study is ATJ graphite, which is the candidate material for rocket nozzle throat. This study presents an experimental method to evaluate the thermal shock resistance and thermal shock fracture toughness of ATJ graphite using laser irradiation. In particular, thermal shock resistance tests are conducted with changes of specimen thickness, with laser source irradiated at the center of the specimen. Temperature distributions on the specimen surface are detected using type K and C thermocouples. Scanning electron microscope (SEM) is used to observe the thermal cracks on specimen surface

  2. Electron-beam-irradiation-induced crystallization of amorphous solid phase change materials

    Science.gov (United States)

    Zhou, Dong; Wu, Liangcai; Wen, Lin; Ma, Liya; Zhang, Xingyao; Li, Yudong; Guo, Qi; Song, Zhitang

    2018-04-01

    The electron-beam-irradiation-induced crystallization of phase change materials in a nano sized area was studied by in situ transmission electron microscopy and selected area electron diffraction. Amorphous phase change materials changed to a polycrystalline state after being irradiated with a 200 kV electron beam for a long time. The results indicate that the crystallization temperature strongly depends on the difference in the heteronuclear bond enthalpy of the phase change materials. The selected area electron diffraction patterns reveal that Ge2Sb2Te5 is a nucleation-dominated material, when Si2Sb2Te3 and Ti0.5Sb2Te3 are growth-dominated materials.

  3. Neutron and gamma irradiation effects on organic insulating materials for fusion magnets

    International Nuclear Information System (INIS)

    Maurer, W.

    1985-10-01

    Available low-temperature neutron and gamma irradiation data for organic insulating materials are collected and compared with room temperature data. Only the most promising polymers in terms of mechanical strength for magnet insulation are taken into account. For characterization and comparison of different materials the 75% dose is used, i.e. the dose, where the mechanical strength is reduced by 25%, and 75% is retained. For room temperature special prepared polyimide and epoxy materials reinforced with glass fibre retained 75% of the mechanical strength up to a dose of 7x10 7 Gy. For 5 K irradiation the best epoxy material retained the 75% dose up to 1x10 7 Gy, the best polyimide material up to 1x10 8 Gy. (orig.) [de

  4. Distribution of products in polymer materials induced by ion-beam irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Masaki; Kudoh, Hisaaki; Sasuga, Tsuneo; Seguchi, Tadao [Japan Atomic Energy Research Inst., Tokyo (Japan); Hama, Yoshimasa; Hamanaka, Ken-ichi; Matsumoto, Hideya

    1997-03-01

    The depth profile of double bond formed in low density polyethylene (LDPE) sheet by ion beams irradiation was observed by a micro FT-IR spectrometer in order to investigate the linear energy transfer (LET) dependency on radiation effects to polymer materials. The distribution of double bond formation in LDPE by irradiation of light ions as H+ was found to be same with the dose distribution calculated from TRIM code, and the yield was also same with that by gamma-rays irradiation, which means that the LET dependency is very small. However, the distribution of double bond to depth was much different from the calculated depth-dose in heavy ions irradiation as Ar and Kr. Then, the dose evaluation was difficult from the TRIM code calculation for heavy ions. (author)

  5. Effects of antioxidant and package materials on the quality of irradiated rugao ham

    International Nuclear Information System (INIS)

    Cao Hong; Chen Xiulan; Bao Jianzhong; Han Yan; Jiang Yunsheng; Wang Zhijun; Dong Jie; Yang Hairong; Xi Jun

    2008-01-01

    Irradiation could extend the shelf life of ham, but irradiation also facilitates the oxidation of fat. Different packaging materials and combination of antioxidants were used to deal with Rugao ham in order to lower the level of antioxidation caused by irradiation treatment. The peroxide value of fat was detected as the reference index. The results were indicated that the fat peroxide value of all samples increased within the storage of 100d, and then decreased. Aluminum film compound packaging showed a better effect than polyethylene plastic bag. The antioxideant combination of 0.5% tea-polyphenol, 0.5% Vc, 0.5% citric acid, 5% sodium alginate, applied on 4 kGy irradiated samples was measured the lowest peroxide value of fat among all the treatments. (authors)

  6. Investigations on neutron irradiated 3D carbon fibre reinforced carbon composite material

    Science.gov (United States)

    Venugopalan, Ramani; Alur, V. D.; Patra, A. K.; Acharya, R.; Srivastava, D.

    2018-04-01

    As against conventional graphite materials carbon-carbon (C/C) composite materials are now being contemplated as the promising candidate materials for the high temperature and fusion reactor owing to their high thermal conductivity and high thermal resistance, better mechanical/thermal properties and irradiation stability. The current need is for focused research on novel carbon materials for future new generation nuclear reactors. The advantage of carbon-carbon composite is that the microstructure and the properties can be tailor made. The present study encompasses the irradiation of 3D carbon composite prepared by reinforcement using PAN carbon fibers for nuclear application. The carbon fiber reinforced composite was subjected to neutron irradiation in the research reactor DHRUVA. The irradiated samples were characterized by Differential Scanning Calorimetry (DSC), small angle neutron scattering (SANS), XRD and Raman spectroscopy. The DSC scans were taken in argon atmosphere under a linear heating program. The scanning was carried out at temperature range from 30 °C to 700 °C at different heating rates in argon atmosphere along with reference as unirradiated carbon composite. The Wigner energy spectrum of irradiated composite showed two peaks corresponding to 200 °C and 600 °C. The stored energy data for the samples were in the range 110-170 J/g for temperature ranging from 30 °C to 700 °C. The Wigner energy spectrum of irradiated carbon composite did not indicate spontaneous temperature rise during thermal annealing. Small angle neutron scattering (SANS) experiments have been carried out to investigate neutron irradiation induced changes in porosity of the composite samples. SANS data were recorded in the scattering wave vector range of 0.17 nm-1 to 3.5 nm-1. Comparison of SANS profiles of irradiated and unirradiated samples indicates significant change in pore morphology. Pore size distributions of the samples follow power law size distribution with

  7. IFMIF - International Fusion Materials Irradiation Facility Conceptual Design Activity/Interim Report

    International Nuclear Information System (INIS)

    Rennich, M.J.

    1995-12-01

    Environmental acceptability, safety, and economic viability win ultimately be the keys to the widespread introduction of fusion power. This will entail the development of radiation- resistant and low- activation materials. These low-activation materials must also survive exposure to damage from neutrons having an energy spectrum peaked near 14 MeV with annual radiation doses in the range of 20 displacements per atom (dpa). Testing of candidate materials, therefore, requires a high-flux source of high energy neutrons. The problem is that there is currently no high-flux source of neutrons in the energy range above a few MeV. The goal, is therefore, to provide an irradiation facility for use by fusion material scientists in the search for low-activation and damage-resistant materials. An accellerator-based neutron source has been established through a number of international studies and workshops' as an essential step for materials development and testing. The mission of the International Fusion Materials Irradiation Facility (IFMIF) is to provide an accelerator-based, deuterium-lithium (D-Li) neutron source to produce high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials up to about a full lifetime of anticipated use in fusion energy reactors. would also provide calibration and validation of data from fission reactor and other accelerator-based irradiation tests. It would generate material- specific activation and radiological properties data, and support the analysis of materials for use in safety, maintenance, recycling, decommissioning, and waste disposal systems

  8. Remarks on the mathematical description of materials damage by irradiation

    International Nuclear Information System (INIS)

    Steinbach, E.

    1987-01-01

    In describing radiation damage processes in materials, the chemical rate theory proves to be the most effective mathematical method. The applicability, and even the validity, of this theory, which has been successfully applied in many other scientific fields, have been questioned by some authors. After rigorous mathematical analysis of the relevant rate equations and the corresponding self-consistent calculation of sink strengths, the main criticisms on this subject can be disproved

  9. Experiment and research on materials irradiated by plasma radiation

    International Nuclear Information System (INIS)

    Hong Wenyu; Yao Lianghua; Tang Sujun; Chang Shufen; Li Guodong

    1992-08-01

    The TiC and SiC coating on the graphite substrate and wall carbonization were studied by plasma radiation in HL-1 tokamak. Samples were analysed with AES (auger electron spectroscopy), SEM (scanning electron microscopy), XPS (X-ray photoelectron spectroscopy) and XDS (X-ray diffraction spectroscopy). The results show that the TiC and SiC materials coated on limiter and wall and wall carbonization can reduce the metal and oxygen impurities and improve the plasma merit

  10. Effect of low temperature neutron irradiation on the magnetoresistivity in stabilizer materials for a superconducting magnet

    International Nuclear Information System (INIS)

    Nakata, Kiyotomo; Tada, Naobumi; Masaoka, Isao; Takamura, Saburo.

    1985-01-01

    Magnetoresistivity changes caused by neutron irradiation at 5 K, annealing up to 300 K and cyclic irradiation are studied in copper and aluminuim stabilizer materials at 4.2 K. The radiation-induced resistivity in Al is about three times as large as that in Cu, and the resistivities in both Al and Cu are independent of the purity and the degree of cold-work of the samples. The radiation-induced magnetoresistivity of the high purity Cu with R.R.R. (R sub(298 K)/R sub(4.2 K)) of 1400 is larger than that of the impure Cu with R.R.R. of 300 and 280. The magnetoresistivities of the high purity Cu and Al with R.R.R. of 1500 increase with the magetic field. Magnetoresistivity change with the magnetic field in the irradiated Cu mostly follows Kohler's rule, and that in the irradiated Al does not follow the rule at high magnetic fields. By the annealing at 300 K after the irradiation, the radiation-induced resistivity is completely annihilated in the Al, but about 20 % of the resistivity retains in the full-annealed Cu and the retained resistivity is accumulated during the cyclic irradiation. Though the accumulated resistivity in the cold-worked Cu is smaller than that in the full-annealed one, the resistivity before irradiation in the cold-worked samples is very large. From the above results, the full-annealed Cu with R.R.R. of about 300 is considered to be the best material as a stabilizer used under irradiation. (author)

  11. Whole-Genome Sequencing in Microbial Forensic Analysis of Gamma-Irradiated Microbial Materials.

    Science.gov (United States)

    Broomall, Stacey M; Ait Ichou, Mohamed; Krepps, Michael D; Johnsky, Lauren A; Karavis, Mark A; Hubbard, Kyle S; Insalaco, Joseph M; Betters, Janet L; Redmond, Brady W; Rivers, Bryan A; Liem, Alvin T; Hill, Jessica M; Fochler, Edward T; Roth, Pierce A; Rosenzweig, C Nicole; Skowronski, Evan W; Gibbons, Henry S

    2016-01-15

    Effective microbial forensic analysis of materials used in a potential biological attack requires robust methods of morphological and genetic characterization of the attack materials in order to enable the attribution of the materials to potential sources and to exclude other potential sources. The genetic homogeneity and potential intersample variability of many of the category A to C bioterrorism agents offer a particular challenge to the generation of attributive signatures, potentially requiring whole-genome or proteomic approaches to be utilized. Currently, irradiation of mail is standard practice at several government facilities judged to be at particularly high risk. Thus, initial forensic signatures would need to be recovered from inactivated (nonviable) material. In the study described in this report, we determined the effects of high-dose gamma irradiation on forensic markers of bacterial biothreat agent surrogate organisms with a particular emphasis on the suitability of genomic DNA (gDNA) recovered from such sources as a template for whole-genome analysis. While irradiation of spores and vegetative cells affected the retention of Gram and spore stains and sheared gDNA into small fragments, we found that irradiated material could be utilized to generate accurate whole-genome sequence data on the Illumina and Roche 454 sequencing platforms. Copyright © 2016, American Society for Microbiology. All Rights Reserved.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  13. Long-term radiation effects on commercial cable-insulating materials irradiated at CERN

    International Nuclear Information System (INIS)

    Maier, P.; Stolarz, A.

    1983-01-01

    Long-term irradiation damage tests have been carried out on a variety of flexible cable-insulating materials offered to CERN by different European cable manufacturers. Tensile test specimens were exposed for a maximum of three years in high-level radiation areas of the Super Proton Synchrotron (SPS) and for comparison at high dose rates in a nuclear reactor. The degradation of mechanical properties after irradiation in air depends not only on the total absorbed dose, but also on the dose rate for most of these polymer compounds. These dose-rate effects vary between material types and for different compounds. The results presented here illustrate the difference in radiation damage between short-term and long-term irradiation conditions in a typical service application for the various materials tested. They also allow safety factors to be estimated for the extrapolation of the limiting exposure in service from accelerated material tests in the range of dose rates covered. A discussion of the available models of the dose-rate effects results in a conservative estimate for extrapolation to low dose rates from measured values at intermediate dose rates of the order of 0.1 Gy/s. Based on short-term irradiation tests only, the safety factors to be applied depend on the end-point criterion used, and may vary between 1 and 10 for the range of dose rates and materials considered here. (orig.)

  14. 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.)

  15. Nano-pulsed laser irradiation scanning system for phase-change materials

    International Nuclear Information System (INIS)

    Kim, Sookyung; Li Xuezhe; Lee, Sangbin; Kim, Kyung-Ho; Lee, Seung-Yop

    2008-01-01

    Recently, the demand of a laser irradiation tester is increasing for phase change random access memory (PRAM) as well as conventional optical storage media. In this study, a nano-pulsed laser irradiation system is developed to characterize the optical property and writing performance of phase-change materials, based on a commercially available digital versatile disk (DVD) optical pick-up. The precisely controlled focusing and scanning on the material's surface are implemented using the auto-focusing mechanism and a voice coil motor (VCM) of the commercial DVD pick-up. The laser irradiation system provides various writing and reading functions such as adjustable laser power, pulse duration, recording pattern (spot, line and area), and writing/reading repetition, phase transition, and in situ reflectivity measurement before/after irradiation. Measurements of power time effect (PTE) diagram and reflectivity map of Ge 2 Sb 2 Te 5 samples show that the proposed laser irradiation system provides the powerful scanning tool to quantify the optical characteristics of phase-change materials

  16. IFMIF, a fusion relevant neutron source for material irradiation current status

    International Nuclear Information System (INIS)

    Knaster, J.; Chel, S.; Fischer, U.; Groeschel, F.; Heidinger, R.; Ibarra, A.; Micciche, G.; Möslang, A.; Sugimoto, M.; Wakai, E.

    2014-01-01

    The d-Li based International Fusion Materials Irradiation Facility (IFMIF) will provide a high neutron intensity neutron source with a suitable neutron spectrum to fulfil the requirements for testing and qualifying fusion materials under fusion reactor relevant irradiation conditions. The IFMIF project, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the Broader Approach (BA) Agreement between Japan Government and EURATOM, aims at the construction and testing of the most challenging facility sub-systems, such as the first accelerator stage, the Li target and loop, and irradiation test modules, as well as the design of the entire facility, thus to be ready for the IFMIF construction with a clear understanding of schedule and cost at the termination of the BA mid-2017. The paper reviews the IFMIF facility and its principles, and reports on the status of the EVEDA activities and achievements

  17. Damage functions generation for polyatomic materials irradiated in test reactors

    International Nuclear Information System (INIS)

    Alberman, A.; Lesueur, D.

    1987-06-01

    Neutron exposure parameters in polyatomic materials is of great importance for fusion technology programs. The COMPOSI code computes the number of displaced atoms of sub-lattice ''j'' induced by one atom of sub-lattice ''i'' either by direct collision or through intermediate knocked atom. The code uses Lindhard equations; it is solved by iterative process. The atomic displacements cross-sections, as a function of neutron energy are derived by folding previous results with ''i'' type PKA. Moreover the COMPOSI code may include recoils from charged particles e.g.: Alpha + Triton from Li 6 capture in Li Al 0 2 . These responses in various spectra are discussed [fr

  18. Analysis of irradiated materials in Ul-chin unit 5

    International Nuclear Information System (INIS)

    Jung, Y. H.; Yoo, B. O.; Kim, H. M.; Joo, Y. S.

    2007-02-01

    The microstructure examination, the fracture surface observation, the composition analysis and the micro-hardness measurement were carried out for investigation of debris apart from structure in Ul-chin uint 5. As the results of investigation, those of debris were found out screw bolts and the washer. The screw bolts and the washer were coincident with materials from ASTM A-193 by quantitative analysis. The screw bolts and the washer were made by STS 304. Finally, all of screw bolts were parts of the LPSI pump case even though one of them was found in different place. The washer was part of the heat exchanger

  19. Similar Survival for Patients Undergoing Reduced-Intensity Total Body Irradiation (TBI) Versus Myeloablative TBI as Conditioning for Allogeneic Transplant in Acute Leukemia

    Energy Technology Data Exchange (ETDEWEB)

    Mikell, John L., E-mail: jmikell@emory.edu [Department of Radiation Oncology, Winship Cancer Institute, Emory University, Atlanta, Georgia (United States); Waller, Edmund K. [Department of Hematology and Oncology, Winship Cancer Institute, Emory University, Atlanta, Georgia (United States); Switchenko, Jeffrey M. [Department of Biostatistics and Bioinformatics, Winship Cancer Institute, Emory University, Atlanta, Georgia (United States); Rangaraju, Sravanti; Ali, Zahir; Graiser, Michael [Department of Hematology and Oncology, Winship Cancer Institute, Emory University, Atlanta, Georgia (United States); Hall, William A. [Department of Radiation Oncology, Winship Cancer Institute, Emory University, Atlanta, Georgia (United States); Langston, Amelia A. [Department of Hematology and Oncology, Winship Cancer Institute, Emory University, Atlanta, Georgia (United States); Esiashvili, Natia [Department of Radiation Oncology, Winship Cancer Institute, Emory University, Atlanta, Georgia (United States); Khoury, H. Jean [Department of Hematology and Oncology, Winship Cancer Institute, Emory University, Atlanta, Georgia (United States); Khan, Mohammad K. [Department of Radiation Oncology, Winship Cancer Institute, Emory University, Atlanta, Georgia (United States)

    2014-06-01

    Purpose: Hematopoietic stem cell transplantation (HSCT) is the mainstay of treatment for adults with acute leukemia. Total body irradiation (TBI) remains an important part of the conditioning regimen for HCST. For those patients unable to tolerate myeloablative TBI (mTBI), reduced intensity TBI (riTBI) is commonly used. In this study we compared outcomes of patients undergoing mTBI with those of patients undergoing riTBI in our institution. Methods and Materials: We performed a retrospective review of all patients with acute leukemia who underwent TBI-based conditioning, using a prospectively acquired database of HSCT patients treated at our institution. Patient data including details of the transplantation procedure, disease status, Karnofsky performance status (KPS), response rates, toxicity, survival time, and time to progression were extracted. Patient outcomes for various radiation therapy regimens were examined. Descriptive statistical analysis was performed. Results: Between June 1985 and July 2012, 226 patients with acute leukemia underwent TBI as conditioning for HSCT. Of those patients, 180 had full radiation therapy data available; 83 had acute lymphoblastic leukemia and 94 had acute myelogenous leukemia; 45 patients received riTBI, and 135 received mTBI. Median overall survival (OS) was 13.7 months. Median relapse-free survival (RFS) for all patients was 10.2 months. Controlling for age, sex, KPS, disease status, and diagnosis, there were no significant differences in OS or RFS between patients who underwent riTBI and those who underwent mTBI (P=.402, P=.499, respectively). Median length of hospital stay was shorter for patients who received riTBI than for those who received mTBI (16 days vs 23 days, respectively; P<.001), and intensive care unit admissions were less frequent following riTBI than mTBI (2.22% vs 12.69%, respectively, P=.043). Nonrelapse survival rates were also similar (P=.186). Conclusions: No differences in OS or RFS were seen between

  20. Two micro fatigue test methods for irradiated materials

    International Nuclear Information System (INIS)

    Nunomura, Shigetomo; Noguchi, Shinji; Okamura, Yuichi; Kumai, Shinji

    1993-01-01

    This paper demonstrates two miniature fatigue test methods in response to the requirements of the fusion reactor wall materials development program. It is known that the fatigue strength evaluated by the axial loading test is independent of the specimen size, while that evaluated by the bend test or torsion test is dependent upon the size of specimen. The new type of gripping system for the axial, tension-tension, fatigue testing of TEM disk-size specimens that has been developed is described in this paper. An alignment tool assists in gripping the miniature specimen. The miniature tension-tension fatigue test method seems to provide reliable S-N curves for SUS304 and SUS316L stainless steels. An indentation method has also been developed to determine fatigue properties. A hard steel ball or ceramic ball was used for cyclically loading the specimen, and an S-N curve was subsequently obtained. The merit of this method is primarily simple handling. S-N curves obtained from four materials by this indentation method compared well with those obtained from the rotary bend fatigue test employing a standard-size specimen

  1. Study of a design criterion for 316L irradiated represented by a strain hardened material

    International Nuclear Information System (INIS)

    Gouin, H.

    1999-01-01

    The aim of this study is to analyse the consequence of radiation on different structure submitted to imposed displacement loading and for damages due to plastic instability or rupture. The main consequence of radiation is a material hardening with a ductility decrease. This effect is similar to initial mechanical hardening: the mechanical properties (determined on smooth tensile specimen) evolve in the same way while irradiation or mechanical hardening increase. So in this study, radiation hardening is simulated by mechanical hardening (swaging). Tests were carried out for which two damages were considered: plastic instability and rupture. These two damages were studied with initial mechanical hardening (5 tested hammering rate 0, 15, 25, 35 and 45% on 316L stainless steel). Likewise two types of loading were studied: tensile or bending loading on specimens with or without geometrical singularities (notches). From tensile tests, two deformation criteria are proposed for prevention against the two quoted damages. Numerical study is carried out allowing to confirm hypothesis made at the time of the tensile test result interpretation and to validate the rupture criterion by applying on bending test. (author)

  2. Nuclear data needs for neutron spectrum tailoring at International Fusion Materials Irradiation Facility (IFMIF)

    International Nuclear Information System (INIS)

    Sugimoto, Masayoshi

    2001-01-01

    International Fusion Materials Irradiation Facility (IFMIF) is a proposal of D-Li intense neutron source to cover all aspects of the fusion materials development in the framework of IEA collaboration. The new activity has been started to qualifying the important technical issues called Key Element technology Phase since 2000. Although the neutron spectrum can be adjusted by changing the incident beam energy, it is favorable to be carried out many irradiation tasks at the same time under the unique beam condition. For designing the tailored neutron spectrum, neutron nuclear data for the moderator-reflector materials up to 50 MeV are required. The data for estimating the induced radioactivity is also required to keep the radiation level low enough at maintenance time. The candidate materials and the required accuracy of nuclear data are summarized. (author)

  3. Nuclear data needs for neutron spectrum tailoring at International Fusion Materials Irradiation Facility (IFMIF)

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    International Fusion Materials Irradiation Facility (IFMIF) is a proposal of D-Li intense neutron source to cover all aspects of the fusion materials development in the framework of IEA collaboration. The new activity has been started to qualifying the important technical issues called Key Element technology Phase since 2000. Although the neutron spectrum can be adjusted by changing the incident beam energy, it is favorable to be carried out many irradiation tasks at the same time under the unique beam condition. For designing the tailored neutron spectrum, neutron nuclear data for the moderator-reflector materials up to 50 MeV are required. The data for estimating the induced radioactivity is also required to keep the radiation level low enough at maintenance time. The candidate materials and the required accuracy of nuclear data are summarized. (author)

  4. Behavior of structural and target materials irradiated in spallation neutron environments

    International Nuclear Information System (INIS)

    Stubbins, J.F.; Wechsler, M.; Borden, M.

    1995-01-01

    This paper describes considerations for selection of structural and target materials for accelerator-driven neutron sources. Due to the operating constraints of proposed accelerator-driven neutron sources, the criteria for selection are different than those commonly applied to fission and fusion systems. Established irradiation performance of various alloy systems is taken into account in the selection criteria. Nevertheless, only limited materials performance data are available which specifically related to neutron energy spectra anticipated for spallation sources

  5. Bulk-shield design for the Fusion Materials Irradiation Test facility

    International Nuclear Information System (INIS)

    Carter, L.L.; Mann, F.M.; Morford, R.J.; Johnson, D.L.; Huang, S.T.

    1982-07-01

    The accelerator-based Fusion Materials Irradiation Test (FMIT) facility will provide a high-fluence, fusion-like radiation environment for the testing of materials. While the neutron spectrum produced in the forward direction by the 35 MeV deuterons incident upon a flowing lithium target is characterized by a broad peak around 14 MeV, a high energy tail extends up to about 50 MeV. Some shield design considerations are reviewed

  6. High flux materials testing reactor HFR Petten. Characteristics of facilities and standard irradiation devices

    International Nuclear Information System (INIS)

    Roettger, H.; Hardt, P. von der; Tas, A.; Voorbraak, W.P.

    1981-01-01

    For the materials testing reactor HFR some characteristic information is presented. Besides the nuclear data for the experiment positions short descriptions are given of the most important standard facilities for material irradiation and radionuclide production. One paragraph deals with the experimental set-ups for solid state and nuclear structure investigations. The information in this report refers to a core type, which is operational since March 1977. The numerical data compiled have been up-dated to January 1981

  7. Behavior of structural and target materials irradiated in spallation neutron environments

    Energy Technology Data Exchange (ETDEWEB)

    Stubbins, J.F. [Univ. of Illinois, Urbana, IL (United States); Wechsler, M. [North Carolina State Univ., Raleigh, NC (United States); Borden, M. [Los Alamos National Laboratory, NM (United States)] [and others

    1995-10-01

    This paper describes considerations for selection of structural and target materials for accelerator-driven neutron sources. Due to the operating constraints of proposed accelerator-driven neutron sources, the criteria for selection are different than those commonly applied to fission and fusion systems. Established irradiation performance of various alloy systems is taken into account in the selection criteria. Nevertheless, only limited materials performance data are available which specifically related to neutron energy spectra anticipated for spallation sources.

  8. International fusion materials irradiation facility and neutronic calculations for its test modules

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.

    1997-01-01

    The International Fusion Material Irradiation Facility (IFMIF) is a projected high intensity neutron source for material testing. Neutron transport calculations for the IFMIF project are performed for variety of here explained reasons. The results of MCNP neutronic calculations for IFMIF test modules with NaK and He cooled high flux test cells are presented in this paper. (author). 3 refs., 2 figs., 3 tabs

  9. Effect of material variables on the irradiation performance of boron carbide

    International Nuclear Information System (INIS)

    Basmajian, J.A.; Hollenberg, G.W.

    1980-01-01

    Boron carbide pellets were fabricated with variations in material parameters. These pellets were irradiated in the Experimental Breeder Reactor-II (EBR-II) to determine the effect of these variations on the performance. Helium release from the material and swelling of the pellets are the primary measures of performance. It was determined that material with a smaller grain size released more helium and swelled less. The pellets with boron-to-carbon ratios greater than 4 to 1 did not perform well. Iron additions improved the performance of the material while density variations had little effect

  10. Tensile properties of irradiated and fatigue exposed stainless steel DIN X 6 CrNi 1811 (similar to AISI type 304) plate and welded joints

    International Nuclear Information System (INIS)

    Vries, M.I. de; Schaaf, B. van der; Elen, J.D.

    1979-10-01

    Test specimens of plate metal and welded joints of stainless steel DIN 1.4948, which is similar to AISI type 304, have been irradiated at 723 K and 823 K up to fluences of 1.10 23 n.m -2 and 5.10 24 n.m -2 (E > 0.1 MeV). These are representative conditions for the SNR-300 reactor vessel and inner components after 16 years of operation. High-rate (depsilon/dt = 1 s -1 ) tensile tests were performed after fatigue exposure up to various fractions of fatigue life (D) ranging from 5% to 95% at the same temperatures as the nominal temperatures of the irradiation series

  11. Data on post irradiation experiments of heat resistant ceramic composite materials. PIE for 97M-13A

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Shin-ichi; Ishihara, Masahiro; Souzawa, Shizuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Sekino, Hajime [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The research on the radiation damage mechanism of heat resistant ceramic composite materials is one of the research subjects of the innovative basic research in the field of high temperature engineering, using the High Temperature engineering Test Reactor (HTTR). Three series of irradiation tests on the heat resistant ceramic composite materials, first to third irradiation test program, were carried out using the Japan Material Testing Reactor (JMTR). This is a summary report on the first irradiation test program; irradiation induced dimensional change, thermal expansion coefficient, X-ray diffraction and {gamma}-ray spectrum are reported. (author)

  12. Change in properties of superconducting magnet materials by fusion neutron irradiation

    International Nuclear Information System (INIS)

    Nishimura, Arata; Nishijima, Shigehiro; Takeuchi, Takao; Nishitani, Takeo

    2007-01-01

    A fusion reactor will generate a lot of high energy neutron and much energy will be taken out of the neutrons by a blanket system. Since some neutrons will stream out of a plasma vacuum vessel through neutral beam injection ports and penetrate a blanket system, a superconducting magnet system, which provides high magnetic field to confirm high energy particles, will be irradiated by a certain amount of neutrons. By developing the new NBI system or by reducing the penetration, the neutron fluence to the superconducting magnet will be able to be reduced. However, it is not easy to achieve the lower streaming and penetration at the present. Therefore, investigations on irradiation behavior of superconducting magnet materials are desired and some novel researches have been performed from 1970s. In general, the critical current of the superconducting wire increases under fast neutron environment comparing with that of the non-irradiated wire, and then decreased to almost zero as an increase of neutron fluence. On the other hand, the critical temperature of the wire starts to get down around 10 22 n/m 2 of neutron fluence and the temperature margin will be decreased during the operation by the neutron irradiation. In this paper, some aspects of irradiated materials will be overviewed and general tendency will be discussed focussing on knock-on effect of fast neutron and long range ordering of A15 compounds

  13. Influence of specimen size/type on the fracture toughness of five irradiated RPV materials

    International Nuclear Information System (INIS)

    Sokolov, Mikhail A; Lucon, Enrico

    2015-01-01

    The Heavy-Section Steel Irradiation (HSSI) Program had previously irradiated five reactor pressure vessel (RPV) steels/welds at fast neutron fluxes of about 4 to 8 x 10 11 n/cm 2 /s (>1 MeV) to fluences from 0.5 to 3.4 10 19 n/cm 2 and at 288 °C. The unirradiated fracture toughness tests were performed by Oak Ridge National Laboratory with 12.7-mm and 25.4-mm thick (0.5T and 1T) compact specimens, while the HSSI Program provided tensile and 5 x 10-mm three-point bend specimens to SCK-CEN for irradiation in the in-pile section of the Belgian Reactor BR2 at fluxes > 10 13 n/cm 2 /s and subsequent testing by SCK-CEN. The BR2 irradiations were conducted at about 2 and 4 x 10 13 n/cm 2 /s with irradiation temperature between 295 °C and 300 °C (water temperature), and to fluences between 6 and 10 x 10 19 n/cm 2 . The irradiation-induced shifts of the Master Curve reference temperatures, ΔT 0 , for most of the materials deviated from the embrittlement correlations much more than expected, motivating the testing of 5 x 10-mm three-point bend specimens of all five materials in the unirradiated condition to eliminate specimen size and geometry as a variable. Tests of the unirradiated small bend specimens resulted in Master Curve reference temperatures, ΔT 0 , 25 °C to 53 °C lower than those from the larger compact specimens, meaning that the irradiation-induced reference temperature shifts, ΔT 0 , were larger than the initial measurements, resulting in much improved agreement between the measured and predicted fracture toughness shifts.

  14. The development of fuel pins and material specimens mixed loading irradiation test rig in the experimental fast reactor Joyo. The development of the fuel-material hybrid rig

    International Nuclear Information System (INIS)

    Oyamatsu, Yasuko; Someya, Hiroyuki

    2013-02-01

    In the experimental fast reactor Joyo, there were many tests using the irradiation rigs that it was possible to be set irradiation conditions for each compartment independently. In case of no alternative fuel element to irradiate after unloading the irradiated compartments, the irradiation test was restarted with the dummy compartment which the fuel elements was not mounted. If the material specimens are mounted in this space, it is possible to use the irradiation space effectively. For these reasons, the irradiation rig (hybrid rig) is developed that is consolidated with material specimens compartment and fuel elements compartment. Fuel elements and material specimens differ greatly with heat generation, so that the most important issue in developing of hybrid rig is being able to distribute appropriately the coolant flow which satisfies irradiation conditions. The following is described by this report. (1) It was confirmed that the flow distribution of loading the same irradiation rig with the compartment from which a flow demand differs could be satisfied. (2) It was confirmed that temperature setting range of hybrid rig could be equivalent to that of irradiation condition. (3) By standardizing the coolant entrance structure of the compartment lower part, the prospect which can perform easily recombination of the compartment from which a type differs between irradiation rigs was acquired. (author)

  15. INAA study of Hg, Se, As, and Br irradiation losses from l-cysteine treated and untreated reference materials

    International Nuclear Information System (INIS)

    Anderson, D.L.

    2013-01-01

    U. S. Food and Drug Administration in-house reference material (RM) Cocoa Powder and National Institute of Standards and Technology Standard RMs (SRMs) 1515 apple leaves, 1547 peach leaves, 1571 orchard leaves, 1566a oyster tissue, and 1568a rice flour were co-irradiated together with polyethylene blanks and analyzed for Hg and Se by anticoincidence instrumental neutron activation analysis. The three botanical SRM portions showed a combined Hg recovery of 70 % while the other portions showed a combined Hg recovery of 169 %, indicating that volatile Hg was lost from botanical SRMs and absorbed by the other irradiated portions. Total Hg recovery for all portions was 82 %. Se results showed no evidence of cross-contamination and all results agreed with certified and known values. National Research Council of Canada Certified RMs DOLT-3 dogfish liver, TORT-2 lobster hepatopancreas, and DORM-3 fish protein were separately analyzed either with no treatment or after treatment with l-cysteine solutions followed by drying over magnesium perchlorate. Each set of portions was co-irradiated with polyethylene and treated filter blanks. Analysis of all components of each treated portion irradiation package showed that essentially all Hg was retained within the package. Treated DOLT-3 portions (inorganic Hg content 53 %) showed a tenfold improvement with 99 % Hg retention. Hg retention for DORM-3 (7 % inorganic Hg) was 85 % (a twofold improvement) while retention for TORT-2 (44 % inorganic Hg), was 94 %, similar to that for untreated portions (96 %). Small irradiation losses (≤0.5 %) of volatile species of Se, As, and Br were observed. (author)

  16. Effect of using type A radiation for dose reconstruction in type B irradiated material: A microdosimetry approach

    International Nuclear Information System (INIS)

    Piters, T.M.; Chernov, V.

    2008-01-01

    A model is proposed to explain that in previously γ irradiated calcite, the yield after additive β irradiation tends to incline to the saturation yield of the β radiation even if that yield is lower than the yield after the γ irradiation. However, the proposed model is not specific for calcite and in fact all calculations are done in a fictive material. The proposed model considers, in contrast to existing models, the track nature of γ and β radiations and that these different types of radiations can be distinguished by the dose distribution inside their tracks. The determination of the dose distribution in the tracks for the different types of irradiations is quite complicated and instead we approximate the γ and β tracks by type A and B tracks that have different but homogeneously distributed dose in their track volumes. The trapping of generated free charges in the track was calculated with a simple one electron-one hole trap model. To obtain the total dose response (the average concentration of occupied traps as a function of dose), the yield in one point was averaged over all possible configurations of track overlapping in that point. We determined the slope of the initial part of the response curve (low dose sensitivity) and the saturation yield as function of the track dose. It is observed that the low dose sensitivity and saturation yield both decrease with increasing track dose. Simulations of the response to sequential irradiation first by type A radiation with a 64 Gy track dose and then followed by type B radiation with a track dose of 128 Gy using our model show a similar effect as observed in calcite demonstrating that the track nature of radiation is a plausible cause for the observed effect

  17. Charged particle and laser irradiation of selected materials

    International Nuclear Information System (INIS)

    Svendsen, W.E.

    1996-11-01

    The main topics of the present thesis are the processes governing electronic sputtering of insulators and laser ablation of metals and insulators. The sputtering yield for electron bombardment of solid deuterium was investigated using quartz crystal microbalances as the measuring technique. The sputtering yield was measured with varying electron energy and deuterium film thickness. Laser ablation measurements of silver and nickel were carried out using a Nd:YAG laser. The effect of various experimental parameters such as background gas pressure (Ar, N 2 ), position of quartz crystals with respect to target position and the optimal number of laser shots for carrying out the experiments were investigated. The deposition rate was measured with varying laser wavelength and laser fluence. The angular distribution of the ablated material was measured for silver as well. A theoretical model based on the thermal properties of laser interaction with metals was applied in the initial phase of ablation. For the non-thermal processes governing laser interaction with the ablated plasma plume, a model developed by Phipps and Dreyfus was used to interpret the results. Laser ablation measurements of water-ice were carried using a Nitrogen laser. Attempts were made to measure the deposition rate for various the laser wavelengths and energies. (au) 8 tabs., 49 ills., 77 refs

  18. Charged particle and laser irradiation of selected materials

    Energy Technology Data Exchange (ETDEWEB)

    Svendsen, W E

    1996-11-01

    The main topics of the present thesis are the processes governing electronic sputtering of insulators and laser ablation of metals and insulators. The sputtering yield for electron bombardment of solid deuterium was investigated using quartz crystal microbalances as the measuring technique. The sputtering yield was measured with varying electron energy and deuterium film thickness. Laser ablation measurements of silver and nickel were carried out using a Nd:YAG laser. The effect of various experimental parameters such as background gas pressure (Ar, N{sub 2}), position of quartz crystals with respect to target position and the optimal number of laser shots for carrying out the experiments were investigated. The deposition rate was measured with varying laser wavelength and laser fluence. The angular distribution of the ablated material was measured for silver as well. A theoretical model based on the thermal properties of laser interaction with metals was applied in the initial phase of ablation. For the non-thermal processes governing laser interaction with the ablated plasma plume, a model developed by Phipps and Dreyfus was used to interpret the results. Laser ablation measurements of water-ice were carried using a Nitrogen laser. Attempts were made to measure the deposition rate for various the laser wavelengths and energies. (au) 8 tabs., 49 ills., 77 refs.

  19. SPECTER-ANL, Neutron Damage for Material Irradiation

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of program or function: SPECTER calculates spectral- averaged displacements, recoil spectra, gas production, and total damage energy (Kerma) for 41 pure elements using ENDF/B-V derived cross sections. The user need only specify a neutron energy spectrum. Because SPECTER does not handle compounds, SPECOMP was developed to determine displacement damage for alloys, insulators, and breeder materials. 2 - Method of solution: In SPECTER elastic scattering is treated exactly including angular distributions from ENDF/B-V. Inelastic scattering calculations consider both discrete and continuous nuclear level distributions. Multiple (n,xn) reactions use a Monte Carlo technique to derive the recoil distributions. The (n,d) and (n,t) reactions are treated as (n,p) and (n, 3 He) as (n, 4 He). The neutron-gamma reaction and subsequent beta-decay are also included, using a new treatment of gamma-gamma coincidences, angular correlations, beta-neutrino correlations and the incident neutron energy. The Lindhard model was used to compute the energy available for nuclear displacement at each recoil energy. SPECOMP reads the required files from SPECTER, computes secondary displacement functions for each combination of recoil and matrix atom, and then integrates over recoil energy to find the net displacement cross section at each neutron energy. Damage due to neutron, gamma-ray and beta decay events is then added in and the results are summed to obtain the total dpa cross section. 3 - Restrictions on the complexity of the problem: The DISCS computer code was used to process ENDF/B-V data for 41 pure elements for use with SPECTER-ANL. SPECOMP can use any combination of four elements in a single run

  20. Safety design of the international fusion materials irradiation facility (IFMIF)

    International Nuclear Information System (INIS)

    Konishi, Satoshi; Yamaki, Daiju; Katsuta, Hiroji; Moeslang, Anton; Jameson, R.A.; Martone, Marcello; Shannon, T.E.

    1997-11-01

    In the Conceptual Design Activity of the IFMIF, major subsystems, as well as the entire facility is carefully designed to satisfy the safety requirements for any possible construction sites. Each subsystem is qualitatively analyzed to identify possible hazards to the workers, public and environments using Failure Mode and Effect Analysis (FMEA). The results are reflected in the design and operation procedure. Shielding of radiation, particularly neutron around the test cell is one of the most important issue in normal operation. Radiation due to beam halo and activation is a hazard for operation personnel in the accelerator system. For the maintenance, remote handling technology is designed to be applied in various facilities of the IFMIF. Lithium loop and target system hold the majority of the radioactive material in the facility. Tritium and beryllium-7 are generated by the nuclear reaction during operation and thus needed to be removed continuously. They are also the potential hazards of airborne source in off-normal events. Minimization of inventory, separation and immobilization, and multiple confinement are considered in the design. Generation of radioactive waste is anticipated to be minor, but waste treatment systems for gas, liquid and solid wastes are designed to minimize the environmental impact. Lithium leak followed by a fire is a major concern, and extensive prevention plan is made in the target design. One of the design option considered is composed of; primary enclosure of the lithium loop, secondary containment filled with positive pressure argon, and an air tight lithium cell made of concrete with a steel lining. This study will report some technical issues considered in the design of IFMIF. It was concluded that the IFMIF can be designed and constructed to meet or exceed current safely standards for workers, public and the environment with existing technology and reasonable construction cost. (J.P.N.)

  1. Study of the temperature dependence of the uniaxial creep property of similar material of new soft rock

    Science.gov (United States)

    Wang, Y. Y.; Wu, Y.; Fan, X. Y.; Zhang, J. L.; Guo, P.; Li, J. G.

    2017-11-01

    Using the experimental method, the experimental research of creep properties were conducted under different temperature ranging from 10°C to 60°C. The similar material of new soft rock consists of paraffin, which can obtain that the deformation contains the instantaneous elastic deformation and creep deformation through the uniaxial creep experimental results. And thus the increase of temperature has great influence on the creep characteristics of similar soft rock according to the creep curve of similar soft rock at 10°C to 60°C. With the increase of temperature, the slope of the stress-strain curve of similar soft rock is increasing, while the average of the creep modulus is decreasing, which means that the capacity of resist deformation is reduced. Therefore, the creeps law of high-temperature and short-time can be shown the creep phenomenon of low-temperature and long-time, and further shorten the creep experimental cycle.

  2. Effect of gamma rays on crystalline materials during irradiation in a reactor

    International Nuclear Information System (INIS)

    Nikolaenko, V.A.; Karpukhin, V.I.; Gordeev, V.G.

    1995-01-01

    The article presents and discusses the results of experiments to determine the effect of gamma rays on the change in the properties of diamond, graphite, and structural steel. The materials were irradiated in a VVER type reactor. For diamonds, the effect on the annealing of defects was investigated. As gamma ray intensity increased, the crystal lattice expansion and defect concentration increased. Graphite lattice expansion and the mechanical properties of structural steel were also examined. Graphite lattice expansion increased with increased neutron flux and decreased irradiation temperature. Changes in the impact toughness of structural steel correlated precisely to the gamma ray flux in the experiments. 6 refs., 3 figs

  3. Comparison of pad detectors produced on different silicon materials after irradiation with neutrons, protons and pions

    International Nuclear Information System (INIS)

    Kramberger, G.; Cindro, V.; Dolenc, I.; Mandic, I.; Mikuz, M.; Zavrtanik, M.

    2010-01-01

    A set of 44 pad detectors produced on p- and n-type MCz and Fz wafers was irradiated with 23 GeV protons, 200 MeV pions and reactor neutrons up to the equivalent fluences of Φ eq =3x10 15 cm -2 . The evolution of the full depletion voltage and the leakage current were monitored during short- and long-term annealing. At selected representative annealing steps, charge collection measurements were performed for all samples with LHC speed electronics. Measurements of full depletion voltage, leakage current and charge collection efficiency were compared for different irradiation particles and silicon materials.

  4. Comparison of pad detectors produced on different silicon materials after irradiation with neutrons, protons and pions

    Energy Technology Data Exchange (ETDEWEB)

    Kramberger, G., E-mail: Gregor.Kramberger@ijs.s [Jozef Stefan Institute and Department of Physics, University of Ljubljana, SI-1000 Ljubljana (Slovenia); Cindro, V.; Dolenc, I.; Mandic, I.; Mikuz, M.; Zavrtanik, M. [Jozef Stefan Institute and Department of Physics, University of Ljubljana, SI-1000 Ljubljana (Slovenia)

    2010-01-01

    A set of 44 pad detectors produced on p- and n-type MCz and Fz wafers was irradiated with 23 GeV protons, 200 MeV pions and reactor neutrons up to the equivalent fluences of PHI{sub eq}=3x10{sup 15}cm{sup -2}. The evolution of the full depletion voltage and the leakage current were monitored during short- and long-term annealing. At selected representative annealing steps, charge collection measurements were performed for all samples with LHC speed electronics. Measurements of full depletion voltage, leakage current and charge collection efficiency were compared for different irradiation particles and silicon materials.

  5. Control of helium effects in irradiated materials based on theory and experiment

    International Nuclear Information System (INIS)

    Mansur, L.K.; Lee, E.H.; Maziasz, P.J.; Rowcliffe, A.F.

    1986-01-01

    Helium produced in materials by (n,α) transmutation reactions during neutron irradiations or subjected in ion bombardment experiments causes substantial changes in the response to displacement damage. In particular, swelling, phase transformations and embrittlement are strongly affected. Present understanding of the mechanisms underlying these effects is reviewed. Key theoretical relationships describing helium effects on swelling and helium diffusion are described. Experimental data in the areas of helium effects on swelling and precipitation is reviewed with emphasis on critical experiments that have been designed and evaluated in conjunction with theory. Confirmed principles for alloy design to control irradiation performance are described

  6. Recommendations on the measurement of irradiation received by the structural materials of reactors

    International Nuclear Information System (INIS)

    Genthon, J.P.; Mas, P.; Wright, S.B.; Zijp, W.L.

    1975-01-01

    The recommendations have been compiled by a working group Radiation Damage which has been set up by the Euratom Working Group for reactor Dosimetry. The parameters are indicated which must be defined for the characterisation of the neutron dose causing radiation-induced damage in construction materials important for reactor technique. Following an explanation of some theoretical aspects, practical guidelines for neutron metrology on irradiation of graphite and of metals are given. A thorough knowledge of the spectrum of the incident neutrons is required for a proper interpretation of the results of irradiation experiments

  7. Consequences of the improvement of fast reactor material behavior under irradiation on fuel element performance

    International Nuclear Information System (INIS)

    Leclere, J.; Dupouy, J.M.; Marcon, J.P.

    1979-01-01

    The most important problems in fast reactor fuel element come from the excessive swelling of the structural materials used. The limitations of irradiation time for a given reactor result from the cladding or hexagonal wrapper deformations. Irradiation creep plays a major role, either in inducing additional deformations, or in providing possible ways of accommodation of bending stresses. Progress has been made in designing swelling resistant and/or low irradiation creep modulus materials. For instance in FRANCE, annealed 316 SS has been eliminated from pin and subassembly, and replaced by cold worked 316; we are now considering introduction of stabilizing elements in 316 SS as a further improvement and studying different alloys (nickel alloys, or ferritic steels). It has to be checked that the improvement of irradiation characteristic is not counterbalanced by losses on other properties (embrittlement for instance). Considering that pushing off or eliminating a limit may lead to the onset of a new one, it is porposed to make a review of the consequences of substantial improvement of structural material behavior

  8. A TEM method for analyzing local strain fields in irradiated materials

    International Nuclear Information System (INIS)

    Bennetch, J.I.; Jesser, W.A.

    1983-01-01

    Of great interest to the field of fracture mechanics is the strain field in front of a crack tip. In irradiated materials, cavities which naturally form as a result of radiation provide convenient internal markers. If a miniaturized irradiated tensile sample is pulled in situ in a transmission electron microscope (TEM), both the relative displacement of these cavities and their distortion in shape provide information on localized strain on a microscopic level. In addition, the TEM method allows direct correlation of active slip systems with crack propagation characteristics. To illustrate this method a strain field map was constructed about a crack propagating in a helium irradiated type 316 stainless steel sample containing large cavities. (orig.)

  9. DBMS Development of Irradiated Materials and Spare parts on master-slave manipulator in IMEF

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Y. S.; Kim, D. S.; Jung, Y. H.; Kim, H. M.; Yoo, B. O.; Baik, S. J.; Hong, K. P.; Ahn, S. B.; Ryu, W. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    The data of irradiated specimens(include nuclear fuel) which are transported from research reactor and commercial power reactor and the spare parts of the master-slave manipulator for the IMEF facility, which is operated since 1996, were controlled and managed through the Hangul and Excel software. But it is recommended to use a special program, which is developed for DBMS, for the beneficial control and systematic management of all irradiated specimens, especially assuming the increase of specimen's kind and amount by increasing customers in the near future. This report summarized the whole logical and physical processes and results about following items : - Management System of Irradiated Materials including nuclear fuel - Management System of spare parts for the master-slave manipulator.

  10. Growth and instability of charged dislocation loops under irradiation in ceramic materials

    CERN Document Server

    Ryazanov, A I; Kinoshita, C; Klaptsov, A V

    2002-01-01

    We have investigated the physical mechanisms of the growth and stability of charged dislocation loops in ceramic materials with very strong different mass of atoms (stabilized cubic zirconia) under different energies and types of irradiation conditions: 100-1000 keV electrons, 100 keV He sup + and 300 keV O sup + ions. The anomalous formation of extended defect clusters (charged dislocation loops) has been observed by TEM under electron irradiation subsequent to ion irradiation. It is demonstrated that very strong strain field (contrast) near charged dislocation loops is formed. The dislocation loops grow up to a critical size and after then become unstable. The instability of the charged dislocation loop leads to the multiplication of dislocation loops and the formation of dislocation network near the charged dislocation loops. A theoretical model is suggested for the explanation of the growth and stability of the charged dislocation loop, taking the charge state of point defects. The calculated distribution...

  11. Observation of He bubbles in ion irradiated fusion materials by conductive atomic force microscopy

    Energy Technology Data Exchange (ETDEWEB)

    Fan, Hongyu [School of Physics and Materials Engineering, Dalian Nationalities University, Dalian 116600 (China); Li, Ruihuan [School of Physics and Optoelectronic Engineering, Dalian University of Technology, Dalian 116024 (China); Yang, Deming [School of Physics and Materials Engineering, Dalian Nationalities University, Dalian 116600 (China); School of Science, Changchun University of Science and Technology, Changchun, Jilin 130022 (China); Wu, Yunfeng; Niu, Jinhai; Yang, Qi [School of Physics and Materials Engineering, Dalian Nationalities University, Dalian 116600 (China); Zhao, Jijun [School of Physics and Optoelectronic Engineering, Dalian University of Technology, Dalian 116024 (China); Liu, Dongping, E-mail: dongping.liu@dlnu.edu.cn [School of Physics and Materials Engineering, Dalian Nationalities University, Dalian 116600 (China); Fujian Key Laboratory for Plasma and Magnetic Resonance, Department of Electronic Science, Aeronautics, School of Physics and Mechanical and Electrical Engineering, Xiamen University, Xiamen, Fujian 361005 (China)

    2013-10-15

    Using a non-destructive conductive atomic force microscope combined with the Ar{sup +} etching technique, we demonstrate that nanoscale and conductive He bubbles are formed in the implanted layer of single-crystalline 6H-SiC irradiated with 100 keV He{sup +}. We find that the surface swelling of irradiated SiC samples is well correlated with the growth of elliptic He bubbles in the implanted layer. First-principle calculations are performed to estimate the internal pressure of the He bubble in the void of SiC. Analysis indicates that nanoscale He bubbles acting as a captor capture the He atoms diffusing along the implanted layer at an evaluated temperature and result in the surface swelling of irradiated SiC materials.

  12. Standardization of accelerator irradiation procedures for simulation of neutron induced damage in reactor structural materials

    Science.gov (United States)

    Shao, Lin; Gigax, Jonathan; Chen, Di; Kim, Hyosim; Garner, Frank A.; Wang, Jing; Toloczko, Mychailo B.

    2017-10-01

    Self-ion irradiation is widely used as a method to simulate neutron damage in reactor structural materials. Accelerator-based simulation of void swelling, however, introduces a number of neutron-atypical features which require careful data extraction and, in some cases, introduction of innovative irradiation techniques to alleviate these issues. We briefly summarize three such atypical features: defect imbalance effects, pulsed beam effects, and carbon contamination. The latter issue has just been recently recognized as being relevant to simulation of void swelling and is discussed here in greater detail. It is shown that carbon ions are entrained in the ion beam by Coulomb force drag and accelerated toward the target surface. Beam-contaminant interactions are modeled using molecular dynamics simulation. By applying a multiple beam deflection technique, carbon and other contaminants can be effectively filtered out, as demonstrated in an irradiation of HT-9 alloy by 3.5 MeV Fe ions.

  13. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Executive summary

    International Nuclear Information System (INIS)

    1997-01-01

    This report is a summary of the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member

  14. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Executive summary

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

    This report is a summary of the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member.

  15. Preparation of a new gamma irradiated PVC-Olive oil cake plastic composite material

    International Nuclear Information System (INIS)

    Messaud, F.A.; Almsmary, Y.A.; Elwerfalli, S.M.; Benayad, S.M.; Haraga, S.O.; Benfaid, N.A.; Kabar, Y.M.

    2003-01-01

    This paper dealt with the investigation on preparing new plastic composite material, utilizing polyvinyl chloride polymer (a commercial product in abu-kammash chemical complex) and olive oil cake (a waste of many olive oil production factories), followed by gamma irradiation (26.3 Kg ry) o induce crosslinking of the polymer. The new material possess good, electrical and mechanical properties as compared to plastic products of (PVC plastic pipe factory), and which could be used as new construction anti corrosive material, such as special roofing and partitioning or household goods

  16. The effects of packaging materials on microbe population in irradiated traditional herbal medicines

    International Nuclear Information System (INIS)

    Bagiawati, Sri; Hilmy, Nazly

    1983-01-01

    Microbial population and moisture content of traditional herbal medicines contaminated with 3 kinds of aerobic microbes, packed in 5 kinds of plastic packaging materials, followed by irradiation at minimum dose of 5 kGy and stored for 6 months were investigated. The highest reduction of microbial counts during storage was observed on samples packed in polyethylene bags. All of packaging materials used were found to be impermeable to microbes and water vapour. Radiation and packaging materials used acted synergistically to inactivate microbes durind storage. The microbial counts decreased as much as 2 to 4 log cycles during storage. (author)

  17. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Final report

    International Nuclear Information System (INIS)

    Martone, M.

    1997-01-01

    This report documents the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member

  18. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Martone, M [ENEA, Centro Ricerche Frascati, Rome (Italy)

    1997-01-01

    This report documents the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member.

  19. Thermal and irradiation effects on high-temperature mechanical properties of materials for SCWR fuel cladding

    International Nuclear Information System (INIS)

    Kano, F.; Tsuchiya, Y.; Oka, K.

    2009-01-01

    The thermal and irradiation effects on high-temperature mechanical properties are examined for candidate alloys for fuel cladding of supercritical water-cooled reactors (SCRWs). JMTR (Japan Materials Testing Reactor) and Experimental Fast Reactor JOYO were utilized for neutron irradiation tests, considering their fluence and temperature. Irradiation was performed with JMTR at 600degC up to 4x10 24 n/m 2 and with JOYO at 600degC and 700degC up to 6x10 25 n/m 2 . Tensile test, creep test and hardness measurement were carried out for high-temperature mechanical properties. Based on the uniaxial creep test, the extrapolation curves were drawn with time-temperature relationships utilizing the Larson and Miller Parameter. Several candidate alloys are expected to satisfy the design requirement from the estimation of the creep rupture stress for 50000 hours. Comparing the creep strengths under irradiated and unirradiated conditions, it was inferred that creep deformation was dominated by the thermal effect rather than the irradiation at SCWR core condition. The microstructure was examined using transmission electron microscope (TEM) analysis, focusing on void swelling and helium (He) bubble formation. Void formation was observed in the materials irradiated with JOYO at 600degC but not at 700degC. However, its effect on the deformation of components was estimated to be tolerable since their size and density were negligibly small. The manufacturability of the thin-wall, small-diameter tube was confirmed for the potential candidate alloys through the trial tests in the factory where the fuel cladding tube is manufactured. (author)

  20. Materials Characterization Center. Second workshop on irradiation effects in nuclear waste forms. Summary report

    International Nuclear Information System (INIS)

    Weber, W.J.; Turcotte, R.P.

    1982-01-01

    The purpose of this second workshop on irradiations effects was to continue the discussions initiated at the first workshop and to obtain guidance for the Materials Characterization Center in developing test methods. The following major conclusions were reached: Ion or neutron irradiations are not substitutes for the actinide-doping technique, as described by the MCC-6 Method for Preparation and Characterization of Actinide-Doped Waste Forms, in the final evaluation of any waste form with respect to the radiation effects from actinide decay. Ion or neutron irradiations may be useful for screening tests or more fundamental studies. The use of these simulation techniques as screening tests for actinide decay requires that a correlation between ion or neutron irradiations and actinide decay be established. Such a correlation has not yet been established and experimental programs in this area are highly recommended. There is a need for more fundamental studies on dose-rate effects, temperature dependence, and the nature and importance of alpha-particle effects relative to the recoil nucleus in actinide decay. There are insufficient data presently available to evaluate the potential for damage from ionizing radiation in nuclear waste forms. No additional test methods were recommended for using ion or neutron irradiations to simulate actinide decay or for testing ionization damage in nuclear waste forms. It was recognized that additional test methods may be required and developed as more data become available. An American Society for Testing and Materials (ASTM) Task Group on the Simulation of Radiation Effects in Nuclear Waste Forms (E 10.08.03) was organized to act as a continuing vehicle for discussions and development of procedures, particularly with regard to ion irradiations

  1. Effect of irradiation on the microbiological status and flavouring materials of selected spices

    International Nuclear Information System (INIS)

    Farag, S.E.D.A.; Aziz, N.H.; Attia, E.S.A.

    1995-01-01

    Spices from Egyptian local markets were irradiated with different recommended doses (0, 5, 10, 20 and 30 kGy). The spices tested included dried leaves of marjoram (Majorana hortensis Moench), rhizomes of ginger (Zingiber officinale Roscoe) and powdered hot pepper (Capsicum annum L.). The study included the isolation and identification of micro-organisms in spices following their irradiation, as well as gas chromatographic (GLC) chemical analysis for the presence and structure of volatile oils, pungent and pigment materials. The results showed that hot pepper was contaminated more (9.2x10 5 /g) than marjoram (4.2x10 3 /g) and ginger (14.3x10 3 /g) with respect to total aerobic bacterial content. The total contents of moulds were 4.8x10 3 /g, 5.7x10 3 /g and 19x10 3 /g in the same spices, respectively, but the pathogenic moulds and bacterial strains differed according to the type of spice. Irradiation at 10, 20 and 30 kGy caused complete elimination of mirco-organisms, whereas 5 kGy was less effective. With the GLC method chosen 18 and 50 compounds could be detected in the extracts of marjoram and ginger, respectively; γ-terpinen and zingiberen being the major compounds in marjoram and ginger, respectively. A noticeable reduction was observed in the amount of terpenes present in irradiated marjoram; they were converted to monoterpensalcohols. Ginger was more sensitive to irradiation, especially at high doses, but moderate changes were detected at low doses (5 and 10 kGy). A slight, but significant effect on the capsaicin (pungent compound) in hot pepper was observed following irradiation, whereas no changes in total pigments resulted at any dose. These results prove that 10 kGy is a sufficiently high dose to eliminate the microorganisms in spices, causing only slight changes in the flavouring materials. (orig.)

  2. Effect of irradiation on the microbiological status and flavouring materials of selected spices

    Energy Technology Data Exchange (ETDEWEB)

    Farag, S.E.D.A. [National Centre for Radiation Research and Technology, Cairo (Egypt); Aziz, N.H. [National Centre for Radiation Research and Technology, Cairo (Egypt); Attia, E.S.A. [Egyptian Starch and Glucose Manufacturing Co., Cairo (Egypt)

    1995-12-31

    Spices from Egyptian local markets were irradiated with different recommended doses (0, 5, 10, 20 and 30 kGy). The spices tested included dried leaves of marjoram (Majorana hortensis Moench), rhizomes of ginger (Zingiber officinale Roscoe) and powdered hot pepper (Capsicum annum L.). The study included the isolation and identification of micro-organisms in spices following their irradiation, as well as gas chromatographic (GLC) chemical analysis for the presence and structure of volatile oils, pungent and pigment materials. The results showed that hot pepper was contaminated more (9.2x10{sup 5}/g) than marjoram (4.2x10{sup 3}/g) and ginger (14.3x10{sup 3}/g) with respect to total aerobic bacterial content. The total contents of moulds were 4.8x10{sup 3}/g, 5.7x10{sup 3}/g and 19x10{sup 3}/g in the same spices, respectively, but the pathogenic moulds and bacterial strains differed according to the type of spice. Irradiation at 10, 20 and 30 kGy caused complete elimination of mirco-organisms, whereas 5 kGy was less effective. With the GLC method chosen 18 and 50 compounds could be detected in the extracts of marjoram and ginger, respectively; {gamma}-terpinen and zingiberen being the major compounds in marjoram and ginger, respectively. A noticeable reduction was observed in the amount of terpenes present in irradiated marjoram; they were converted to monoterpensalcohols. Ginger was more sensitive to irradiation, especially at high doses, but moderate changes were detected at low doses (5 and 10 kGy). A slight, but significant effect on the capsaicin (pungent compound) in hot pepper was observed following irradiation, whereas no changes in total pigments resulted at any dose. These results prove that 10 kGy is a sufficiently high dose to eliminate the microorganisms in spices, causing only slight changes in the flavouring materials. (orig.)

  3. Materials characterization center workshop on the irradiation effects in nuclear waste forms

    International Nuclear Information System (INIS)

    Roberts, F.P.; Turcotte, R.P.; Weber, W.J.

    1981-01-01

    The Workshop on Irradiation Effects in Nuclear Waste Forms sponsored by the Materials Characterization Center (MCC) brought together experts in radiation damage in materials and waste-management technology to review the problems associated with irradiation effects on waste-form integrity and to evaluate standard methods for generating data to be included in the Nuclear Waste Materials Handbook. The workshop reached the following conclusions: the concept of Standard Test for the Effects of Alpha-Decay in Nuclear Waste Solids, (MCC-6) for evaluating the effects of alpha decay is valid and useful, and as a result of the workshop, modifications to the proposed procedure will be incorpoated in a revised version of MCC-6; the MCC-6 test is not applicable to the evaluation of radiation damage in spent fuel; plutonium-238 is recommended as the dopant for transuranic and defense high-level waste forms, and when high doses are required, as in the case of commercial high-level waste forms, 244 Cm can be used; among the important property changes caused by irradiation are those that lead to greater leachability, and additionally, radiolysis of the leachant may increase leach rates; research is needed in this area; ionization-induced changes in physical properties can be as important as displacement damage in some materials, and a synergism is also likely to exist from the combined effects of ionization and displacement damage; and the effect of changing the temperature and dose rates on property changes induced by radiation damage needs to be determined

  4. Importance diagrams - a novel presentation of the response of a material to neutron irradiation

    International Nuclear Information System (INIS)

    Forrest, R.A.

    1998-01-01

    Activation of fusion materials following neutron irradiation is of great technological importance, especially in the study of safety and environmental impacts. Currently, activation calculations are performed for a particular neutron spectrum, appropriate to a region in a particular fusion device, which makes it difficult to extract generic information. The present work gives details of a method to present the dominant nuclides for the radiological responses or an irradiated material in a fashion that is independent of the neutron spectrum and almost independent of the flux. The importance diagrams show regions in the decay time versus neutron energy space where a nuclide contributes >50% of the response. The importance diagrams for pure iron and SS316 are described, and it is noted that the shapes of the various regions vary very little with the total neutron flux. Variation of the diagrams with irradiation time occurs at short decay times in a systematic fashion. The use of the diagrams in a realistic spectrum relies on an expansion, which while not generally true, does hold approximately for many of the nuclides of interest. The diagrams are therefore a valuable summary of the universal, device-independent, response of the materials, and when combined with pathway information give a comprehensive description of activation for that material. (orig.)

  5. Inert materials for the GFR fuel. Characterizations, chemical interactions and irradiation damage

    International Nuclear Information System (INIS)

    Audubert, Fabienne; Carlot, Gaoelle; Lechelle, Jacques; David, Laurent; Gomes, Severine

    2005-01-01

    In the framework of an extensive R and D Program on GFR fuel, studies on inert materials have been performed at the French Atomic Energy Commission (CEA). The inert materials would be associated with the fuel with the aim of featuring an efficient barrier to radiotoxic species with regard to the cooling circuit of the reactor. Potential matrices identified for dispersion fuels or particles fuels are SiC, TiN, ZrN, ZrC, TiC. Physical microstructural and thermal properties have been determined in order to evaluate elaboration process effects. The evolution under irradiation of thermal properties (such as conductivity, diffusivity) of the materials has been studied using heavy ions to simulate fission product irradiation. After irradiation, scanning thermal microscopy is used to investigate the thermal degradation of the materials. Thermal conductivity variations were obtained on TiC irradiated with krypton ion at an energy of 86 MeV and a fluence of 5.10 15 ions.cm -2 . They are quantified at 19 W.m -1 .K -1 . On other materials such as SiC, ZrC, TiN, no thermal conductivity contrast was shown. Reactivity between the inert matrix (SiC or TiN) and the fuel (U, Pu)N have been evaluated on powders and on ceramic samples in contact by a thermal treatment under several atmospheres. It was shown that SiC reacts with (U, Pu)N in various atmospheres making secondary phases as PuSi 2 , USi 2 , U 20 Si 16 C 3 . TiN behaviour seems to be better: the only reactivity which may take place would be a variation of the nitrogen stoichiometry in TiN and (U, Pu)N at the interface. (author)

  6. Simulated Irradiation of Samples in HFIR for use as Possible Test Materials in the MPEX (Material Plasma Exposure Experiment) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, Ronald James [ORNL; Rapp, Juergen [ORNL

    2014-01-01

    The importance of Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) facility will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. The project presented in this paper involved performing assessments of the induced radioactivity and resulting radiation fields of a variety of potential fusion reactor materials. The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR; generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. These state-of-the-art simulation methods were used in addressing the challenge of the MPEX project to minimize the radioactive inventory in the preparation of the samples for inclusion in the MPEX facility.

  7. Irradiation effects on C/C composite materials for high temperature nuclear applications

    International Nuclear Information System (INIS)

    Eto, M.; Ugachi, H.; Baba, S.I.; Ishiyama, S.; Ishihara, M.; Hayashi, K.

    2000-01-01

    Excellent characteristics such as high strength and high thermal shock resistance of C/C composite materials have led us to try to apply them to the high temperature components in nuclear facilities. Such components include the armour tile of the first wall and divertor of fusion reactor and the elements of control rod for the use in HTGR. One of the most important aspects to be clarified about C/C composites for nuclear applications is the effect of neutron irradiation on their properties. At the Japan Atomic Energy Research Institute (JAERI), research on the irradiation effects on various properties of C/C composite materials has been carried out using fission reactors (JRR-3, JMTR), accelerators (TANDEM, TIARA) and the Fusion Neutronics Source (FNS). Additionally, strength tests of some neutron-irradiated elements for the control rod were carried out to investigate the feasibility of C/C composites. The paper summarises the R and D activities on the irradiation effects on C/C composites. (authors)

  8. Mechanical Tests Plan after Neutron Irradiation for SMART SG Tube Materials in a Hot Cell

    International Nuclear Information System (INIS)

    Ahn, Sang Bok; Baik, Seung Jai; Kim, Do Sik; Yoo, Byung Ok; Jung, Yang Hong; Song, Woong Sub; Choo, Kee Nam; Park, Jin Seok; Lee, Yong Sun; Ryu, Woo Seog

    2010-01-01

    An advanced integral PWR, SMART (System- Integrated Modular Advanced ReacTor) is being developed in KAERI. It has compact size and a relatively small power rating compared to a conventional reactor. The main components such as the steam generators, main circulation pumps are located in the reactor vessel. Therefore they are damaged from neutron irradiations generated from nuclear fuel fissions during operation. The SMART SG tubes which are 17 mm in a diameter and 2.5 mm in a thickness will be made of Alloy 690. To ensure the operation safety the post irradiation examinations is necessary to evaluate the deterioration levels of various original properties. Specially the amount of mechanical properties change should be reflected and revised to design data. For that tensile, fracture, hardness test are planned and under preparations. In this paper the detailed plans are reviewed. Three kinds of materials having different heat treatment procedures are prepared to fabricate specimens. The capsules installed the specimens are going to be irradiated in HANARO. Finally the tests for them will be performed in IMEF, Irradiated Materials Examination Facility at KAERI

  9. Effect of thermal annealing on property changes of neutron-irradiated non-graphitized carbon materials and nuclear graphite

    International Nuclear Information System (INIS)

    Matsuo, Hideto

    1991-06-01

    Changes in dimension of non-graphitized carbon materials and nuclear graphite, and the bulk density, electrical resistivity, Young's modulus and thermal expansivity of nuclear graphite were studied after neutron irradiation at 1128-1483 K and the successive thermal annealing up to 2573 K. Carbon materials showed larger and anisotropic dimensional shrinkage than that of nuclear graphite after the irradiation. The irradiation-induced dimensional shrinkage of carbon materials decreased during annealing at temperatures from 1773 to 2023 K, followed by a slight increase at higher temperatures. On the other hand, the irradiated nuclear graphite hardly showed the changes in length, density and thermal expansivity under the thermal annealing, but the electrical resistivity and Young's modulus showed a gradual decrease with annealing temperature. It has been clarified that there exists significant difference in the effect of thermal annealing on irradiation-induced dimensional shrinkage between graphitized nuclear graphite and non-graphitized carbon materials. (author)

  10. Developing Ultra-small Scale Mechanical Testing Methods and Microstructural Investigation Procedures for Irradiated Materials

    Energy Technology Data Exchange (ETDEWEB)

    Hosemann, Peter; Kaoumi, Djamel

    2018-04-02

    Nuclear materials are an essential aspect of nuclear engineering. While great effort is spent on designing more advanced reactors or enhancing a reactor’s safety, materials have been the bottleneck of most new developments. The designs of new reactor concepts are driven by neutronic and thermodynamic aspects, leading to unusual coolants (liquid metal, liquid salt, gases), higher temperatures, and higher radiation doses than conventional light water reactors have. However, any (nuclear) engineering design must consider the materials used in the anticipated application in order to ever be realized. Designs which may look easy, simple and efficient considering thermodynamics or neutronic aspects can show their true difficulty in the materials area, which then prevents them from being deployed. In turn, the materials available are influencing the neutronic and thermodynamic designs and therefore must be considered from the beginning, requiring close collaborations between different aspects of nuclear engineering. If a particular design requires new materials, the licensing of the reactor must be considered, but licensing can be a costly and time consuming process that results in long lead times to realize true materials innovation. Extensive materials evaluation and irradiation campaigns need to be conducted in order to introduce a new material in a nuclear system. For licensing purposes, standard materials testing is key. However, basic scientific studies on new materials or even already used materials have the potential to accelerate the process of materials development or foster predictability of materials that are already in service and therefore are essential in order not to face difficulties later in the development or service stage. Therefore a combination of engineering scale materials evaluation as well as basic scientific understanding of the materials property changes under service condition is key to address potential issues in the process. Ion

  11. Size-Tuned Plastic Flow Localization in Irradiated Materials at the Submicron Scale

    Science.gov (United States)

    Cui, Yinan; Po, Giacomo; Ghoniem, Nasr

    2018-05-01

    Three-dimensional discrete dislocation dynamics (3D-DDD) simulations reveal that, with reduction of sample size in the submicron regime, the mechanism of plastic flow localization in irradiated materials transitions from irradiation-controlled to an intrinsic dislocation source controlled. Furthermore, the spatial correlation of plastic deformation decreases due to weaker dislocation interactions and less frequent cross slip as the system size decreases, thus manifesting itself in thinner dislocation channels. A simple model of discrete dislocation source activation coupled with cross slip channel widening is developed to reproduce and physically explain this transition. In order to quantify the phenomenon of plastic flow localization, we introduce a "deformation localization index," with implications to the design of radiation-resistant materials.

  12. Pyrolysis responses of kevlar/epoxy composite materials on laser irradiating

    Science.gov (United States)

    Liu, Wei-ping; Wei, Cheng-hua; Zhou, Meng-lian; Ma, Zhi-liang; Song, Ming-ying; Wu, Li-xiong

    2017-05-01

    The pyrolysis responses of kevlar/epoxy composite materials are valuable to study in a case of high temperature rising rate for its widely application. Distinguishing from the Thermal Gravimetric Analysis method, an apparatus is built to research the pyrolysis responses of kevlar/epoxy composite materials irradiated by laser in order to offer a high temperature rising rate of the sample. By deploying the apparatus, a near real-time gas pressure response can be obtained. The sample mass is weighted before laser irradiating and after an experiment finished. Then, the gas products molecular weight and the sample mass loss evolution are derived. It is found that the pressure and mass of the gas products increase with the laser power if it is less than 240W, while the molecular weight varies inversely. The variation tendency is confusing while the laser power is bigger than 240W. It needs more deeper investigations to bring it to light.

  13. Study on ceramic breeder and related materials by means of work function measurement under irradiation

    International Nuclear Information System (INIS)

    Luo, G.N.; Terai, T.; Yamawaki, M.; Yamaguchi, K.

    2002-01-01

    Ceramic breeder materials, Li 2 O, LiAlO 2 and Li 4 SiO 4 , under irradiation have been studied using a Kelvin probe that measures work function changes of materials. Surface charging was observed to influence greatly the probe output, which can be explained qualitatively employing a model concerning induction electric field due to external field and free charges on ceramic surface. It is found that the insulating ceramics could not be studied properly with the Kelvin probe. A probable solution is to heat the ceramics, so as to raise their electric conductivities high enough to root out the surface charging. Also briefly discussed is the application of the probe to metals under ion irradiation. (orig.)

  14. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  15. Cytogenetics for dosimetry in cases of radiation accidents and assessing the safety of irradiated food material

    International Nuclear Information System (INIS)

    Natarajan, A.T.; Kesavan, P.C.

    2005-01-01

    One of the many areas of research initiated by Swaminathan at the Botany Division of the Indian Agricultural Research Institute, New Delhi was radiation cytogenetics, which involves study of induced chromosomal aberrations. These studies had impact not only on elucidating basic mechanisms involved in the formation of chromosomal aberrations, but also several practical applications related to human health. In this review, we briefly summarize two applications, namely biological dosimetry following radiation accidents and safety of irradiated food material. (author)

  16. Device for the generation of homogeneous dose distributions in irradiated materials

    International Nuclear Information System (INIS)

    Leonhardt, J.; Schulze, H.; Boes, J.; Decker, U.; Schmidt, J.

    1985-01-01

    The invention has been directed at a device for the generation of homogeneous dose distributions in materials irradiated by charged particles. This device can be applied to the initiation of radiation-chemical reactions in solids, of cross-linking and vulcanizing reactors, of crystal defect annealings, etc. A movable absorber (e.g. a wedge or a solid of revolution) which periodically changes the energy of particles striking the specimen has been installed in the beam hole of the beam generating system

  17. Influence of high dose irradiation on core structural and fuel materials in advanced reactors

    International Nuclear Information System (INIS)

    1998-08-01

    The IAEA International Working Group on Fast Reactors (IWGFR) periodically organizes meeting to discuss and review important aspects of fast reactor technology. The fifth meeting held in Obninsk, Russian Federation, 16-19 June 1997, was devoted to the influence of high dose irradiation on the mechanical properties of reactor core structural and fuel materials. The proceedings includes the papers submitted at this meeting each with a separate abstract

  18. Development of a Device for a Material Irradiation Test in the OR Test Hole

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Kang, Y. H.; Kim, B. G.; Choo, K. N.; Sohn, J. M.; Shin, Y. T.; Park, S. J.; Seo, C. K

    2008-05-15

    To develop a technology and a device for the irradiation test for utilization of the OR/IP holes according to the various requirements of users, the properties of the OR/IP holes were investigated and an irradiation device for the OR hole was designed and fabricated. The OR-4, 5 and the IP-9, 10, 11 holes were selected as those suitable to irradiation tests among the test holes located in the out core area. The conceptual design was performed to design a device to irradiate materials using the OR and IP holes. The capsule for the OR holes is fixed by pressing the protection tube using a clamping device, on the other hand the IP capsule is inserted in the hole without a special clamping device. In the basic design of the irradiation device for the OR hole, the capsules having the outside diameter of 50, 52, 54, 56mm were reviewed theoretically to investigate if they meet the hydraulic and vibration conditions required in the HANARO. The results of the pressure drop test showed that the 3 kinds of capsules having diameter of 52, 54, 56mm satisfied the requirement for the pressure difference and flow rate in HANARO. The capsule of {phi}56mm out of the above three satisfied the vibration condition and was finally selected giving consideration of a capacity of specimens. The capsule having a diameter of {phi}56mm was fabricated and the flow rate was measured. Using the velocity data measured at the out-core facility, the heat transfer coefficient, and the temperature on the surface of the capsule was evaluated to confirm it less than the ONB temperature. As a result, the capsule of {phi}56mm was selected for the irradiation test at the OR holes.

  19. PIREX II, a new irradiation facility for testing fusion first wall materials

    International Nuclear Information System (INIS)

    Marmy, P.; Daum, M.; Gavillet, D.; Green, S.; Green, W.V.; Hegedues, F.; Pronnecke, S.; Rohrer, U.; Stiefel, U.; Victoria, M.

    1988-12-01

    A new irradiation facility, PIREX II, became operational in March 1987. It is located on a dedicated beam line split from the main beam of the 590 MeV proton accelerator at the Paul Scherrer Institute (PSI). Irradiation with protons of this energy introduces simultaneously displacement damage, helium and other impurities. Because of the penetration range of 590 MeV protons, both damage and impurities are homogeneously distributed in the target. The installation has its own beam line optics that can support a proton current of up to 50 μA. At a typical beam density of 4 μA/mm 2 , the damage rate in steels is 0.7 x 10 -5 dpa/sec (dpa: displacements per atom) and the helium production rate is 170 appm He/dpa. Both flat tensile specimens of up to 0.4 mm thickness and tubular fatigue samples of 3 mm diameter can be irradiated. Cooling of the temperatures can be controlled between 100 o and 800 o C. Installation of an in situ low cycle fatigue device is foreseen. Beams of up to 20 μA have been obtained, the beam having approximately a gaussian distribution of elliptical cross section with 4 σ between 0.8 and 3 mm by 10 mm. Irradiations for a dosimetry program have been completed on samples of Al, Cu, Fe, Ni, Au, W, and the 1.4914 ferritic steel. The evaluation of results allows the correct choice of reactions to be used for determining total dose, from the standpoint of half life and gamma energy. A program of irradiations on candidate materials for the Next European Torus (NET) design (Cu and Cu alloys, the 1.4914 ferritic martensitic steel, W and W-Re alloys and Mo alloys), where the above mentioned characteristics of this type of irradiation can be used advantageously, is now under way. (author) 11 figs., 4 tabs., 20 refs

  20. Report Summarizing the Effort Required to Initiate Welding of Irradiated Materials within the Welding Cubicle

    Energy Technology Data Exchange (ETDEWEB)

    Frederick, Greg [Electric Power Research Institute (EPRI), Palo Alto, CA (United States); Sutton, Benjamin J. [Electric Power Research Institute (EPRI), Palo Alto, CA (United States); Tatman, Jonathan K. [Electric Power Research Institute (EPRI), Palo Alto, CA (United States); Vance, Mark Christopher [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Smith, Allen W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Clark, Scarlett R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Feng, Zhili [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Roger G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chen, Jian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Tang, Wei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hu, Xunxiang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gibson, Brian T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    The advanced welding facility within a hot cell at the Radiochemical Engineering Development Center of Oak Ridge National Laboratory (ORNL), which has been jointly funded by the U.S. Department of Energy (DOE), Office of Nuclear Energy, Light Water Reactor Sustainability Program and the Electric Power Research Institute, Long Term Operations Program and the Welding and Repair Technology Center, is in the final phase of development. Research and development activities in this facility will involve direct testing of advanced welding technologies on irradiated materials in order to address the primary technical challenge of helium induced cracking that can arise when conventional fusion welding techniques are utilized on neutron irradiated stainless steels and nickel-base alloys. This report details the effort that has been required since the beginning of fiscal year 2017 to initiate welding research and development activities on irradiated materials within the hot cell cubicle, which houses welding sub-systems that include laser beam welding (LBW) and friction stir welding (FSW) and provides material containment within the hot cell.

  1. STUDY STRUCTURE OF THREE-COMPONENT POLYMERIC MATERIAL UNDER INFLUENCE OF γ-IRRADIATION

    Directory of Open Access Journals (Sweden)

    V. T. Tarasyuk

    2017-01-01

    Full Text Available The polymer material (РА/РЕ/Eva with a width of 55 μm was studied. Sterilization was carried out on the unit GU–200 at doses from 3 to 18 kGy in the Research Institute of Technical Physics and Automation, Rosatom, Moscow, Russia. The structure of the polymermaterial samples was studied by IR spectroscopy before and after irradiation in a range of 400–5000 сm–1. According to the results of the analysis of the IR spectrum structure, the changes in the structure were insignificant upon irradiation at doses up to 6 kGy. Upon irradiation at doses from 9 kGy and higher, an increase in quantity of ester groups (2340 сm–1 and insignificant increase in other functional groups were observed, which can suggest a simultaneous process of intra-molecular cross-linking with the intermediatestage of cross-linking occurring with formation of vinylene groups. This causes destruction of a polymer material and radiation oxidation. These disorders can lead to changes in physico-mechanical and barrier parameters of a polymer material, which can be notably reflected in the shelf life of agricultural products.

  2. Effect of packaging materials on the quality of irradiated ground spices

    International Nuclear Information System (INIS)

    Saputra, T.S.; Maha, Munsiah; Purwanto, Z.I.

    1985-01-01

    These experiments were carried out to determine the suitable packaging materials to be used for irradiated ground spices produced in Indonesia. The materials used were white pepper (Piper album), black pepper (Piper nigrum) nutmeg (Myristica fragrans), turmeric (Curcuma domestica), and ginger (Zangiber officinale R.) packaged in transparent polypropylene bottles, in pouches made of cellophane-aluminum foil and lithopaper-polyethylene laminates. The samples were irradiated at 5 kGy, stored at ambient conditions, and then examined every 3 months from 0 up to 9 months of storage. The parameters observed were total bacterial counts, total moulds and yeast counts, water activity (Aw), moisture content, and organoleptic scores of the samples. Piperine content of white pepper and black pepper, colour of turmeric extract, and rancidity of ginger were also determined. The results showed that the packaging materials used had no significant effect on bacterial load of the samples. Prolonged storage, however, could reduce the microbial load of the ground spices. Irradiation at 5 kGy could effectively increase the hygienic condition as well as storage life of the ground spices under investigation without affecting their organoleptic properties. (author). 8 refs

  3. Irradiation of gelatin. Important applications for the development of new materials

    Energy Technology Data Exchange (ETDEWEB)

    Del Mastro, N. L., E-mail: nlmastro@ipen.br [Instituto de Pesquisas Energeticas e Nucleares IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242, 05008-900 Sao Paulo (Brazil)

    2011-11-15

    Gelatin is obtained from a naturally occurring protein, collagen, by chemical and thermal hydrolysis. As a protein, gelatin is biodegradable. Among biomaterials, gelatin is an interesting material because the dehydrated gelatin is a partially crystalline polymer and has a relatively low melting point. Gelatin melts to a liquid when heated and solidifies when cooled again. Together with water, it forms a semi-solid colloid gel. If gelatin is placed in contact with cold water, some of the material dissolves, their solubility being determined by their manufacturing method. Ionizing radiation acting on aqueous biological systems produces labile intermediates. Macromolecular free radicals may be diffusion ally mobile in homogenous solutions but relatively immobile in gelled systems. In this paper, different works, some of them performed in our laboratory, are firstly described showing gelatin irradiation in diverse systems. We have already studied the mechanical properties of gelatin composites prepared with a natural fiber, plasticizer and treated by electron beam irradiation. Departing from that, the range of novel applications for gelatin composites like gelatin nanoparticles as biodegradable s and low cell toxic alternative carrier delivery systems are outlined. The potential and the possibilities of using gelatin irradiation for important applications for the development of new materials for medical and food industry are presented. (Author)

  4. Components production and assemble of the irradiation capsule of the Surveillance Program of Materials of the nuclear power plant of Laguna Verde

    International Nuclear Information System (INIS)

    Medrano, A.

    2009-01-01

    To predict the effects of the neutrons radiation and the thermal environment about the mechanical properties of the reactor vessel materials of the nuclear power plant of Laguna Verde, a surveillance program is implemented according to the outlines settled by Astm E185-02 -Standard practice for design of surveillance programs for light-water moderated nuclear power reactor vessels-. This program includes the installation of three irradiation capsules of similar materials to those of the reactor vessels, these samples are test tubes for mechanical practices of impact and tension. In the National Institute of Nuclear Research and due to the infrastructure as well as of the actual human resources of the Pilot Plant of Nuclear Fuel Assembles Production it was possible to realize the materials rebuilding extracted in 2005 of Unit 2 of nuclear power plant of Laguna Verde as well as the production, assemble and reassignment of the irradiation capsule made in 2006. At the present time the surveillance materials extracted in 2008 of Unit 1 of the nuclear power plant of Laguna Verde are reconstituting and the components are manufactured for the assembles of the irradiation capsule that will be reinstalled in the reactor vessel in 2010. The purpose of the present work is to describe the necessary components as well as its disposition during the assembles of the irradiation capsule for the surveillance program of the reactors vessel of the nuclear power plant of Laguna Verde. (Author)

  5. Magnetization and flux pinning in high-Tc cuprates: Irradiated and oxygen deficient materials

    International Nuclear Information System (INIS)

    Thompson, J.R.; Civale, L.; Marwick, A.D.; Holtzberg, F.

    1992-11-01

    This work surveys recent studies of the intragrain current density J and vortex pinning in high Tc superconductors. Materials include Y 1 Ba 2 Cu 3 O 7-δ and Bi 2 Sr 2 Ca 1 Cu 2 O 8 single crystals and aligned polycrystals. To probe the flux pinning, we modified the strength, number, and morphology of defects. Varying the oxygen content (7-δ) in YBa 2 Cu 3 O 7-δ or irradiating the materials with ions, having either light or heavy masses, gives systematic changes in the character of the all-important defects

  6. Displacement per atom profile in carbon nanotube bulk material under gamma irradiation

    International Nuclear Information System (INIS)

    Leyva, A.; Pinnera, I.; Leyva, D.; Cruz, C.; Abreu, Y.

    2011-01-01

    Taking into account the physical properties and the displacement threshold energy values reported in literature for C atoms in single and multiple walled carbon nanotubes, the effective atomic displacement cross-section in carbon nanotube bulk materials exposed to the gamma rays were calculated. Then, using the mathematical simulation of photons and particles transport in the matter, energy fluxes distribution of electrons and positrons within the irradiated object were also calculated. Finally, considering both results, the atomic displacement damage profiles inside the analyzed carbon nanotube bulk materials were determined. (Author)

  7. IFMIF (International Fusion Materials Irradiation Facility) key element technology phase interim report

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-03-01

    Activities of International Fusion Materials Irradiation Facility (IFMIF) have been performed under an IEA collaboration since 1995. IFMIF is an accelerator-based deuteron (D{sup +})-lithium (Li) neutron source designed to produce an intense neutron field (2 MW/m{sup 2}, 20 dpa/year for Fe) in a volume of 500 cm{sup 3} for testing candidate fusion materials. In 2000, a 3 year Key Element technology Phase (KEP) of IFMIF was started to reduce the key technology risk factors. This interim report summarizes the KEP activities until mid 2001 in the major project work-breakdown areas of accelerator, target, test facilities and design integration. (author)

  8. Design of a high-flux test assembly for the Fusion Materials Irradiation Test Facility

    International Nuclear Information System (INIS)

    Opperman, E.K.; Vogel, M.A.

    1982-01-01

    The Fusion Material Test Facility (FMIT) will provide a high flux fusion-like neutron environment in which a variety of structural and non-structural materials irradiations can be conducted. The FMIT experiments, called test assemblies, that are subjected to the highest neutron flux magnitudes and associated heating rates will require forced convection liquid metal cooling systems to remove the neutron deposited power and maintain test specimens at uniform temperatures. A brief description of the FMIT facility and experimental areas is given with emphasis on the design, capabilities and handling of the high flux test assembly

  9. HiRadMat: A high‐energy, pulsed beam, material irradiation facility

    CERN Multimedia

    Charitonidis, Nikolaos

    2016-01-01

    HiRadMat is a facility constructed in 2011, designed to provide high-intensity pulsed beams to an irradiation area where different material samples or accelerator components can be tested. The facility, located at the CERN SPS accelerator complex, uses a 440 GeV proton beam with a pulse length up to 7.2 μs and a maximum intensity up to 1E13 protons / pulse. The facility, a unique place for performing state-of-the art beam-to-material experiments, operates under transnational access and welcomes and financially supports, under certain conditions, experimental teams to perform their experiments.

  10. IFMIF (International Fusion Materials Irradiation Facility) key element technology phase interim report

    International Nuclear Information System (INIS)

    Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki

    2002-03-01

    Activities of International Fusion Materials Irradiation Facility (IFMIF) have been performed under an IEA collaboration since 1995. IFMIF is an accelerator-based deuteron (D + )-lithium (Li) neutron source designed to produce an intense neutron field (2 MW/m 2 , 20 dpa/year for Fe) in a volume of 500 cm 3 for testing candidate fusion materials. In 2000, a 3 year Key Element technology Phase (KEP) of IFMIF was started to reduce the key technology risk factors. This interim report summarizes the KEP activities until mid 2001 in the major project work-breakdown areas of accelerator, target, test facilities and design integration. (author)

  11. Newly developed non-destructive testing method for evaluation of irradiation brittleness of structural materials using ultrasonic

    International Nuclear Information System (INIS)

    Ishii, Toshimitsu; Ooka, Norikazu; Kato, Yoshiaki; Saito, Junichi; Hoshiya, Taiji; Shibata, Saburo; Kobayashi, Hideo

    1999-01-01

    Surveillance testing is important to evaluate neutron irradiation embrittlement of reactor pressure vessel material for long life operation. An alternative test method for evaluating the irradiation embrittlement of the pressure vessel material will have to be proposed to support the limited number of surveillance test specimens in order to manage the plant life to be extended. In this study, ultrasonic testing for irradiated A533B-1 steel and weld metal was applied to examine material degradation nondestructively. With increasing the shift of Charpy 41 J transition temperature, ultrasonic velocity decreased and attenuation coefficient of ultrasonic wave increased. Especially, the difference of ultrasonic velocity for 5 MHz shear wave between as-received and irradiated material is corresponding to the shift of transition temperature showing material degradation. (author)

  12. Material modifications in lithium niobate and lithium tantalate crystals by ion irradiation

    International Nuclear Information System (INIS)

    Raeth, Niels Lennart

    2017-01-01

    The artificially produced crystals lithium niobate (LiNbO 3 ) and the closely related lithium tantalate (LiTaO 3 ) are proven starting materials for producing active and passive devices that can guide, amplify, switch and process light. For this purpose, it is often necessary to be able to influence the refractive index of the substrate targeted, which is possible in addition to other methods by irradiation of the materials with fast light ions. In this work, lithium niobate and lithium tantalate crystals are irradiated with alpha particles, 3 He ions, deuterons, and protons at projectile energies of up to 14 MeV / nucleon. Energy and crystal thickness are chosen so that the projectiles penetrate the entire sample and are not implanted. All isotopes responsible for the unwanted nuclear activation of the crystals due to the irradiation are relatively short-lived and overall the activation decreases fast enough to allow the safe handling of the irradiated samples after a storage period of a few days to a few weeks. The refractive index changes produced in lithium niobate and lithium tantalate by irradiation with the different projectiles are determined interferometrically and can also be measured by suitable choice of the sample geometry as a function of the ion penetration depth: In LiNbO 3 the ordinary refractive index decreases, the extraordinary increases equally. In LiTaO 3 , both the ordinary and the extraordinary refractive indices decrease as a result of the irradiation; the ordinary refractive index change is many times stronger than the extraordinary one. There is an enormous long-term stability at room temperature for both crystal systems: Even after eleven (LiNbO 3 ) or three (LiTaO 3 ) years, no decrease in the ion beam-induced refractive index change can be observed. The ion beam-induced refractive index changes are probably the result of atomic displacements such as vacancies, defect clusters or ''latent tracks''. An explanation for

  13. Influence of specimen size/type on the fracture toughness of five irradiated RPV materials

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lucon, Enrico [National Inst. of Standards and Technology (NIST), Boulder, CO (United States)

    2015-01-01

    The Heavy-Section Steel Irradiation (HSSI) Program had previously irradiated five reactor pressure vessel (RPV) steels/welds at fast neutron fluxes of about 4 to 8 x 1011 n/cm2/s (>1 MeV) to fluences from 0.5 to 3.4 1019 n/cm2 and at 288 °C. The unirradiated fracture toughness tests were performed by Oak Ridge National Laboratory with 12.7-mm and 25.4-mm thick (0.5T and 1T) compact specimens, while the HSSI Program provided tensile and 5 x 10-mm three-point bend specimens to SCK-CEN for irradiation in the in-pile section of the Belgian Reactor BR2 at fluxes > 1013 n/cm2/s and subsequent testing by SCK-CEN. The BR2 irradiations were conducted at about 2 and 4 x 1013 n/cm2/s with irradiation temperature between 295 °C and 300 °C (water temperature), and to fluences between 6 and 10 x 1019n/cm2. The irradiation-induced shifts of the Master Curve reference temperatures, ΔT0, for most of the materials deviated from the embrittlement correlations much more than expected, motivating the testing of 5 x 10-mm three-point bend specimens of all five materials in the unirradiated condition to eliminate specimen size and geometry as a variable. Tests of the unirradiated small bend specimens resulted in Master Curve reference temperatures, T0, 25 °C to 53 °C lower than those from the larger compact specimens, meaning that the irradiation-induced reference temperature shifts, ΔT0, were larger than the initial measurements, resulting in much improved agreement between the measured and predicted fracture toughness shifts.

  14. Lattice strain in irradiated materials unveils a prevalent defect evolution mechanism

    Science.gov (United States)

    Debelle, Aurélien; Crocombette, Jean-Paul; Boulle, Alexandre; Chartier, Alain; Jourdan, Thomas; Pellegrino, Stéphanie; Bachiller-Perea, Diana; Carpentier, Denise; Channagiri, Jayanth; Nguyen, Tien-Hien; Garrido, Frédérico; Thomé, Lionel

    2018-01-01

    Modification of materials using ion beams has become a widespread route to improve or design materials for advanced applications, from ion doping for microelectronic devices to emulation of nuclear reactor environments. Yet, despite decades of studies, major issues regarding ion/solid interactions are not solved, one of them being the lattice-strain development process in irradiated crystals. In this work, we address this question using a consistent approach that combines x-ray diffraction (XRD) measurements with both molecular dynamics (MD) and rate equation cluster dynamics (RECD) simulations. We investigate four distinct materials that differ notably in terms of crystalline structure and nature of the atomic bonding. We demonstrate that these materials exhibit a common behavior with respect to the strain development process. In fact, a strain build-up followed by a strain relaxation is observed in the four investigated cases. The strain variation is unambiguously ascribed to a change in the defect configuration, as revealed by MD simulations. Strain development is due to the clustering of interstitial defects into dislocation loops, while the strain release is associated with the disappearance of these loops through their integration into a network of dislocation lines. RECD calculations of strain depth profiles, which are in agreement with experimental data, indicate that the driving force for the change in the defect nature is the defect clustering process. This study paves the way for quantitative predictions of the microstructure changes in irradiated materials.

  15. EFFECT OF IRRADIATION AND PACKAGING MATERIALS TYPES ON SHELF-LIFE AND QUALITY ATTRIBUTES OF MINCED MEAT DURING COLD STORAGE

    International Nuclear Information System (INIS)

    OSHEBA, A.S.; NAGY, KH.S.; ANWAR, M.M.

    2008-01-01

    Minced meat is considered one of the most meat products that exposed to contamination which led to many changes in its quality and reduced its shelf-life.Therefore, this investigation was carried out to extend the shelf-life of minced meat for consumption and maintaining its quality during cold storage by using irradiation with various doses (3, 6 and 9 kGy) and different packing materials. The results indicated that irradiation,especially at 3 and 6 kGy, had no effect on chemical composition and some physical properties of minced meat. On the other hand, pH values of all irradiated samples were slightly decreased with decreasing irradiation doses.Irradiation at the highest dose used, i.e. 9 kGy, slightly increased total volatile nitrogen (TVN) of minced meat. Thiobarbituric acid (TBA) value of irradiated samples was tended to increase with increasing irradiation dose from 3 to 9 kGy either directly after irradiation or during storage.Regardless of irradiation effect on TVN and TBA values at zero time, there were no marked differences in TVN and TBA values of irradiated minced meat according to differentiate packaging materials (PE, PA/PE and PET/Al/PE). During cold storage, the TVN and TBA values of all minced meat samples either non-irradiated or irradiated were progressively increased as the time of cold storage increased. The higher increasing rate in TVN and TBA of irradiated samples was recorded for samples packaged in PE (one layer) followed by PA/PE (two layers) and finally PET/Al/PE (three layers) at the same irradiation dose. Irradiation of minced meat with 3 kGy reduced the counts of total bacteria, coliform bacteria, Staphylococcus aureus and yeasts and molds counts as well as eliminating Salmonella spp. Irradiation doses of 6 and 9 kGy completely eliminated coliform bacteria, Staphylococcus aureus and yeasts and molds. Also, type of packaging materials which used had no effect on counts of all studied microorganisms. Irradiation of minced meat with

  16. ESR (Electronic Spin Resonance Spectroscopy) study of irradiated paper for biomedical material wrapping

    International Nuclear Information System (INIS)

    Huarte, Monica; Rubin de Celis, Emilio; Kairiyama, Eulogia; Zapata, Miguel; Santoro, Natalia; Magnavacca, Cecilia

    2009-01-01

    Ionising radiation treatments are used for sterilization, microbiological decontamination, disinfection, insect disinfestation and food preservation. This ionising radiation generates free radicals (FR) in matter, which can be detected by Electronic Spin Resonance Spectroscopy (ESR). For this work it had analysed different kind of irradiated package papers of syringes, surgical gloves and dressings by ESR. These were irradiated with doses between 20 and 35 kGy of gamma radiation (Cobalt 60). The processed samples were measured in a Bruker ECS 106 spectrometer. The obtained results were: 1-) The irritated samples showed a central peak and two satellites induced by the applied radiation; 2-) The non-irradiated samples did not show the characteristic satellite peaks of the irritated ones; 3-) A linear relationship between the signal heights per unit mass and the applied doses was found; and 4-) The signals were highly stable, with half-time values between 240 and 370 days for 20 and 30 kGy, permitting more than one year of monitoring proceedings. In conclusion, the ESR allows the detection, quantification and time monitoring processes of this kind of irradiated materials. (author) [es

  17. Mechanical properties of organic composite materials irradiated with 2 MeV electrons

    International Nuclear Information System (INIS)

    Egusa, S.; Kirk, M.A.; Birtcher, R.C.; Argonne National Lab., IL; Hagiwara, M.; Kawanishi, S.

    1983-01-01

    Four kinds of cloth-filled organic composites (filter: glass or carbon fiber; matrix; epoxy or polyimide resin) were irradiated with 2 MeV electrons at room temperature, and were examined with regard to the mechanical properties. Following irradiation the Young's (tensile) modulus of these composites remains practically unchanged even after irradiation up to 15.000 Mrad. The shear modulus and the ultimate strength, on the other hand, begin to decrease after the absorbed dose reaches about 2.000 Mrad for the glass/epoxy composite and about 5.000-10.000 Mrad for the other composites. This result is ascribed to the decrease in the capacity of load transfer from the matrix to the fiber due to the radiation damage at the interface, and the dose dependence is interpreted and formulated based on the mechanics of composite materials and the target theory used in radiation biology. As to the fracture behavior, the propagation energy increases from the beginning of irradiation. This result is attributed to the radiation-induced decrease in the bonding energy at the interface. (orig.)

  18. LAMI - a planned Brazilian facility to investigate the mechanical and physical properties of structural materials under irradiation

    International Nuclear Information System (INIS)

    Andrade, Arnaldo H.P.; Lobo, Raquel M.

    2011-01-01

    The LAMI (Laboratorio de Materiais Irradiados) is a hot laboratory designed to the characterization of irradiated structural material and will constitute one of the main installations of the Brazilian Multipurpose Reactor (RMB). The strong points of LAMI are: to contribute, through theoretical and experimental investigations, to the development of knowledge in materials science in order to be able to predict the evolution of the physical and mechanical material properties under service conditions (irradiation, thermomechanical solicitation, influence of the environment, etc); to characterize the properties of the materials used in the nuclear industry in order to determine their performance and to be able to predict their life expectancy; to establish, maintain and make use of the database generated by these data and to provide expertise on industrial components, in particular to investigate strain or rupture mechanisms. The test materials can be irradiated or not, and originate from surveillance programs, experimental neutron irradiations or simulated irradiation with charged particles. The main line of LAMI will have 10 shielded hot cells. The building also will have an area dedicated to micro and nano structural materials analysis. The mechanical characterization to be carried out within LAMI includes mechanical tests on irradiated materials, comprehension of behavior and damage processes and the incorporation of the test data results in a data bank for capitalization of test results. Planned materials to be tested are going to be metallic alloys used in industrial and experimental reactor: pressure vessel steels, internal stainless steels, austeno-ferritic steels, zirconium alloys and aluminum alloys. (author)

  19. Design and Demonstration of a Material-Plasma Exposure Target Station for Neutron Irradiated Samples

    International Nuclear Information System (INIS)

    Rapp, Juergen; Aaron, A. M.; Bell, Gary L.; Burgess, Thomas W.; Ellis, Ronald James; Giuliano, D.; Howard, R.; Kiggans, James O.; Lessard, Timothy L.; Ohriner, Evan Keith; Perkins, Dale E.; Varma, Venugopal Koikal

    2015-01-01

    Fusion energy is the most promising energy source for the future, and one of the most important problems to be solved progressing to a commercial fusion reactor is the identification of plasma-facing materials compatible with the extreme conditions in the fusion reactor environment. The development of plasma-material interaction (PMI) science and the technology of plasma-facing components are key elements in the development of the next step fusion device in the United States, the so-called Fusion Nuclear Science Facility (FNSF). All of these PMI issues and the uncertain impact of the 14-MeV neutron irradiation have been identified in numerous expert panel reports to the fusion community. The 2007 Greenwald report classifies reactor plasma-facing materials (PFCs) and materials as the only Tier 1 issues, requiring a ''. . . major extrapolation from the current state of knowledge, need for qualitative improvements and substantial development for both the short and long term.'' The Greenwald report goes on to list 19 gaps in understanding and performance related to the plasma-material interface for the technology facilities needed for DEMO-oriented R&D and DEMO itself. Of the 15 major gaps, six (G7, G9, G10, G12, G13) can possibly be addressed with ORNL's proposal of an advanced Material Plasma Exposure eXperiment. Establishing this mid-scale plasma materials test facility at ORNL is a key element in ORNL's strategy to secure a leadership role for decades of fusion R&D. That is to say, our end goal is to bring the ''signature facility'' FNSF home to ORNL. This project is related to the pre-conceptual design of an innovative target station for a future Material-Plasma Exposure eXperiment (MPEX). The target station will be designed to expose candidate fusion reactor plasma-facing materials and components (PFMs and PFCs) to conditions anticipated in fusion reactors, where PFCs will be exposed to dense high-temperature hydrogen plasmas providing steady-state heat fluxes of

  20. Similarities in the induction of synthesis of a cell-surface polypeptide in Arthrobacter sp. by near-UV irradiation and photodynamic conditions

    International Nuclear Information System (INIS)

    Hoober, J.K.; Franzi, J.

    1983-01-01

    Irradiation of aerobic suspensions of Arthrobacter sp. with near-UV light (310-400 nm) induced synthesis of a 21 000 dalton, cell-surface polypeptide. Synthesis of this polypeptide also was induced by visible light in the presence of photodynamic dyes. Induction of the polypeptide in ear-UV light and with visible light plus dyes was inhibited by histidine. Hemin inhibited induction in near-UV light and in visible light with methylene blue, neutral red and acriflavin, which are cationic dyes, but failed to inhibit induction in visible light with rose bengal, an anionic dye. These results suggested that inhibition by hemin required electrostatically favored interaction between the anionic porphyrin and the sensitizer, and that the near-UV light effect was mediated by a cationic or neutral endogenous sensitizer. The similarities in the responses of the cells to near-UV irradiation and visible light plus dyes suggested that the mechanism of induction under the two conditions was the same. (author)

  1. Design and Demonstration of a Material-Plasma Exposure Target Station for Neutron Irradiated Samples

    Energy Technology Data Exchange (ETDEWEB)

    Rapp, Juergen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Aaron, A. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bell, Gary L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burgess, Thomas W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ellis, Ronald James [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giuliano, D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kiggans, James O. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lessard, Timothy L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ohriner, Evan Keith [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Perkins, Dale E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Varma, Venugopal Koikal [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-20

    Fusion energy is the most promising energy source for the future, and one of the most important problems to be solved progressing to a commercial fusion reactor is the identification of plasma-facing materials compatible with the extreme conditions in the fusion reactor environment. The development of plasma–material interaction (PMI) science and the technology of plasma-facing components are key elements in the development of the next step fusion device in the United States, the so-called Fusion Nuclear Science Facility (FNSF). All of these PMI issues and the uncertain impact of the 14-MeV neutron irradiation have been identified in numerous expert panel reports to the fusion community. The 2007 Greenwald report classifies reactor plasma-facing materials (PFCs) and materials as the only Tier 1 issues, requiring a “. . . major extrapolation from the current state of knowledge, need for qualitative improvements and substantial development for both the short and long term.” The Greenwald report goes on to list 19 gaps in understanding and performance related to the plasma–material interface for the technology facilities needed for DEMO-oriented R&D and DEMO itself. Of the 15 major gaps, six (G7, G9, G10, G12, G13) can possibly be addressed with ORNL’s proposal of an advanced Material Plasma Exposure eXperiment. Establishing this mid-scale plasma materials test facility at ORNL is a key element in ORNL’s strategy to secure a leadership role for decades of fusion R&D. That is to say, our end goal is to bring the “signature facility” FNSF home to ORNL. This project is related to the pre-conceptual design of an innovative target station for a future Material–Plasma Exposure eXperiment (MPEX). The target station will be designed to expose candidate fusion reactor plasma-facing materials and components (PFMs and PFCs) to conditions anticipated in fusion reactors, where PFCs will be exposed to dense high-temperature hydrogen plasmas providing steady

  2. Ablation, surface activation, and electroless metallization of insulating materials by pulsed excimer laser irradiation

    International Nuclear Information System (INIS)

    Lowndes, D.H.; Godbole, M.J.; Pedraza, A.J.

    1993-01-01

    Pulsed-laser irradiation of wide bandgap ceramic substrates, using photons with sub-bandgap energies, activates the ceramic surface for subsequent electroless copper deposition. The copper deposit is confined within the irradiated region when the substrate is subsequently immersed in an electroless copper bath. However, a high laser fluence (typically several j/cm 2 ) and repeated laser shots are needed to obtain uniform copper coverage by this direct-irradiation process. In contrast, by first applying an evaporated SiO x thin film (with x ∼1), laser ablation at quite low energy density (∼0.5 J/cm 2 ) results in re-deposition on the ceramic substrate of material that is catalytic for subsequent electroless copper deposition. Experiments indicate that the re-deposited material is on silicon, on which copper nucleates. Using an SiO x film on a laser-transparent substrate, quite fine (∼12 μm) copper lines can be formed at the boundary of the region that is laser-etched in SiO x . Using SiO x with an absorbing (polycrystalline) ceramic substrate, more-or-less uniform activation and subsequent copper deposition are obtained. In the later case, interactions with the ceramic substrate also may be important for uniform deposition

  3. Induction of materials for mutation breeding of strawberry (FragariaxAnanassa) by gamma irradiation (Phase 2)

    International Nuclear Information System (INIS)

    Le Tien Thanh; Huynh Thi Trung; Pham Van Nhi; Vu Thi Trac

    2016-01-01

    In this study, New Zaeland strawberry runners was propagated in vitro to create clump of buds for Gamma irradiation. The experimental result showed that LD_5_0 was 52 Gy. Basing on the LD_5_0, we selected the 5 doses of 20, 40, 60, 80 and 100 Gy to irradiate in vitro materials for creating the potential mutants. Irradiated materials were propagated continuously in vitro to complete 300 in vitro plants per dose. On farm, ex vitro plants were planted on the spout (from the ground) by hydroponic method with the number of 200 plants per dose. Some mutant characteristics increased gradually toward the increasing of gamma doses as dwarf plants (in the dose of 60 Gy, 80 Gy and 100 Gy), plants had small fruits (in the dose of 60 Gy, 80 Gy, 100 Gy), plants had deformed fruits (in the dose of 20 Gy, 40 Gy, 60 Gy, 80 Gy and 100 Gy). In this study, we selected four mutants with 2 mutation fruits were changed to fruit heart-shape with symbol DT 1 (dose 60 Gy) and DT 2 (dose 80 Gy); 2 mutants were dwarf plants, the ungrown bud, wrinkled leaf, deformed fruits, high sweetness with symbol DN 1 (dose 60 Gy) and DN 2 (dose 80 Gy). (author)

  4. Estimation of irradiation-induced material damage measure of FCM fuel in LWR core

    International Nuclear Information System (INIS)

    Lee, Kyung-Hoon; Lee, Chungchan; Park, Sang-Yoon; Cho, Jin-Young; Chang, Jonghwa; Lee, Won Jae

    2014-01-01

    An irradiation-induced material damage measure on tri-isotropic (TRISO) multi-coating layers of fully ceramic micro-encapsulated (FCM) fuel to replace conventional uranium dioxide (UO 2 ) fuel for existing light water reactors (LWRs) has been estimated using a displacement per atom (DPA) cross section for a FCM fuel performance analysis. The DPA cross sections in 47 and 190 energy groups for both silicon carbide (SiC) and graphite are generated based on the molecular dynamics simulation by SRIM/TRIM. For the selected FCM fuel assembly design with FeCrAl cladding, a core depletion analysis was carried out using the DeCART2D/MASTER code system with the prepared DPA cross sections to evaluate the irradiation effect in the Korean OPR-1000. The DPA of the SiC and IPyC coating layers is estimated by comparing the discharge burnup obtained from the MASTER calculation with the burnup-dependent DPA for each coating layer calculated using DeCART2D. The results show that low uranium loading and hardened neutron spectrum compared to that of high temperature gas-cooled reactor (HTGR) result in high discharge burnup and high fast neutron fluence. In conclusion, it can be seen that the irradiation-induced material damage measure is noticeably increased under LWR operating conditions compared to HTGRs. (author)

  5. Irradiation-induced modification of the material parameters in magnesium-doped lithium niobate

    International Nuclear Information System (INIS)

    Jentjens, Lena

    2010-01-01

    In the framework of this thesis the material properties of lithium niobate are directedly influenced by the irradiation with 3 He ions with an energy of 40 MeV. In the first part the irradiation-induced material changes are intensively studied. Long-time stable changes of the refractive index are measured in the range of up to 6.10 -3 , which depend on the radiation dose and exhibit until now no saturation behaviour. Accompanied is this change by an also dose-dependent deformation as well as a brownish change of color of the crystals. Furthermore a by several orders of magnitude increased electrical dark- and photoconductivity, which depends on the ion dose and exhibits until now also no saturation behaviour. An effect independent on the ion dose is the reduction of the coercive field strength by about 10%. Furthermore it was stated the quantity of the effects not only depends on the absolute dose, but also on the irradiation direction in view of the crystallographic c-axis. The second part of this thesis deals with the generation of microscopic structures in lithium niobate. By an ion microbeam respectively a shiftable slit aperture the fabrication of refractive-index gratings is pursued. Grating with periodicity lengths in the range of 12-160 μm could until now be detected and promise in comparison with photorefractive gratings the advance of larger stability.

  6. Reactivation of X-irradiated cell material during limb regeneration in Urodeles Amphibians

    International Nuclear Information System (INIS)

    Desselle, J.C.

    1979-10-01

    In amputated members irradiated with X-rays the regeneration power is inhibited. This power is restored by grafts of healthy tissue in the irradiated members. The origin of the cell material of the restored regeneration blastema has been studied by an original labelling technique. The different amounts of DNA in the graft cells and those of the stump mark the graft cells during the regeneration process. It was shown that the graft causes a reactivation of the inhibited stump cells and the reactivation stages are the same as the activation stages of the member regenerating normally. It was also established that during restored regeneration the cell material implanted in the irradiated members contributes, by the 160th day of regeneration, 4.5% of the cartilaginous regenerate cells and 12% of the muscle cells. All the other regenerate cells are supplied by the cells of the stump; these are reactivated and together with the activated graft cells lead to the restitution of the amputated member [fr

  7. Reactor similarity for plasma–material interactions in scaled-down tokamaks as the basis for the Vulcan conceptual design

    International Nuclear Information System (INIS)

    Whyte, D.G.; Olynyk, G.M.; Barnard, H.S.; Bonoli, P.T.; Bromberg, L.; Garrett, M.L.; Haakonsen, C.B.; Hartwig, Z.S.; Mumgaard, R.T.; Podpaly, Y.A.

    2012-01-01

    Highlights: ► Discussion of similarity scalings for reduced-size tokamaks. ► Proposal of a new set of scaling laws for divertor similarity. ► Discussion of how the new scaling provides fidelity to a reactor. ► The new scaling is used as the basis for the Vulcan conceptual design. - Abstract: Dimensionless parameter scaling techniques are a powerful tool in the study of complex physical systems, especially in tokamak fusion experiments where the cost of full-size devices is high. It is proposed that dimensionless similarity be used to study in a small-scale device the coupled issues of the scrape-off layer (SOL) plasma, plasma–material interactions (PMI), and the plasma-facing material (PFM) response expected in a tokamak fusion reactor. Complete similarity is not possible in a reduced-size device. In addition, “hard” technological limits on the achievable magnetic field and peak heat flux, as well as the necessity to produce non-inductive scenarios, must be taken into account. A practical approach is advocated, in which the most important dimensionless parameters are matched to a reactor in the reduced-size device, while relaxing those parameters which are far from a threshold in behavior. “Hard” technological limits are avoided, so that the reduced-size device is technologically feasible. A criticism on these grounds is offered of the “P/R” model, in which the ratio of power crossing the last closed flux surface (LCFS), P, to the device major radius, R, is held constant. A new set of scaling rules, referred to as the “P/S” scaling (where S is the LCFS area) or the “PMI” scaling, is proposed: (i) non-inductive, steady-state operation; (ii) P is scaled with R 2 so that LCFS areal power flux P/S is constant; (iii) magnetic field B constant; (iv) geometry (elongation, safety factor q * , etc.) constant; (v) volume-averaged core density scaled as n≈n ¯ e ∼R −2/7 ; and (vi) ambient wall material temperature T W,0 constant. It is

  8. New electron beam facility for irradiated plasma facing materials testing in hot cell

    International Nuclear Information System (INIS)

    Sakamoto, N.; Kawamura, H.; Akiba, M.

    1995-01-01

    Since plasma facing components such as the first wall and the divertor for the next step fusion reactors are exposed to high heat loads and high energy neutron flux generated by the plasma, it is urgent to develop of plasma facing components which can resist these. Then, we have established electron beam heat facility (open-quotes OHBISclose quotes, Oarai Hot-cell electron Beam Irradiating System) at a hot cell in JMTR (Japan Materials Testing Reactor) hot laboratory in order to estimate thermal shock resistivity of plasma facing materials and heat removal capabilities of divertor elements under steady state heating. In this facility, irradiated plasma facing materials (beryllium, carbon based materials and so on) and divertor elements can be treated. This facility consists of an electron beam unit with the maximum beam power of 50kW and the vacuum vessel. The acceleration voltage and the maximum beam current are 30kV (constant) and 1.7A, respectively. The loading time of electron beam is more than 0.1ms. The shape of vacuum vessel is cylindrical, and the mainly dimensions are 500mm in inner diameter, 1000mm in height. The ultimate vacuum of this vessel is 1 x 10 -4 Pa. At present, the facility for thermal shock test has been established in a hot cell. And performance estimation on the electron beam is being conducted. Presently, the devices for heat loading tests under steady state will be added to this facility

  9. New electron beam facility for irradiated plasma facing materials testing in hot cell

    International Nuclear Information System (INIS)

    Shimakawa, S.; Akiba, M.; Kawamura, H.

    1996-01-01

    Since plasma facing components such as the first wall and the divertor for the next step fusion reactors are exposed to high heat loads and high energy neutron flux generated by the plasma, it is urgent to develop plasma facing components which can resist these. We have established electron beam heat facility ('OHBIS', Oarai hot-cell electron beam irradiating system) at a hot cell in JMTR (Japan materials testing reactor) hot laboratory in order to estimate thermal shock resistivity of plasma facing materials and heat removal capabilities of divertor elements under steady state heating. In this facility, irradiated plasma facing materials (beryllium, carbon based materials and so on) and divertor elements can be treated. This facility consists of an electron beam unit with the maximum beam power of 50 kW and the vacuum vessel. The acceleration voltage and the maximum beam current are 30 kV (constant) and 1.7 A, respectively. The loading time of the electron beam is more than 0.1 ms. The shape of vacuum vessel is cylindrical, and the main dimensions are 500 mm in inside diameter, 1000 mm in height. The ultimate vacuum of this vessel is 1 x 10 -4 Pa. At present, the facility for the thermal shock test has been established in a hot cell. The performance of the electron beam is being evaluated at this time. In the future, the equipment for conducting static heat loadings will be incorporated into the facility. (orig.)

  10. Study on dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment

    International Nuclear Information System (INIS)

    Abe, K.; Kohyama, A.; Namba, C.; Wiffen, F.W.; Jones, R.H.

    2001-01-01

    A Japan-USA Program of irradiation experiments for fusion research, 'JUPITER', has been established as a 6 year program from 1995 to 2000. The goal is to study the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment using fission reactors. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects. The irradiation capsules for in-situ measurement and varying temperature were developed successfully. It was found that insulating ceramics were worked up to 3 dpa. The property changes and related issues in low activation structural materials were summarized. (author)

  11. On possibility of high energy electron irradiation usage for material alloying

    International Nuclear Information System (INIS)

    Vladimirskij, R.A.; Livshits, V.B.; Payuk, V.A.; Plotnikov, S.V.; Kuz'minykh, V.A.

    1988-01-01

    Review of papers concerning over 2.5 MeV fast electron beam (FEB) irradiation of metals and semiconductors is made. It makes possible to transform physical and mechanical properties ofsurface layers due to their alloyage with different elements or due to redistribution of alloy impurities at the essential depth. It is shown, that electron beam irradiation of materials results in the formation of essential temperature gradient in the sample near the surface and defect nonequilibrium concentration. Along with the increase of diffusion effective ratio the heterogeneous distribution of temperature and defects results in the formation of atom nucompensated fluxes within the sample, which result in element redistribution. Drift of one element through the layer of the second one occurs as a result. Gradient of temperature and defects, amfient temperature and correlation of migration activation energies of element atoms are considered as determining factors at anomalous mass transfer

  12. Microbial decontamination of cosmetic raw materials and personal care products by irradiation

    International Nuclear Information System (INIS)

    Katusin-Razem, B.; Mihaljevic, B.; Razem, D.

    2005-01-01

    Typical levels of sporadically occurring (dynamic) microbial contamination of cosmetic raw materials: pigments, abrasives and liposomes, as well as of final products for personal care, i.e. toothpaste, crayons, shampoos, cleansers and creams, were evaluated. In most cases, contamination was dominated by a single population of microorganisms, either Gram-negative bacteria or molds. The feasibility of microbial decontamination by irradiation was studied by determining the resistance to gamma radiation of contaminating microflora in situ. It was expressed as a dose required for the first 90% reduction, D first 9 0% red. The values in the range 1-2 kGy for molds and 0.1-0.6 kGy for Gram-negative bacteria were obtained. This relatively high susceptibility to irradiation allowed inactivation factors close to 6 to be achieved with doses generally not exceeding 3 kGy, and yielding endpoint contamination less than 10 g -1 . (author)

  13. A spallation-based irradiation test facility for fusion and future fission materials

    International Nuclear Information System (INIS)

    Samec, K.; Fusco, Y.; Kadi, Y.; Luis, R.; Romanets, Y.; Behzad, M.; Aleksan, R.; Bousson, S.

    2014-01-01

    The EU's FP7 TIARA program for developing accelerator-based facilities has recently demonstrated the unique capabilities of a compact and powerful spallation source for irradiating advanced nuclear materials. The spectrum and intensity of the neutron flux produced in the proposed facility fulfils the requirements of the proposed DEMO fusion reactor, ADS reactors and also Gen III / IV reactors. Test conditions can be modulated, covering temperature from 400 to 550 deg. C, liquid metal corrosion, cyclical or static stress up to 500 MPa and neutron/proton irradiation damage of up to 25 DPA per annum over a volume occupying one litre. The entire 'TMIF' facility fits inside a cube 2 metres on a side, and is dimensioned for an accelerator beam power of 100 kW, thus reducing costs and offering great versatility and flexibility. (authors)

  14. Microbial decontamination of cosmetic raw materials and personal care products by irradiation

    International Nuclear Information System (INIS)

    Katusin-Razem, Branka; Mihaljevic, Branka; Razem, D.

    2003-01-01

    Typical levels of sporadically occurring (dynamic) microbial contamination of cosmetic raw materials: pigments, abrasives and liposomes, as well as of final products for personal care: toothpaste, crayons, shampoos, cleansers and creams, were evaluated. In most cases the contamination was dominated by a single population of microorganisms, either Gram-negative bacteria or molds. The feasibility of microbial decontamination by irradiation was studied by determining the resistance to gamma radiation of contaminating microflora in situ. It was expressed as a dose required for the first 90% reduction, D first 9 0% r ed . The values in the range 1-2 kGy for molds and 0.1-0.6 kGy for Gram-negative bacteria were obtained. This relatively high susceptibility to irradiation allowed inactivation factors close to 6 to be achieved with doses generally not exceeding 3 kGy, and yielding endpoint contamination less than 10/g

  15. A spallation-based irradiation test facility for fusion and future fission materials

    CERN Document Server

    Samec, K; Kadi, Y; Luis, R; Romanets, Y; Behzad, M; Aleksan, R; Bousson, S

    2014-01-01

    The EU’s FP7 TIARA program for developing accelerator-based facilities has recently demonstrated the unique capabilities of a compact and powerful spallation source for irradiating advanced nuclear materials. The spectrum and intensity of the neutron flux produced in the proposed facility fulfils the requirements of the DEMO fusion reactor for ITER, ADS reactors and also Gen III / IV reactors. Test conditions can be modulated, covering temperature from 400 to 550°C, liquid metal corrosion, cyclical or static stress up to 500 MPa and neutron/proton irradiation damage of up to 25 DPA per annum. The entire “TMIF” facility fits inside a cube 2 metres on a side, and is dimensioned for an accelerator beam power of 100 kW, thus reducing costs and offering great versatility and flexibility.

  16. Optical transmittance investigation of 1-keV ion-irradiated sapphire crystals as potential VUV to NIR window materials of fusion reactors

    Directory of Open Access Journals (Sweden)

    Keisuke Iwano

    2016-10-01

    Full Text Available We investigate the optical transmittances of ion-irradiated sapphire crystals as potential vacuum ultraviolet (VUV to near-infrared (NIR window materials of fusion reactors. Under potential conditions in fusion reactors, sapphire crystals are irradiated with hydrogen (H, deuterium (D, and helium (He ions with 1-keV energy and ∼ 1020-m-2 s-1 flux. Ion irradiation decreases the transmittances from 140 to 260 nm but hardly affects the transmittances from 300 to 1500 nm. H-ion and D-ion irradiation causes optical absorptions near 210 and 260 nm associated with an F-center and an F+-center, respectively. These F-type centers are classified as Schottky defects that can be removed through annealing above 1000 K. In contrast, He-ion irradiation does not cause optical absorptions above 200 nm because He-ions cannot be incorporated in the crystal lattice due to the large ionic radius of He-ions. Moreover, the significant decrease in transmittance of the ion-irradiated sapphire crystals from 140 to 180 nm is related to the light scattering on the crystal surface. Similar to diamond polishing, ion irradiation modifies the crystal surface thereby affecting the optical properties especially at shorter wavelengths. Although the transmittances in the VUV wavelengths decrease after ion irradiation, the transmittances can be improved through annealing above 1000 K. With an optical transmittance in the VUV region that can recover through simple annealing and with a high transparency from the ultraviolet (UV to the NIR region, sapphire crystals can therefore be used as good optical windows inside modern fusion power reactors in terms of light particle loadings of hydrogen isotopes and helium.

  17. Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

    International Nuclear Information System (INIS)

    Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

    2009-01-01

    The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

  18. 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)

  19. Investigation of electrophysical properties of electrical insulating materials under neutron irradiation

    International Nuclear Information System (INIS)

    Skornyakov, Yu.A.; Stepanov, A.N.; Lapenas, A.A.

    1978-01-01

    The possibilities of applicaiton of insulating materials on the basis of glass cloths in electric windings for operation under neutron radiation of thermonuclear devices are studied. Changes in the specimen resistance, tangent of the angle of dielectric losses, electric strength according to the value of neutron fluence are determined. The temperature regimes are also studied. The data indicate the irreversible changes in the composition and structure of the polymer material under irradiation. The LSMI 228L-80 glass cloth has the highest radiation resistance. The necessity of forced cooling of large-sized specimens under the neutron radiation the IRT-200 reactor is established. The presence of impurities leading to the long-term induced activity of the insulating materials ( 59 Fe, 60 Co) is determined

  20. Reduction of beta-interference in gamma-spectrometric measurements of neutron-irradiated geological material

    International Nuclear Information System (INIS)

    Garmann, L.B.

    1986-01-01

    The analytical technique for INAA, when applied to geological materials, is improved by introducing an electromagnetic field between sample and detector. This field lowers the bremsstrahlung background intensity in the gamma-spectrum by reducing the number of beta-particles reaching the detector. Thus precision, accuracy and lower detection limit are improved. The technique was used on alkalisyenite and on meteoritic material, rocks containing high quantities of sodium and iron, respectively. After neutron irradiation, the induced nuclides sup(24)Na and sup(59)Fe are responsible for high bremsstrahlung interference, which under normal analyitical conditions would mask any x-ray or gamma-ray peaks of interest. The technique is easily applied to multielement analysis of geological and biological materials. It can be combined with sophisticated spectrum-treating techniques such as spectrum stripping and spectrum smoothing, or coincidence-anticoincidence circuits. (author)

  1. Erosion and mass transfer of Mo, W and Nb under neutron irradiation of high temperature materials

    International Nuclear Information System (INIS)

    Berzhatyj, V.I.; Luk'yanov, A.N.; Zavalishin, A.A.; Tkach, V.N.; Fedorenko, A.I.

    1980-01-01

    Studies have been made of the medium composition in thermionic fuel elements of two types during reactor tests; erosion and mass transfer of electrode materials have been investigated in the after-reactor analysis of the tested fuel elements. The studies of electrode material evaporation at the conditions approaching (in environment temperature and composition) those of reactor tests of thermionic fuel elements have shown that the process proceeds in the form of metal oxides. Evaporation rates are determined, the mechanism of evaporation is discussed, and the analytical dependences are obtained for calculating the evaporation rates of Mo and W at certain temperature and gaseous medium composition. It is found that the main contribution to the material transfer off the Mo and Nb surfaces under a high-temperature reactor irradiation comes through the thermal evaporation; in the case of tungsten at the same experimental conditions the rates of mass transfer due to thermal evaporation and neutron sputtering are nearly the same [ru

  2. Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor

    International Nuclear Information System (INIS)

    Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon

    2005-01-01

    A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams

  3. Characterization of the proton irradiation induced luminescence of materials and application in radiation oncology dosimetry

    Science.gov (United States)

    Darafsheh, Arash; Zhang, Rongxiao; Kassaee, Alireza; Finlay, Jarod C.

    2018-03-01

    Visible light generated as the result of interaction of ionizing radiation with matter can be used for radiation therapy quality assurance. In this work, we characterized the visible light observed during proton irradiation of poly(methyl methacrylate) (PMMA) and silica glass fiber materials by performing luminescence spectroscopy. The spectra of the luminescence signal from PMMA and silica glass fibers during proton irradiation showed continuous spectra whose shape were different from that expected from Čerenkov radiation, indicating that Čerenkov radiation cannot be the responsible radioluminescence signal. The luminescence signal from each material showed a Bragg peak pattern and their corresponding proton ranges are in agreement with measurements performed by a standard ion chamber. The spectrum of the silica showed two peaks at 460 and 650 nm stem from the point defects of the silica: oxygen deficiency centers (ODC) and non-bridging oxygen hole centers (NBOHC), respectively. The spectrum of the PMMA fiber showed a continuous spectrum with a peak at 410 nm whose origin is connected with the fluorescence of the PMMA material. Our results are of interest for various applications based on imaging radioluminescent signal in proton therapy and will inform on the design of high-resolution fiber probes for proton therapy dosimetry.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  5. Nuclear data for the production of radioisotopes in fusion materials irradiation facility

    International Nuclear Information System (INIS)

    Cheng, E.T.; Schenter, R.E.; Mann, F.M.; Ikeda, Y.

    1991-01-01

    The fusion materials irradiation facility (FMIF) is a neutron source generator that will produce a high-intensity 14-MeV neutron field for testing candidate fusion materials under reactor irradiation conditions. The construction of such a facility is one of the very important development stages toward realization of fusion energy as a practical energy source for electricity production. As a result of the high-intensity neutron field, 10 MW/m 2 or more equivalent neutron wall loading, and the relatively high-energy (10- to 20-MeV) neutrons, the FMIF, as future fusion reactors, also bears the potential capability of producing a significant quantity of radioisotopes. A study is being conducted to identify the potential capability of the FMIF to produce radioisotopes for medical and industrial applications. Two types of radioisotopes are involved: one is already available; the second might not be readily available using conventional production methods. For those radioisotopes that are not readily available, the FMIF could develop significant benefits for future generations as a result of the availability of such radioisotopes for medical or industrial applications. The current production of radioisotopes could help finance the operation of the FMIF for irradiating the candidate fusion materials; thus this concept is attractive. In any case, nuclear data are needed for calculating the neutron flux and spectrum in the FMIF and the potential production rates of these isotopes. In this paper, the authors report the result of a preliminary investigation on the production of 99 Mo, the parent radioisotope for 99m Tc

  6. A new device for X-ray Diffraction analyses of irradiated materials

    International Nuclear Information System (INIS)

    Valot, Christophe; Blay, Thierry; Caillot, Laurent; Ferroud-Plattet, Marie Pierre

    2008-01-01

    A new X-Ray Diffraction (XRD) equipment is being implemented in the LECA (Cea - Cadarache) hot laboratory. The device will be dedicated to structural characterization on irradiated fuels, as PWR fuels, transmutation targets and innovative fuels. The paper will present the specific design that was decided in order to reduce the number of components in contaminated volume and to make servicing easier. The analytical performances of this new equipment will be illustrated on some model samples: -) micro-diffraction capabilities will be detailed on heterogeneous material; -) strain and stress analyses on fresh uranium oxide pellets. (authors)

  7. Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

    Science.gov (United States)

    Gelles, D. S.

    1990-05-01

    Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development.

  8. Impacts of damage production and accumulation on materials performance in irradiation environment

    DEFF Research Database (Denmark)

    Singh, B.N.

    1998-01-01

    and needs to be extended to temperatures below stage V and to materials of practical interests. This requires, however, that the information regarding the effects of alloying elements and impurity atoms on the nature of the primary damage state are available from molecular dynamics and kinetic Monte Carlo...... the damage accumulation, irradiation hardening and the loss of ductility. The recently developed production bias model together with one-dimensional glide of interstitial clusters produced in the cascades has been shown to describe the damage accumulation at temperatures above stage V for pure metals...... type of simulations. (C) 1998 Elsevier Science B.V. All rights reserved....

  9. Corrosion of copper-based materials in irradiated moist air systems

    International Nuclear Information System (INIS)

    Reed, D.T.; Van Konynenburg, R.A.

    1991-06-01

    The atmospheric corrosion of oxygen-free copper (CDA-102), 70/30 copper-nickel (CDA-715), and 7% aluminum bronze (CDA-613) in an irradiated moist air environment was investigated. Experiments were performed in both dry and 40% RH (at sign 90 degree C) air at temperatures of 90 and 150 degree C. Initial corrosion rates were determined based on a combination of weight gain and weight loss measurements. Corrosion products observed were identified. These experiments support efforts by the Yucca Mountain Project (YMP) to evaluate possible metallic barrier materials for nuclear waste containers. 8 refs., 1 fig., 2 tabs

  10. Non-destructive diagnostics of irradiated materials using neutron scattering from pulsed neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Korenev, Sergey E-mail: sergey_korenev@steris.com; Sikolenko, Vadim

    2004-10-01

    The advantage of neutron-scattering studies as compared to the standard X-ray technique is the high penetration of neutrons that allow us to study volume effects. The high resolution of instrumentation on the basis neutron scattering allows measurement of the parameters of lattice structure with high precision. We suggest the use of neutron scattering from pulsed neutron sources for analysis of materials irradiated with pulsed high current electron and ion beams. The results of preliminary tests using this method for Ni foils that have been studied by neutron diffraction at the IBR-2 (Pulsed Fast Reactor at Joint Institute for Nuclear Research) are presented.

  11. Non-destructive diagnostics of irradiated materials using neutron scattering from pulsed neutron sources

    Science.gov (United States)

    Korenev, Sergey; Sikolenko, Vadim

    2004-09-01

    The advantage of neutron-scattering studies as compared to the standard X-ray technique is the high penetration of neutrons that allow us to study volume effects. The high resolution of instrumentation on the basis neutron scattering allows measurement of the parameters of lattice structure with high precision. We suggest the use of neutron scattering from pulsed neutron sources for analysis of materials irradiated with pulsed high current electron and ion beams. The results of preliminary tests using this method for Ni foils that have been studied by neutron diffraction at the IBR-2 (Pulsed Fast Reactor at Joint Institute for Nuclear Research) are presented.

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

    International Nuclear Information System (INIS)

    Dietz, W.; Breitling, H.

    1995-01-01

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

  13. Radiation effects on polymer materials. Ionizing radiation induces degradation or improvement? (2) Gas evolution by irradiation

    International Nuclear Information System (INIS)

    Sasuga, Tsuneo

    2005-01-01

    The present article reviews gas evolution from organic polymers induced by ionizing radiations, focusing on gamma-ray irradiation of PE (polyethylene) and PP (polypropylene)-model compounds at temperatures from -77 to 55degC. In the polyolefins, the main gas evolved by irradiation is hydrogen with G-value of 3-4 at room temperatures and G(H 2 ) is 1.8 at 77K. For PE, G(H 2 ) is higher for the low-density PE than for higher-density PE. For the halogenated polymers as PVC, etc., evolved gas is hydrogen halogenated: G(HCl)=6.8 for PVC. For the case where the irradiation is accompanied with the oxidation of polymers, the de-oxygenation and formation of carboxylic radicals are remarkably high and known to emit a bad smell which depends on the thickness of oxidized layers. In conclusion, the gas evolution can be estimated by considering the molecular structure of polymer materials. (S.Ohno)

  14. Effectiveness of shield materials in the design of the PFBR irradiated fuel subassembly shipping cask

    International Nuclear Information System (INIS)

    Radhakrishnan, G.

    2003-01-01

    Fuel subassemblies are irradiated inside the reactor core till they achieve the required burn up and after that they are cooled to permissible decay power level in in-vessel and ex-vessel storage places. Subsequently they are transported to reprocessing plants by means of shipping casks. Shield for the shipping cask has to be designed such a way that it has to comply with the ICRP recommended dose levels of less than 2 mSv/h on contact at the outer surface of the cask and less than 100 mSv/h at 1 m distance from the outer surface of the cask. In this paper, shield design of a typical PFBR irradiated fuel subassembly, which can transport three subassemblies at a time, is narrated. Considering the neutron and fission product and induced gamma rays emitted by typical PFBR irradiated core central subassembly subjected to a maximum burn up, as the source term shield design optimizations have been done. One-dimensional discrete ordinates transport theory computer code ANISN and point kernel computer code QAD-CGGP have been used in complement to carry out the shield design optimizations. Cast-iron, carbon steel, stainless steel 304 and lead and permali have been considered for shield materials. Shield requirements on top, bottom and along the axial height of the shipping cask have also been estimated. (author)

  15. Computational science simulation of laser materials processing and provision of their irradiation conditions

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu

    2016-01-01

    In laser processing, it is necessary for achieving the intended performance and product, to understand the complex physical courses including melting and solidification phenomena occurring in laser processing, and thus to set proper laser irradiation conditions. This condition optimization work requires an enormous amount of overhead due to repeated efforts, and has become a cause for inhibiting the introduction of laser processing technology into the industrial field that points to the small lot production of many products. JAEA tried to make it possible to quantitatively handle the complex physical course from the laser light irradiation to the fabricating material until the completion of processing, and is under development of the computational science simulation code SPLICE that connects micro behavior and macro behavior through a multi-level scale model. This SPLICE is able to visualize the design space and to reduce the overhead associated with the setting of laser irradiation conditions and the like, which gives the prospect of being effective as a tool for front-loading. This approach has been confirmed to be effective for the welding and fusing process. (A.O.)

  16. Practice of adding value to materials by a neutron irradiation in research reactors

    International Nuclear Information System (INIS)

    Oh, Soo Youl; Jun, Byung Jin

    2008-01-01

    Adding more value to materials by a neutron irradiation is one of the utilization areas of research reactors. A neutron transmutation doping (NTD) of Si, a gemstone coloration, and a track etched membrane production are the three important subjects and these are being serviced on a commercial basis. Nevertheless, a further enhancement of the technologies is necessary, and the IAEA/RCA has supported the development and dissemination of the relevant technologies. This article reports on what was achieved during the IAEA/RCA Regional Training Course on the Design and Operation of Neutron Irradiation Facilities which was held in Korea in April 2008 for two weeks. The Course, which was open to the Asian RCA member states, consisted of lectures, presentations from each country, and design and experiment exercises concerning the above three subjects. The lectures covered not only the principles but also the practice in every detail. The experience of Korea on the NTD and those of Indonesia and Thailand on the gemstone coloration drew concentrated attention of the participants. Meanwhile, as one of the exercises the participants designed a device for the NTD or the gemstone irradiation and analyzed its performance from the neutronics point of view. The Course was successful in deepening the understanding on the practice of value adding technologies and also in sharing some ideas for their enhancement

  17. Practice of adding value to materials by a neutron irradiation in research reactors

    International Nuclear Information System (INIS)

    Oh, Soo-Youl; Jun, Byung Jin

    2008-01-01

    Adding more value to materials by a neutron irradiation is one of the utilization areas of research reactors. A neutron transmutation doping (NTD) of Si, a gemstone coloration, and a track-etched membrane production are the three important subjects and these are being serviced on a commercial basis. Nevertheless, a further enhancement of the technologies is necessary, and the IAEA/RCA has supported the development and dissemination of the relevant technologies. This article reports on what was achieved during the IAEA/RCA Regional Training Course on the Design and Operation of Neutron Irradiation Facilities which was held in Korea in April 2008 for two weeks. The Course, which was open to the Asian RCA member states, consisted of lectures, presentations from each country, and design and experiment exercises concerning the above three subjects. The lectures covered not only the principles but also the practice in every detail. The experience of Korea on the NTD and those of Indonesia and Thailand on the gemstone coloration drew concentrated attention of the participants. Meanwhile, as one of the exercises the participants designed a device for the NTD or the gemstone irradiation and analyzed its performance from the neutronics point of view. The Course was successful in deepening the understanding on the practice of value-adding technologies and also in sharing some ideas for their enhancement. (author)

  18. Practice of adding value to materials by a neutron irradiation in research reactors

    International Nuclear Information System (INIS)

    Oh, Soo Youl; Jun, Byung Jin

    2008-01-01

    Adding value to materials by a neutron irradiation is one of the utilization areas of research reactors. A neutron transmutation doping (NTD) of Si, a gemstone coloration, and a track-etched membrane production are the three important subjects and these are being serviced on a commercial basis. Nevertheless, a further enhancement of the technologies is necessary, and the IAEA/RCA has supported the development and dissemination of the relevant technologies. This article report on what was achieved during the IAEA/RCA regional Training Course on the Design and Operation of Neutron Irradiation Facilities which was held in Korea in April 2008 for two weeks. The Course, which was open to the Asian RCA member states, consisted of lectures, presentations from each country, and design and experiment exercises concerning the above three subjects. The lectures covered not only the principles but also the practice in every detail. The experience of Korea on the NTD and those of indonesia and Thailand on the gemstone coloration drew concentrated attention of the participants. Meanwhile, as one of the exercises the participants designed a device for the NTD or the gemstone irradiation and analyzed its performance from the neutronics point of view. The Course was successful in deepening the understanding on the practice of value-adding technologies and also in sharing some ideas for their enhancement

  19. New insight on bubble-void transition effects in irradiated materials

    International Nuclear Information System (INIS)

    Dubinko, V.I.

    1993-01-01

    An account of elastic interaction between cavities and point defects is shown to result in new critical quantities for bubblevoid transition effects in irradiated cubic crystals. In contrast to previous theories, the present one gives not only critical quantities which determine the onset of bias-driven void swelling but the maximum stationary number density and the corresponding mean radius of voids as well as the duration of the bimodal regime. The void density and swelling rate are shown to be independent from the gas level. In the region of low temperatures/high dose rates, the void density appears to be independent from irradiation parameters as well. The relationships among material constants are found at which the stabilization of gas bubbles occurs via the dislocation loop punching mechanism resulting in a drastic change in the cavity behaviour under irradiation such as the saturation (or even suppression) of void swelling and void lattice formation. The theoretical results are compared with experimental data and further experimental tests are proposed. (author). 38 refs., 1 tab., 11 figs

  20. Sink efficiency calculation of dislocations in irradiated materials by phase-field modelling

    International Nuclear Information System (INIS)

    Rouchette, Adrien

    2015-01-01

    The aim of this work is to develop a modelling technique for diffusion of crystallographic migrating defects in irradiated metals and absorption by sinks to better predict the microstructural evolution in those materials.The phase field technique is well suited for this problem, since it naturally takes into account the elastic effects of dislocations on point defect diffusion in the most complex cases. The phase field model presented in this work has been adapted to simulate the generation of defects by irradiation and their absorption by the dislocation cores by means of a new order parameter associated to the sink morphology. The method has first been validated in different reference cases by comparing the sink strengths obtained numerically with analytical solutions available in the literature. Then, the method has been applied to dislocations with different orientations in zirconium, taking into account the anisotropic properties of the crystal and point defects, obtained by state-of-the-art atomic calculations.The results show that the shape anisotropy of the point defects promotes the vacancy absorption by basal loops, which is consistent with the experimentally observed zirconium growth under irradiation. Finally, the rigorous investigation of the dislocation loop case proves that phase field simulations give more accurate results than analytical solutions in realistic loop density ranges. (author)

  1. Light irradiance through novel CAD-CAM block materials and degree of conversion of composite cements.

    Science.gov (United States)

    Lise, Diogo Pedrollo; Van Ende, Annelies; De Munck, Jan; Yoshihara, Kumiko; Nagaoka, Noriyuki; Cardoso Vieira, Luiz Clovis; Van Meerbeek, Bart

    2018-02-01

    To assess light irradiance (LI) delivered by two light-curing units (LCU's) and to measure the degree of conversion (DC) of three composite cements, when cured through different thicknesses of two novel CAD-CAM block materials. 100-μm-thick films of a dual-curable composite cement (G-CEM LinkAce, GC), a light-curable flowable resin-based composite (RBC) (G-ænial Universal Flo, GC) and a micro-hybrid RBC (G-ænial Posterior, GC) were investigated as luting agents. Two 'polymer-ceramic' CAD-CAM blocks (Cerasmart, GC; Enamic, Vita Zahnfabrik) were sectioned in slabs with different thicknesses (1, 3 and 5mm). LI at the bottom of the specimens was measured using a calibrated spectrometer, while being light-cured through the CAD-CAM block slabs for 40s with a low- (±500mW/cm 2 ) or high- (±1,600mW/cm 2 ) irradiance LCU (n=5). After light-curing, micro-Raman spectra of the composite films were acquired to determine DC at 5min, 10min, 1h and 24h. LI data were statistically analyzed by Kruskal-Wallis followed by post-hoc comparisons, while a linear mixed-effect model was applied for the DC analysis. In addition, the CAD-CAM blocks ultrastructure was characterized upon argon-ion slicing using scanning transmission electron microscopy (STEM). Finally, light transmission (LT) through each CAD-CAM block material was assessed using a spectrophotometer. Curing-light attenuation and DC were significantly influenced by thickness and type of the overlying material. LCU only had a significant effect on DC of the micro-hybrid RBC. DC significantly increased over time for all composite cements. CAD-CAM block structural analysis revealed a relatively small and homogenous filler configuration (mean filler size of 0.2-0.5μm) for Cerasmart, while Enamic contained ceramic grains varying in shape and size (1-10μm), which were interconnected by the polymer-based network. LT was much higher at a wavelength range of 300-800nm for Cerasmart than for Enamic. Light-curable composite cements

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  3. Status and possible prospects of an international fusion materials irradiation facility

    International Nuclear Information System (INIS)

    Cozzani, F.

    1999-01-01

    Structural materials for future DT fusion power reactors will have to operate under intense neutron fields with energies up to 14 MeV and fluences in the order of 2 MW/m 2 per year. As environmental acceptability, safety considerations and economic viability will be ultimately the keys to the widespread introduction of fusion power, the development of radiation-resistant and low activation materials would contribute significantly to fusion development. For this purpose, testing of materials under irradiation conditions close to those expected in a fusion power station would require the availability, in an appropriate time framework, of an intense, high-energy neutron source. Recent advances in linear accelerator technology, in small specimens testing technology, and in the comprehension of damage phenomena, lead to the conclusion that an accelerator-based D-Li neutron source, with beam energy variability, would provide the most realistic option for a fusion materials testing facility. Under the auspices of the IEA, an international effort (EU, Japan, US, RF) to carry out the conceptual design activities (CDA) of an international fusion materials irradiation facility (IFMIF), based on the D-Li concept, have been carried out successfully. A final conceptual design report was produced at the end of 1996. A phase of conceptual design evaluation (CDE), presently underway, is extending and further refining some of the conceptual design details of IFMIF. The results indicate that an IFMIF-class installation would be technically feasible and could meet its mission objectives. However, a suitable phase of Engineering Validation, to carry out some complementary R and D and prototyping, would still be needed to resolve a few key technical uncertainties before the possibility to proceed toward detailed design and construction could be explored. (orig.)

  4. Electron beam irradiation to the allogeneic, xenogenic and synthetic bone materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Soung Min; Park, Min Woo; Jeong, Hyun Oh [School of Dentistry Seoul National University, Seoul (Korea, Republic of); and others

    2013-07-01

    For the development of the biocompatible bony regeneration materials, allogenic, xenogenic and synthetic bone were irradiated by electron beam to change the basic components and structures. For the efficient electron beam irradiating condition of these allogenic, xenogenic and artificial bone substitutes, the optimal electron beam energy and their individual dose were established, to maximize the bony regeneration capacity. Commercial products of four allogenic bones, such as Accell (ISOTIS OrthogBiologics Co., USA), Allotis (Korea Bone Bank Co., Korea), Oragraft (LifeNet Co., USA), and Orthoblast (Integra Orthobiologics Inc., USA), six xenogenic bones, such as BBP (OscoTec Co., Korea), Bio-cera (OscoTec Co., Korea), Bio-oss (Geistlich Pharma AG, Switzerland), Indu-cera (OscoTec Co., Korea), OCS-B (Nibec Co., Korea), and OCS-H (Nibec Co., Korea), and six synthetic bones, such as BMP (Couellmedi Co., Korea), BoneMedik (Meta Biomed Co., Korea), Bone plus (Megagen Co., Korea), MBCP (Biomatlante Co., France), Osteon (Genoss Co., Korea), and Osteogen (Impladent LTD., USA), were used. We used 1.0 and 2.0 MeV superconduction accelerator, and/or microtrone with different individual 60, 120 kGy irradiation dose. Different dose irradiated specimens were divided 6 portions each, so total 360 groups were prepared. 4 portions were analyzed each by elementary analysis using FE-SEM (Field Emission Scanning Microscopy) and another 2 portions were grafted to the calvarial defect of Sprague-Dawley rat, following histologic, immunohistochemical analysis and TEM study were processed at the 8th and 16th weeks, in vivo. This work was supported by the National Research Foundation of Korea(NRF) grant funded by the Korea government(MEST)

  5. Application of FE-SEM with elemental analyzer for irradiated fuel materials

    International Nuclear Information System (INIS)

    Sasaki, Shinji; Maeda, Koji; Yamada, A.

    2012-01-01

    It is important to study the irradiation behavior of the uranium-plutonium mixed oxide fuels (MOX fuels) for development of fast reactor fuels. During irradiation in a fast reactor, the changes of microstructures and the changes of element distributions along radial direction occur in the MOX fuels because of a radial temperature gradient. In order to make detailed observations of microstructure and elemental analyses of fuel samples, a field emission scanning electron microscope (FE-SEM) equipped with a wavelength-dispersive X-ray spectrometer (WDX) and an energy-dispersive X-ray spectrometer (EDX) were installed in a hot laboratory. Because fuel samples have high radioactivities and emit α-particles, the instrument was modified correspondingly. The notable modified points were as follows. 1) To prevent leakage of radioactive materials, the instrument was attached to a remote control air-tight sample transfer unit between a shielded hot cell and the FE-SEM. 2) To protect operators and the instruments from radiation, the FE-SEM was installed in a lead shield box and the control unit was separately located outside the box. After the installation, the microscopy and elemental analyses were made on low burnup fuel samples. High resolution images were obtained on the fuel sample surface. The characteristic X-rays (U, Pu) emitted from the fuel sample surface measured along radial direction successfully. Thereby, it was able to grasp the change of U, Pu radial distribution after irradiation. The technique has the great advantage of being able to evaluate the changes of microstructures and the changes of element distributions of MOX fuels due to irradiation. In future work, samples of even higher radioactivity will be observed and analyzed. (author)

  6. Effect of ion-irradiation on the microstructure and microhardness of the W-2Y2O3 composite materials fabricated by sintering and hot forging

    International Nuclear Information System (INIS)

    Battabyal, M.; Spätig, P.; Baluc, N.

    2013-01-01

    Highlights: • W-2Y 2 O 3 material is fabricated using sintering and hot forging method with 99.3 vol.% density. • Microstructure and microhardness of the material after heavy ion irradiation are almost similar irrespective of the sample holder heating temperatures. • Dislocation loops are found on the W grains of irradiated sample where as radiation induced fine voids are observed on yttria particles. • We also observe few radiation loops on yttria particles. • No surface crack at the grain boundary is observed and significant difference in radiation hardening is confirmed. -- Abstract: A W-2Y 2 O 3 material was developed in collaboration with the Plansee Company (Austria). An ingot of the material having approximate dimension of 95 mm × 20 mm was fabricated by mixing the elemental powders followed by pressing, sintering and hot forging. The microstructure of the W-2Y 2 O 3 composite was investigated using transmission electron microscopy (TEM). The microhardness was studied using nano-indentation technique. We observed that the W-grains having a mean size of about 1 μm already formed and these grains contain very low density of dislocations. The size of the yttria particles was between 300 nm and 1 μm and the Berkovich hardness was about 4.8 GPa. The specimens were irradiated/implanted with Fe and He ions at JANNuS facility located at Orsay/Saclay, France. The TEM disks kept were irradiated/implanted at 300 and 700 °C using Fe and He ions with an energy of 24 and 2 MeV, respectively. The calculated radiation dose was about 5 dpa produced by Fe ions and total He content is 75 appm at both 300 and 700 °C. From the TEM investigation of irradiated samples, few radiation loops are present on the W grains, whereas on yttria particles, the radiation induced damages appear as voids. Berkovich hardness of the irradiated sample is higher than that of the non-irradiated sample. Results on the microstructure and microhardness of the ion-irradiated W-2Y 2 O 3

  7. Bacterial Protein Characterization of Streptococcus agalactiae by SDS-page Method for Subclinical Mastitis Irradiated Vaccine Materials in Dairy Cattle

    International Nuclear Information System (INIS)

    Tuasikal, B.J.; Wibawan, I.W.T.; Pasaribu, F.H; Estuningsih, S.

    2012-01-01

    A study have been conducted to isolate and characterize bacterial protein S. agalactiae, which is antigenic and can be used to test immunogenicity of vaccine in order to manufacture irradiated mastitis (inflammation of the udder) vaccine in ruminant. The study aims to determine the Molecular Weight (MW) bacterial protein S. agalactiae irradiation, which can be used to test the nature of its antigenic caharacteristic. The character of S. agalactiae antigenic stimulates antibody induction of the immune system, in which case is the body's defense system against mastitis disease in cattle. In this study, irradiation of gamma ray is used to attenuate the pathogenicity of bacteria by reducing S. agalactiae antigenic characteristic. Previous research, in irradiation dose orientation before antigenic protein isolation of S. agalactiae, indicated that irradiation lethal dose to 50% (LD 50 ) is 17 Gy. The characterization of S. agalactiae bacteria isolate using SDS-page method results in no significance different between irradiated and non-irradiated group, which indicated by MW range 75 - 100 kDa base on marker standard which used, or 99 kDa by the linier equation of Y = 11,60 - 0.05X (where Y = bands distance; X = MW standard protein); r 2 = 0.99. In conclusion, 17 Gy irradiation dose does not impair antigenic property of S. agalactiae and therefore, can be applied to produce base material of irradiated vaccine for mastitis. (author)

  8. First results of the post-irradiation examination of the Ceramic Breeder materials from the Pebble Bed Assemblies Irradiation for the HCPB Blanket concept

    International Nuclear Information System (INIS)

    Hegeman, J.; Magielsen, A.J.; Peeters, M.; Stijkel, M.P.; Fokkens, J.H.; Laan, J.G. van der

    2006-01-01

    In the framework of developing the European Helium Cooled Pebble-Bed (HCPB) blanket an irradiation test of pebble-bed assemblies is performed in the HFR Petten. The experiment is focused on the thermo-mechanical behavior of the HCPB type breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. To achieve representative conditions a section of the HCPB is simulated by EUROFER-97 cylinders with a horizontal bed of ceramic breeder pebbles sandwiched between two beryllium beds. Floating Eurofer-97 steel plates separate the pebble-beds. The structural integrity of the ceramic breeder materials is an issue for the design of the Helium Cooled Pebble Bed concept. Therefore the objective of the post irradiation examination is to study deformation of pebbles and the pebble beds and to investigate the microstructure of the ceramic pebbles from the Pebble Bed Assemblies. This paper concentrates on the Post Irradiation Examination (PIE) of the four ceramic pebble beds that have been irradiated in the Pebble Bed Assembly experiment for the HCPB blanket concept. Two assemblies with Li 4 SiO 4 pebble-beds are operated at different maximum temperatures of approximately 600 o C and 800 o C. Post irradiation computational analysis has shown that both have different creep deformation. Two other assemblies have been loaded with a ceramic breeder bed of two types of Li 2 TiO 3 beds having different sintering temperatures and consequently different creep behavior. The irradiation maximum temperature of the Li 2 TiO 3 was 800 o C. To support the first PIE result, the post irradiation thermal analysis will be discussed because thermal gradients have influence on the pebble-bed thermo-mechanical behavior and as a result it may have impact on the structural integrity of the ceramic breeder materials. (author)

  9. IFMIF-KEP. International fusion materials irradiation facility key element technology phase report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-03-01

    The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m{sup 2}, 20 dpa/y in Fe, in a volume of 500 cm{sup 3} and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration. (author)

  10. IFMIF-KEP. International fusion materials irradiation facility key element technology phase report

    International Nuclear Information System (INIS)

    2003-03-01

    The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m 2 , 20 dpa/y in Fe, in a volume of 500 cm 3 and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration. (author)

  11. Setup for in situ x-ray diffraction study of swift heavy ion irradiated materials.

    Science.gov (United States)

    Kulriya, P K; Singh, F; Tripathi, A; Ahuja, R; Kothari, A; Dutt, R N; Mishra, Y K; Kumar, Amit; Avasthi, D K

    2007-11-01

    An in situ x-ray diffraction (XRD) setup is designed and installed in the materials science beam line of the Pelletron accelerator at the Inter-University Accelerator Centre for in situ studies of phase change in swift heavy ion irradiated materials. A high vacuum chamber with suitable windows for incident and diffracted X-rays is integrated with the goniometer and the beamline. Indigenously made liquid nitrogen (LN2) temperature sample cooling unit is installed. The snapshots of growth of particles with fluence of 90 MeV Ni ions were recorded using in situ XRD experiment, illustrating the potential of this in situ facility. A thin film of C60 was used to test the sample cooling unit. It shows that the phase of the C60 film transforms from a cubic lattice (at room temperature) to a fcc lattice at around T=255 K.

  12. Investigation of corrosion of materials of the irradiation device in the RA reactor

    International Nuclear Information System (INIS)

    Zaric, M.; Mance, A.; Vlajic, M.

    1963-12-01

    Devices for sample irradiation in the vertical RA reactor channels will be made of aluminium alloys. According to the regulations concerned with introducing materials into the RA reactor core, corrosion characterisation of these materials is an obligation. Corrosion properties of four aluminium alloys were investigated both in contact with stainless steel and without it. First part of this report deals with the corrosion testing of aluminium alloys in water by gravimetric and electrochemical methods. Bi-distilled water at temperatures less than 100 deg C was used. Second part is related to aluminium alloys corrosion in carbon dioxide gas under experimental conditions. The second part of research was initiated by the design of the head of the independent CO 2 loop for samples cooling [sr

  13. Setup for in situ x-ray diffraction study of swift heavy ion irradiated materials

    Science.gov (United States)

    Kulriya, P. K.; Singh, F.; Tripathi, A.; Ahuja, R.; Kothari, A.; Dutt, R. N.; Mishra, Y. K.; Kumar, Amit; Avasthi, D. K.

    2007-11-01

    An in situ x-ray diffraction (XRD) setup is designed and installed in the materials science beam line of the Pelletron accelerator at the Inter-University Accelerator Centre for in situ studies of phase change in swift heavy ion irradiated materials. A high vacuum chamber with suitable windows for incident and diffracted X-rays is integrated with the goniometer and the beamline. Indigenously made liquid nitrogen (LN2) temperature sample cooling unit is installed. The snapshots of growth of particles with fluence of 90MeV Ni ions were recorded using in situ XRD experiment, illustrating the potential of this in situ facility. A thin film of C60 was used to test the sample cooling unit. It shows that the phase of the C60 film transforms from a cubic lattice (at room temperature) to a fcc lattice at around T =255K.

  14. Development of in-pile instruments for fuel and material irradiation tests

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Akira; Kitagishi, Shigeru; Kimura, Nobuaki; Saito, Takashi; Nakamura, Jinichi; Ohmi, Masao; Izumo, Hironobu; Tsuchiya, Kunihiko [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    To get measurement data with high accuracy for fuel and material behavior studies in irradiation tests, two kinds of measuring equipments have been developed; these are the Electrochemical Corrosion Potential (ECP) sensor and the Linear Voltage Differential Transformer (LVDT) type gas pressure gauge. The ECP sensor has been developed to determine the corrosive potential under high temperature and high pressure water conditions. The structure of the joining parts was optimized to avoid stress concentration. The ECP sensor showed enough performance at 288degC and at 9MPa conditions. The LVDT type rod inner gas pressure gauge has been developed to measure gas pressure in a fuel element during neutron irradiation. To perform stable measurements with high accuracy under high temperature, high pressure and high dosed environment, the coil material of LVDT was changed to MI cable. As a result of this development, the LVDT type gas pressure gauge showed high accuracy within 1.8% of a full scale, and good stability. (author)

  15. Motion of Defect Clusters and Dislocations at a Crack Tip of Irradiated Material

    International Nuclear Information System (INIS)

    Moon, Won Jin; Kwon, Sang Chul; Kim, Whung Whoe

    2007-01-01

    Effects of defect clusters on mechanical properties of irradiated materials have not been clarified until now. Two radiation hardening models have been proposed. One is a dispersed barrier hardening mechanism based on the Orowan hardening model. This explains defect clusters as barriers to a dislocation motion. Generally the dislocation would rather shear or remove the defect clusters than make so-called Orowan loops. And the other is a cascade induced source hardening mechanism, which explains defect clusters as a Cottrell atmosphere for dislocation motions. However, the above mechanisms can not explain the microstructure of deformed material after irradiation and the phenomenon of yield softening. These mechanisms are based on an immobility of clusters. But we observed defect clusters could move into a specific crystallographic direction easily. Through 3 times of High Voltage Electron Microscope analysis, defect clusters have been observed to make one dimensional motion without applied external stress. If very small defect clusters could move under a stress gradient due to interactions between clusters, we can suggest that the clusters will move more actively when a stress gradient is applied externally. In-situ tensile test at TEM, we confirmed that kind of motion. We suggest defect clusters can move into crack tip, a stress-concentrated area due to tensile stress gradient and dislocations move out from the area by shear stress. Therefore radiation hardening can be explained agglomeration of defect clusters at stress concentrated area prohibits a generation of dislocation and make an increase of yield point

  16. Reduced cost design of liquid lithium target for international fusion material irradiation facility (IFMIF)

    International Nuclear Information System (INIS)

    Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki

    2001-01-01

    The International Fusion Materials Irradiation Facility (IFMIF) is being jointly planned to provide an accelerator-based D-Li neutron source to produce intense high energy neutrons (2 MW/m 2 ) up to 200 dpa and a sufficient irradiation volume (500 cm 3 ) for testing the candidate materials and components up to about a full lifetime of their anticipated use in ITER and DEMO. To realize such a condition, 40 MeV deuteron beam with a current of 250 mA is injected into high speed liquid lithium flow with a speed of 20 m/s. Following Conceptual Design Activity (1995-1998), a design study with focus on cost reduction without changing its original mission has been done in 1999. The following major changes to the CAD target design have been considered in the study and included in the new design: i) number of the Li target has been changed from 2 to 1, ii) spare of impurity traps of the Li loop was removed although the spare will be stored in a laboratory for quick exchange, iii) building volume was reduced via design changes in lithium loop length. This paper describes the reduced cost design of the lithium target system and recent status of Key Element Technology activities. (author)

  17. The role of strain localization in the fracture of irradiated pressure tube material

    International Nuclear Information System (INIS)

    Dutton, R.

    1989-04-01

    This report reviews those phenomena that lead to strain localization in zirconium alloys, with particular reference to the role played by the formation of shear bands in fracture processes. The important influence of plastic deformation, in general, on fracture mechanisms is emphasized. This is to be expected when elastic-plastic fracture mechanics is the chosen analytical technique. Intensely inhomogeneous characteristics of strain localization cause an abrupt bifurcation in the evolution of deformation strain and lead to plastic instability linked with intrinsic material behaviour (e.g., work softening) or of geometric origin (e.g., localized necking). Both of these effects are discussed in relation to measurable deformation parameters, such as the work hardening rate and strain rate sensitivity, which determine the degree of resistance to plastic instability. The modifying effect of irradiation on these quantities is given specific attention, the appropriate literature pertaining to Zircaloy and Zr-2.5% Nb being reviewed. Recommendations are made for a combined experimental and theoretical program to characterize strain localization and reduced ductility in irradiated cold-worked Zr-2.5% Nb pressure tube material. The relationship between the deformation properties and the fracture behaviour is discussed

  18. Effect of gamma irradiation on fluoride release and antibacterial activity of resin dental materials

    International Nuclear Information System (INIS)

    Carvalho, Fabiola Galbiatti de; Fucio, Suzana Beatriz Portugal de; Correr-Sobrinho, Lourenco; Pascon, Fernanda Miori; Kantovitz, Kamila Rosamilia; Puppin-Rontani, Regina Maria

    2009-01-01

    This study evaluated the effect of gamma irradiation on fluoride release and antibacterial activity of FluroShield (FS) and Clearfil Protect Bond (CPB). Four groups were formed: G1-FS + gamma; G2-FS without gamma; G3-CPB + gamma; G4-CPB without gamma. For fluoride release analysis, 12 disks of each material were prepared and covered with nail polish, except for one side (50.4 mm 2 area). G1 and G3 were sterilized with a 14.5 KGy dose at 27 deg C for 24 h, while G2 and G4 (controls) were not sterilized and were maintained under the same time and temperature conditions. Fluoride release measurements were made in duplicate (n=6) by an ion specific electrode. The antibacterial activity of the CPB and FS against Streptococcus mutans after gamma sterilization was evaluated by the agar-disc diffusion method. The diameter of the zones of microbial growth inhibition was recorded after 48 h. Data were analyzed statistically by ANOVA and Tukey's test (alpha=5%). Gamma sterilization decreased the fluoride release of FS by approximately 50%, while CPB was not affected. There was no statistically significant difference (p>0.05) in the antibacterial effect of CPB between gamma and non-gamma sterilization groups. FS presented no antibacterial activity. Gamma irradiation decreased the fluoride release of FS, but did not affect the antibacterial activity of the studied materials. (author)

  19. Effect of gamma irradiation on fluoride release and antibacterial activity of resin dental materials

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Fabiola Galbiatti de; Fucio, Suzana Beatriz Portugal de; Correr-Sobrinho, Lourenco [Universidade Estadual de Campinas (UNICAMP), Piracicaba, SP (Brazil). Piracicaba Dental School. Dept. of Dental Materials; Pascon, Fernanda Miori; Kantovitz, Kamila Rosamilia; Puppin-Rontani, Regina Maria [Universidade Estadual de Campinas (UNICAMP), Piracicaba, SP (Brazil). Piracicaba Dental School. Dept. of Pedriatric Dentistry], e-mail: rmpuppin@fop.unicamp.br

    2009-07-01

    This study evaluated the effect of gamma irradiation on fluoride release and antibacterial activity of FluroShield (FS) and Clearfil Protect Bond (CPB). Four groups were formed: G1-FS + gamma; G2-FS without gamma; G3-CPB + gamma; G4-CPB without gamma. For fluoride release analysis, 12 disks of each material were prepared and covered with nail polish, except for one side (50.4 mm{sup 2} area). G1 and G3 were sterilized with a 14.5 KGy dose at 27 deg C for 24 h, while G2 and G4 (controls) were not sterilized and were maintained under the same time and temperature conditions. Fluoride release measurements were made in duplicate (n=6) by an ion specific electrode. The antibacterial activity of the CPB and FS against Streptococcus mutans after gamma sterilization was evaluated by the agar-disc diffusion method. The diameter of the zones of microbial growth inhibition was recorded after 48 h. Data were analyzed statistically by ANOVA and Tukey's test (alpha=5%). Gamma sterilization decreased the fluoride release of FS by approximately 50%, while CPB was not affected. There was no statistically significant difference (p>0.05) in the antibacterial effect of CPB between gamma and non-gamma sterilization groups. FS presented no antibacterial activity. Gamma irradiation decreased the fluoride release of FS, but did not affect the antibacterial activity of the studied materials. (author)

  20. Effect of grain size on void swelling in irradiated materials: A phase-field approach

    International Nuclear Information System (INIS)

    Chang, Kunok; Lee, Gyeonggeun; Kwon, Junhyun

    2014-01-01

    The progress of swelling is retarded as the average grain diameter increases in a pure copper case. Within the framework of the production bias model (PBM), their experimental results were quantitatively explained. The phase-field method has already been used to investigate the void/bubble behavior in the irradiated materials. In particular, Millett et al. already incorporated the interaction between the point defect and the grain boundary in their study. Therefore, they described the void denuded zones and void peaked zones adjacent to the grain boundaries, which are already observed in the experimental investigations. We performed the phase-field simulation in order to verify the role of the grain diameter on the void swelling in the cascade damage condition. In addition, our results will be compared with the experimental observations or the theoretical works, such as PBM. Two-dimensional phase-field simulations were performed to investigate the void swelling process in the irradiated materials. We clearly observed the void denuded and void peaked zones, which were already observed in formal experimental and computational approaches. We also found that the progress of swelling was retarded as the average grain diameter increased. The triple junctions, which are believed to be a critical factor t affecting the fracture, are the main cites for the void nucleation and growth in our simulations

  1. Development of in-pile instruments for fuel and material irradiation tests

    International Nuclear Information System (INIS)

    Shibata, Akira; Kitagishi, Shigeru; Kimura, Nobuaki; Saito, Takashi; Nakamura, Jinichi; Ohmi, Masao; Izumo, Hironobu; Tsuchiya, Kunihiko

    2012-01-01

    To get measurement data with high accuracy for fuel and material behavior studies in irradiation tests, two kinds of measuring equipments have been developed; these are the Electrochemical Corrosion Potential (ECP) sensor and the Linear Voltage Differential Transformer (LVDT) type gas pressure gauge. The ECP sensor has been developed to determine the corrosive potential under high temperature and high pressure water conditions. The structure of the joining parts was optimized to avoid stress concentration. The ECP sensor showed enough performance at 288degC and at 9MPa conditions. The LVDT type rod inner gas pressure gauge has been developed to measure gas pressure in a fuel element during neutron irradiation. To perform stable measurements with high accuracy under high temperature, high pressure and high dosed environment, the coil material of LVDT was changed to MI cable. As a result of this development, the LVDT type gas pressure gauge showed high accuracy within 1.8% of a full scale, and good stability. (author)

  2. How can we predict microstructural changes caused by the multiscale irradiation process occurred in materials having complicated and hierarchical structures?

    International Nuclear Information System (INIS)

    Morishita, Kazunori; Watanabe, Yoshiyuki; Yoshimatsu, Jun-ichi

    2008-01-01

    Challenging efforts are discussed to establish an advanced methodology for prediction of material's property and performance changes by irradiation, which will be necessary by all means for the advanced reactor maintenance technology in the future. The changes of material's properties and performance caused by irradiation, such as irradiation-induced hardening, ductility loss, and material's degradation leading to reduction in reactor lifetime, are primarily determined by microstructural changes in materials during irradiation, where athermal lattice defects are continuously produced by collisions between an irradiating particle and a target material atom, and subsequently the defects are aggregated via diffusion in the form of dislocation loops, voids, and solute precipitation. These radiation damage processes are in essence multiscale phenomena, which involve varying time- and length-scales, from ballistic binary collisions to collective atomic motion in the thermal spike stage followed by the thermal activation process. In this report, the multiscale modeling approach is proposed to understand the processes in materials having complicated and hierarchical structures. (author)

  3. Radiation-thermal effects change of physico-mechanical properties in reactor materials irradiated with neutrons and energetic charged particles

    International Nuclear Information System (INIS)

    Hofman, A.

    1999-01-01

    In the first part of the report (chapter 1) the earlier results of the important scientific and technological investigations which were performed in the seventies years in Poland have been presented. They concerned the fabrication, corrosion, mechanical properties of materials for research and power reactors. Being of the general survey character, the chapter includes own, original results of research of thermal irradiation effects on microstructure evolution phase transformations and mechanical properties of reactor materials. The kinetics of isothermal transformation β→α in U-Cr 0.4% wt. alloy has been studied. Factors affecting stress-corrosion cracking of zirconium in iodine vapour have been investigated. The rings and loops for irradiation specimens and Hot Laboratory for postirradiation examination of construction materials is described. In the second part (chapters 2, 3, 4, 5) performed the investigations and simulations of radiation damage in metals by heavy ion beams (E > 1 MeV/a.m.n.) were described scientific base and technical problems of the method of irradiation of heavy ions and of the examination of irradiated samples is presented. It is followed by a summary of the results of simulation and reactor experiments on different materials. Radiation hardening of a number metals (Al, Zr, Cu, Ni, U) irradiated by heavy ion and neutrons, mechanical properties and microstructural evolution in ion and neutron irradiated austenitic stainless steel is described. The last chapter is a description of practical aspects of the presented studies in nuclear science and technology. (author)

  4. Ion irradiation studies of construction materials for high-power accelerators

    Science.gov (United States)

    Mustafin, E.; Seidl, T.; Plotnikov, A.; Strašík, I.; Pavlović, M.; Miglierini, M.; Stanćek, S.; Fertman, A.; Lanćok, A.

    The paper reviews the activities and reports the current results of GSI-INTAS projects that are dealing with investigations of construction materials for high-power accelerators and their components. Three types of materials have been investigated, namely metals (stainless steel and copper), metallic glasses (Nanoperm, Finemet and Vitrovac) and organic materials (polyimide insulators and glass fiber reinforced plastics/GFRP). The materials were irradiated by different ion beams with various fluencies and energies. The influence of radiation on selected physical properties of these materials has been investigated with the aid of gamma-ray spectroscopy, transmission Mössbauer spectroscopy (TMS), conversion electrons Mössbauer spectroscopy (CEMS), optical spectroscopy (IR and UV/VIS) and other analytical methods. Some experiments were accompanied with computer simulations by FLUKA, SHIELD and SRIM codes. Validity of the codes was verified by comparison of the simulation results with experiments. After the validation, the codes were used to complete the data that could not be obtained experimentally.

  5. Nano lead oxide and epdm composite for development of polymer based radiation shielding material: Gamma irradiation and attenuation tests

    Science.gov (United States)

    Özdemir, T.; Güngör, A.; Akbay, I. K.; Uzun, H.; Babucçuoglu, Y.

    2018-03-01

    It is important to have a shielding material that is not easily breaking in order to have a robust product that guarantee the radiation protection of the patients and radiation workers especially during the medical exposure. In this study, nano sized lead oxide (PbO) particles were used, for the first time, to obtain an elastomeric composite material in which lead oxide nanoparticles, after the surface modification with silane binding agent, was used as functional material for radiation shielding. In addition, the composite material including 1%, 5%, 10%, 15% and 20% weight percent nano sized lead oxide was irradiated with doses of 81, 100 and 120 kGy up to an irradiation period of 248 days in a gamma ray source with an initial dose rate of 21.1 Gy/h. Mechanical, thermal properties of the irradiated materials were investigated using DSC, DMA, TGA and tensile testing and modifications in thermal and mechanical properties of the nano lead oxide containing composite material via gamma irradiation were reported. Moreover, effect of bismuth-III oxide addition on radiation attenuation of the composite material was investigated. Nano lead oxide and bismuth-III oxide particles were mixed with different weight ratios. Attenuation tests have been conducted to determine lead equivalent values for the developed composite material. Lead equivalent thickness values from 0.07 to 0.65 (2-6 mm sample thickness) were obtained.

  6. Results from the CDE phase activity on neutron dosimetry for the international fusion materials irradiation facility test cell

    CERN Document Server

    Esposito, B; Maruccia, G; Petrizzi, L; Bignon, G; Blandin, C; Chauffriat, S; Lebrun, A; Recroix, H; Trapp, J P; Kaschuck, Y

    2000-01-01

    The international fusion materials irradiation facility (IFMIF) project deals with the study of an accelerator-based, deuterium-lithium source, producing high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials for fusion energy reactors. IFMIF would also provide calibration and validation of data from fission reactor and other accelerator based irradiation tests. This paper describes the activity on neutron/gamma dosimetry (necessary for the characterization of the specimens' irradiation) performed in the frame of the IFMIF conceptual design evaluation (CDE) neutronics tasks. During the previous phase (conceptual design activity (CDA)) the multifoil activation method was proposed for the measurement of the neutron fluence and spectrum and a set of suitable foils was defined. The cross section variances and covariances of this set of foils have now been used for tests on the sensitivity of the IFMIF neutron spectrum determination to cross section uncertainties...

  7. Creation of a plant for gamma-sterilization of the medical equipment and irradiation of industrial materials

    International Nuclear Information System (INIS)

    Maltceva, F.; Petukhov, V.; Chekushin, A.

    1996-01-01

    The purpose of the project. To create a powerful irradiation devices allowing to fulfill the irradiation of large-sized objects with dozes up to 4 Rad, and caring out the irradiation technologies by using of it. The basic tasks of the project and sequence of fulfillment. The realization of irradiating technologies is possible at WWR-K research reactor that is located near of Almaty . At the reactor complex, there are the means for creation powerful irradiating devices, and quite qualified personnel having the experience of works with radioactive materials is available. All of it allows, both for the staff and for the population living nearby, to have the radiation safety during of caring out of the work with the use of high intensive radioactive sources. Technical problems. The creation of plant has the following stages: - fulfilling of the calculation and design work; - caring out of mechanical works on manufacturing the irradiating devices, cartridges for irradiating sources and a hermetic tank for irradiated objects; - installation of devices, fulfilling of the tests; - purchasing or manufacturing of radioactive sources (-10-60); - loading of the sources into devices, and realization of measurements of the fields of gamma field inside the container; - making up the working instruction; - training of the personnel

  8. Development of Manufacturing Method of Highly Functional Material Gallic acid-CLA Ester Using Irradiation

    International Nuclear Information System (INIS)

    Cho, C. H.; Byun, M. W.; Jeong, I. Y.; Kim, D. H.

    2006-01-01

    Increasing interest and current trends for natural materials with various health beneficial functions by radiation (RT)-biotechnology (BT) fusion by developed countries. However, the information and development of new functional materials using the RT-BT fusion technology is still limited. The target material developed and manufactured by RT-BT fusion technology may have a multi-functional effect on human health and it can be applied for pharmaceutical materials as well as functional food ingredient. The market of functional new materials has been grown dramatically and a multi-functional material manufactured by RT-BT fusion technology may have a great economic impact for both the domestic and overseas market. Development of GA-CLA ester by chemical synthetic method. Transformation of linoleic acid to conjugated linoleic acid by irradiation. Identification and confirmation of the biological functions including antioxidative, cancer cell proliferation inhibition, anti-microbial, enhancement of immune response and lipid metabolism of GA-CLA ester. Increase industrial applicability of the new materials. Development of GA-CLA ester by chemical synthetic method(2 patents submitted). Development of the optimum methodology of GA-CLA and its derivative, octadeca-9,12-dienyl-3,4,5-trihydroxy benzoate). Identification and confirmation of biological activities of GA-CLA. Extramural funding from the Ministry of Commerce, Industry, and Energy subjected by gallic acid-fatty acid derivatives (205,000,000 Won). Provides the basic data for successful project 'Development of cosmeceutical and cosmetics using gallic acid fatty acid derivatives' funded by Ministry of Commerce, Industry, and Energy and collaboration with the Technology-invested venture company, SunBiotech, Co. and problem-solving for industrial application. Complete the patent procedure and publish the results to international or domestic peer-reviewed journals

  9. Studies on Development of Polymeric Materials Using Gamma Irradiation for Contact and Intraocular Lenses

    Directory of Open Access Journals (Sweden)

    Pranshu Chhabra

    2009-01-01

    Full Text Available For the development of materials for contact lenses and intraocular lenses, the selection criteria is based on the (i capacity to absorb and retain water, (ii hydrophilicity and hydrophobicity, (iii refractive index and (iv hardness besides the other essential properties. Various monomers are being studied to develop suitable materials for such applications. Selection of suitable monomers that can be converted into optical materials of desired characteristics is the most essential step. In the present paper, an attempt has been made to develop suitable optical polymers based on 2-hydroxy ethyl methacrylate (HEMA, N-vinyl pyrrolidone (NVP, methyl methacrylate (MMA, methacrylic acid (MAA, and styrene. Compositions were prepared in such a way that polymers of varying hydrophilicity or hydrophobicity could be obtained keeping HEMA as the base (main monomer. For polymerization, gamma irradiation (Co-60 as a source was used. The results of the study showed that: (i an increase in NVP and MAA content brought in an increase in hydrophilicity of polymerized HEMA (pHEMA, while the addition of styrene and MMA decreased hydrophilicity of polymerized HEMA (pHEMA, (ii polymers for contact lenses with water retention capacity as high as >50 wt.% and as low as <10 wt% with varying content of suitable comonomers can be designed, (iii polymeric materials for contact lenses can be made by using radiation processing such as Co-60 and (iv a dose of 40 kGy was found to be ideal for purpose.

  10. Method of polymerizing ethylenically unsaturated materials by irradiation and composition for use therein

    International Nuclear Information System (INIS)

    Nemcek, J.; Heap, N.

    1976-01-01

    This patent concerns photopolymerizable compositions consisting essentially of at least one polymerizable ethylenically unsaturated material and a photosensitive catalyst comprising (a) from 0.5 to 5 percent based on the ethylenically unsaturated material of at least one photosensitizer having the structure Ph(CO)C 2 A(CO)Ph, where Ph is phenyl, halogen-substituted phenyl, phenylene or halogen-substituted phenylene and A is a cyclic hydrocarbyl group, a halogen-substituted cyclic hydrocarbyl group, or a group of the formula X(NR)COCONRY, where X and Y each is hydrogen, a hydrocarbyl, or a halogen-substituted hydrocarbyl group, and (b) from 1 to 5 percent by weight based on the ethylenically unsaturated material of a reducing agent capable of reducing the photosensitizer when the photosensitizer is in an excited state. Also described is a process of preparing polymeric materials by irradiating the foregoing polymerizable composition at a wavelength capable of exciting the photosensitizer to an excited state

  11. Irradiation effects on weld heat-affected zone and plate materials (series 11)

    International Nuclear Information System (INIS)

    Nanstad, R.K.; McCabe, D.E.

    1995-01-01

    The purpose of this task is to examine the effects of neutron irradiation on the fracture toughness (ductile and brittle) of the HAZ of welds and of A 302 grade B (A302B) plate materials typical of those used fabricating older RPVs. The initial plate material of emphasis will be A302B steel, not the A302B modified with nickel additions. This decision was made by the NRC following a survey of the materials of construction for RPBs in operating U.S. nuclear plants. Reference 1 was used for the preliminary survey, and the information from that report was revised by NRC staff based on information contained in the licensee responses to Generic Letter (GL) 92-01, open-quotes Reactor Vessel Structural Integrity, 10CFR50.54(f).close quotes The resulting survey showed a total of eight RPVs with A302B, ten with A302B (modified), and one with A302 grade A plate. Table 5.1 in the previous semiannual report provides a summary of that survey. For the HAZ portion of the program, the intent is to examine HAZ material in the A302B (i.e., with low nickel content) and in A302B (modified) or A533B-1 (i.e., with medium nickel content). During this reporting period, two specific plates were identified as being applicable to this task. One plate is A302B and the other is A302B (modified). The A302B plate (43 x 42 x 7 in.) will be prepared for welding, while the A302B (modified) plate already contains a commercially produced weld (heat 33A277, Linde 0091 flux). These plates were identified from a list of ten materials provided by Mr. E. Biemiller of Yankee Atomic Electric Company (YAEC). The materials have been requested from YAEC for use in this irradiation task, and arrangements are being made with YAEC for procurement of the plates mentioned above

  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. Microbiological decontamination of botanical raw materials and corresponding pharmaceutical products by irradiation

    International Nuclear Information System (INIS)

    Katusin-Razem, B.; Novak, B.; Razem, D.

    2001-01-01

    Microbiological contamination typical of botanical raw materials used in the manufacture of pharmaceuticals decreases with the increasing level of processing, on going from flowers and leaves (10 4 -10 8 CFU/g), to fruits and seeds (10 2 -10 6 CFU/g), to liquid extracts (10 4 -10 6 CFU/g), and to dry extracts (10 2 -10 5 CFU/g). At the same time the resistivity of microflora to irradiation, expressed as a dose required for the first 90% reduction, increases along the same assortment as 2, 4, 5 and 5 kGy, respectively. This results in doses between 4 and 30 kGy required to treat typical contamination, or between 10 and 40 kGy for severe cases. The contamination of final products, phyto-therapeutic ointments (10 4 -10 7 CFU/g), is relatively sensitive to irradiation (D first90%red =1 kGy) and usually does not require doses higher than 8 kGy

  14. 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

  15. Perspectives for online analysis of raw material by pulsed neutron irradiation

    Science.gov (United States)

    Bach, Pierre; Le Tourneur, P.; Poumarede, B.

    1997-02-01

    On-line analysis by pulsed neutron irradiation is an example of an advanced technology application of nuclear techniques, concerning real problems in the cement, mineral and coal industries. The most significant of these nuclear techniques is their capability of continuous measurement without contact and without sampling, which can lead to improved control of processes and resultant large financial savings. Compared to Californium neutron sources, the use of electrical pulsed neutron generators allows to obtain a higher signal/noise ratio for a more sensitive measurement, and allows to overcome a number of safety problems concerning transportation, installation and maintenance. An experiment related to a possible new on-line raw material analyzer is described, using a pulsed neutron generator. The key factors contributing to an accurate measurement are related to a suitable generator, to a high count rate gamma ray spectroscopy electronics, and to computational tools. Calculation and results for the optimization of the neutron irradiation time diagram are reported. One of the operational characteristics of such an equipment is related to neutron flux available: it is possible to adjust it to the requested accuracy, i.e. for a high accuracy during a few hours/day and for a lower accuracy the rest of the time. This feature allows to operate the neutron tube during a longer time, and then to reduce the cost of analysis.

  16. H2 formation by electron irradiation of SBA-15 materials and the effect of Cu(II) grafting

    International Nuclear Information System (INIS)

    Brodie-Linder, N.; Le Caer, S.; Shahdo Alam, M.; Renault, J.P.; Alba-Simionesco, Ch.

    2010-01-01

    Measurement of H 2 production from electron irradiation (10 MeV) on SBA-15 materials has shown that adsorbed water is attacked preferentially. Silanol groups are only attacked when they are in the majority with respect to adsorbed water, however they are much less efficient at producing H 2 . The comparison between water content before and after electron irradiation and the corresponding H 2 production indicates that water desorption is the main route to adsorbed water loss for SBA-15 materials. On the other hand, surface silanol groups are more susceptible to attack,leading to H 2 production when SBA-15 samples have undergone extensive thermal treatment. Electron irradiation of SBA-15-Cu materials has shown that the presence of Cu(II) on the surface reduces and inhibits the production of H 2 . This inhibiting power affects adsorbed water bonded to grafted copper but not surface silanol groups. (authors)

  17. Assessment of the gas dynamic trap mirror facility as intense neutron source for fusion material test irradiations

    International Nuclear Information System (INIS)

    Fischer, U.; Moeslang, A.; Ivanov, A.A.

    2000-01-01

    The gas dynamic trap (GDT) mirror machine has been proposed by the Budker Institute of nuclear physics, Novosibirsk, as a volumetric neutron source for fusion material test irradiations. On the basis of the GDT plasma confinement concept, 14 MeV neutrons are generated at high production rates in the two end sections of the axially symmetrical central mirror cell, serving as suitable irradiation test regions. In this paper, we present an assessment of the GDT as intense neutron source for fusion material test irradiations. This includes comparisons to irradiation conditions in fusion reactor systems (ITER, Demo) and the International Fusion Material Irradiation Facility (IFMIF), as well as a conceptual design for a helium-cooled tubular test assembly elaborated for the largest of the two test zones taking proper account of neutronics, thermal-hydraulic and mechanical aspects. This tubular test assembly incorporates ten rigs of about 200 cm length used for inserting instrumented test capsules with miniaturized specimens taking advantage of the 'small specimen test technology'. The proposed design allows individual temperatures in each of the rigs, and active heating systems inside the capsules ensures specimen temperature stability even during beam-off periods. The major concern is about the maximum achievable dpa accumulation of less than 15 dpa per full power year on the basis of the present design parameters of the GDT neutron source. A design upgrading is proposed to allow for higher neutron wall loadings in the material test regions

  18. Irradiation study of different silicon materials for the CMS tracker upgrade

    International Nuclear Information System (INIS)

    Erfle, Joachim

    2014-05-01

    Around 2022, an upgrade of the LHC collider complex is planned to significantly increase the luminosity (the High Luminosity LHC, HL-LHC). This means that the experiments have to cope with a higher number of collisions per bunch crossing and survive in a radiation environment much harsher than that at the present LHC. Especially the tracking detectors have to be improved for the HL-LHC. The increased number of tracks requires an increase of the number of readout channels while the higher radiation makes new sensor materials necessary. Within CMS, a measurement campaign was initiated to study the performance of different silicon materials in a corresponding radiation environment. To simulate the expected radiation the samples were irradiated with neutrons and with protons with two different energies. Radiation damage can be divided in two categories. First, ionizing energy loss in the surface isolation layers of the sensor leads to a change of the concentration of charged states in the sensor surface and therefore alters the distribution of the electrical fields in the sensor. Second, non-ionizing energy loss in the bulk of the sensor material leads to a variety of defects in the silicon lattice. Electrically active defects can influence the material properties. The three properties under investigation are the reverse current, the full depletion voltage and the charge collection. While the reverse current and full depletion voltage influence the power dissipation and the noise of the detector, the charge collection directly influences the measurement. The material properties were studied using pad and strip sensor. The structures were electrically characterized before and after irradiation with different fluences of neutrons and protons, corresponding to the expected fluences at different radii of the outer tracker after 3000 fb -1 . The charge collection measurements were mainly performed using the ALiBaVa readout system and the charge was induced with

  19. Material correlations and models for the irradiation behavior of fissile and fertile material in SNR-300, Mark-II and KNK II, third core

    International Nuclear Information System (INIS)

    Fenneker; Steinmetz; Toebbe

    1986-07-01

    The report contains the material correlations and models used in the fuel pin design code IAMBUS for the irradiation behavior of PuO 2 -UO 2 fissile materials and UO 2 fertile materials of the SNR-300 Mark-II reload and the KNK II third core. They are applicable for pellet densities of more than 90 % of the theoretical density. The presented models of the fuel behavior and the applied material correlations have been derived either from single experiments or from the comparison of theoretically predicted integral fuel behavior with the results of fuel pin irradiation experiments. The material correlations have been examined and extended in the frame of the collaborations INTERATOM/KWU and INTERATOM/KfK. French and British results were included, when available from the European fast reactor knowledge exchange [de

  20. A correlation between micro- and nano-indentation on materials irradiated by high-energy heavy ions

    Science.gov (United States)

    Yang, Yitao; Zhang, Chonghong; Ding, Zhaonan; Su, Changhao; Yan, Tingxing; Song, Yin; Cheng, Yuguang

    2018-01-01

    Hardness testing is an efficient means of assessing the mechanical properties of materials due to the small sampling volume requirement. Previous studies have established the correlation between flow stress and Vickers hardness. However, the damage layer produced by ions irradiation with low energy is too thin to perform Vickers hardness test, which is usually measured by nano-indentation. Therefore, it is necessary to correlate the Vickers hardness and nano-hardness for the convenience of assessing mechanical properties of materials under irradiation. In this study, various materials (pure nickel, nickel base alloys and oxide dispersion strengthened steel) were irradiated with high-energy heavy ions to different damage levels. After irradiation, micro- and nano-indentation were performed to characterize the change in hardness. Due to indentation size effect (ISE), the hardness was dependent of load or depth. Therefore, Nix-Gao model was used to obtain the hardness without ISE (Hv0 and Hnano_0). The determined Hv0 was plotted as a function of the corresponding Hnano_0, then a good linear relation was found between Vickers hardness and nano-hardness, and a coefficient was determined to be 81.0 ± 10.5, namely, Hv 0 = 81.0Hnano _ 0 (Hv0 with unit of kgf/mm2, Hnano_0 with unit of GPa). This correlation was based on the data from various materials, therefore it was independent of materials. Based on the established correlation and nano-indentation results, the change fraction in yield stress of Inconel 718 and pure Ni with ion irradiation was compared with that with neutron irradiation. The data of Inconel 718 with heavy ion irradiation was in good agreement with the data with neutron irradiation, which was a good demonstration for the validation of the established correlation. However, a distinctive difference in change fraction of yield stress was seen for pure Ni under heavy ion irradiation and neutron irradiation, which was attributed to the difference in samples

  1. Use of the National Low-Temperature Neutron Irradiation Facility (NLTNIF) for fusion materials research

    International Nuclear Information System (INIS)

    Coltman, R.R. Jr.; Kerchner, H.R.; Klabunde, C.E.

    1986-01-01

    In May 1983 the Division of Materials Sciences, Office of Basic Energy Sciences of the Department of Energy authorized the establishment of a National Low-Temperature Neutron Irradiation Facility (NLTNIF) at ORNL's Bulk Shielding Reactor (BSR). The NLTNIF, which will be available for qualified experiments at no cost to users, will provide a combination of high radiation intensities and special environmental and testing conditions that have not been previously available in the US. Since the DOE authorization, work has proceeded on the design and construction of the new facility without interruption. This report describes the present status of the development of NLTNIF and, for the information of new candidate users, a recounting of the major specifications and capabilities is also given

  2. Modification induced by laser irradiation on physical features of plastics materials filled with nanoparticles

    Directory of Open Access Journals (Sweden)

    Scolaro Cristina

    2018-01-01

    Full Text Available The Thermal Laser Welding (TLW process involves localized heating at the interface of two pieces of plastic that will be joined. Polymeric materials of Ultra High Molecular Weight Polyethylene (UHMWPE, both pure and containing nanostructures at different concentrations (titanium and silver nanoparticles, were prepared as thin foils in order to produce an interface between a substrate transparent to the infrared laser wavelength and an highly absorbent substrate, in order to be welded by the laser irradiation. The used diode laser operates at 970 nm wavelength, in continuum, with a maximum energy of 100 mJ, for times of the order of 1 -60 s, with a spot of 300 μm of diameter. The properties of the polymers and of nanocomposite sheets, before and after the laser welding process, were measured in terms of optical characteristics, wetting ability, surface roughness and surface morphology.

  3. Fusion Materials Irradiation Test (FMIT) facility lithium system: a design and development status

    International Nuclear Information System (INIS)

    Brackenbury, P.J.; Bazinet, G.D.; Miller, W.C.

    1983-01-01

    The design and development of the Fusion Materials Irradiation Test (FMIT) Facility lithium system is outlined. This unique liquid lithium recirculating system, the largest of its kind in the world, is described with emphasis on the liquid lithium target assembly and other important components necessary to provide lithium flow to the target. The operational status and role of the Experimental Lithium System (ELS) in the design of the FMIT lithium system are discussed. Safety aspects of operating the FMIT lithium system in a highly radioactive condition are described. Potential spillage of the lithium is controlled by cell liners, by argon flood systems and by remote maintenance features. Lithium chemistry is monitored and controlled by a side-stream loop, where impurities measured by instruments are collected by hot and cold traps

  4. Fusion Materials Irradiation Test (FMIT) facility lithium system: a design and development status

    Energy Technology Data Exchange (ETDEWEB)

    Brackenbury, P.J.; Bazinet, G.D.; Miller, W.C.

    1983-01-01

    The design and development of the Fusion Materials Irradiation Test (FMIT) Facility lithium system is outlined. This unique liquid lithium recirculating system, the largest of its kind in the world, is described with emphasis on the liquid lithium target assembly and other important components necessary to provide lithium flow to the target. The operational status and role of the Experimental Lithium System (ELS) in the design of the FMIT lithium system are discussed. Safety aspects of operating the FMIT lithium system in a highly radioactive condition are described. Potential spillage of the lithium is controlled by cell liners, by argon flood systems and by remote maintenance features. Lithium chemistry is monitored and controlled by a side-stream loop, where impurities measured by instruments are collected by hot and cold traps.

  5. Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials

    Energy Technology Data Exchange (ETDEWEB)

    Nanstad, Randy K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Odette, George R. [Univ. of California, Santa Barbara, CA (United States); Almirall, Nathan [Univ. of California, Santa Barbara, CA (United States); Robertson, Janet [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Server, W. L. [ATI Consulting, Pinehurst, NC (United States); Yamamoto, T. [Univ. of California, Santa Barbara, CA (United States); Wells, Peter [Univ. of California, Santa Barbara, CA (United States)

    2017-05-01

    The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughness loss dependent on the radiation sensitivity of the materials. The available embrittlement predictive models and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.

  6. The RaDIATE High-Energy Proton Materials Irradiation Experiment at the Brookhaven Linac Isotope Producer Facility

    Energy Technology Data Exchange (ETDEWEB)

    Ammigan, Kavin; et al.

    2017-05-01

    The RaDIATE collaboration (Radiation Damage In Accelerator Target Environments) was founded in 2012 to bring together the high-energy accelerator target and nuclear materials communities to address the challenging issue of radiation damage effects in beam-intercepting materials. Success of current and future high intensity accelerator target facilities requires a fundamental understanding of these effects including measurement of materials property data. Toward this goal, the RaDIATE collaboration organized and carried out a materials irradiation run at the Brookhaven Linac Isotope Producer facility (BLIP). The experiment utilized a 181 MeV proton beam to irradiate several capsules, each containing many candidate material samples for various accelerator components. Materials included various grades/alloys of beryllium, graphite, silicon, iridium, titanium, TZM, CuCrZr, and aluminum. Attainable peak damage from an 8-week irradiation run ranges from 0.03 DPA (Be) to 7 DPA (Ir). Helium production is expected to range from 5 appm/DPA (Ir) to 3,000 appm/DPA (Be). The motivation, experimental parameters, as well as the post-irradiation examination plans of this experiment are described.

  7. Sterilisation of allograft cortical bone using gamma irradiation: effect on strength and material ultrastructure

    International Nuclear Information System (INIS)

    Price, R.; Walters, M.

    1996-01-01

    Full text: The use of allograft bone in revision joint and limb salvage surgery is widespread and increasing (Buck B.E. et al, Clin Orthop 303: 8-17, 1994). To reduce the risk of disease transmission from donor graft contamination (particularly HIV and hepatitis) sterilisation is practiced worldwide. Gamma (γ)-irradiation using a dose of 1.5 - 2.5 Mrads is common. However, γ-irradiation is known to reduce bone strength, though the extent and mechanisms are controversial (eg Bright RW et al, Trans Orthop Res Soc 3: 210, 1978). We measured the effect of γ-irradiation on bone strength and properties reflecting bone material ultrastructure. Diaphyseal bone was obtained from the femur of a 47 year-old male would-be donor with suspicious hepatitis serology. Beams of cortical bone (long axes parallel to the femur) were cut using a low speed diamond saw bathed in Ringer's solution. Four groups were irradiated with γ-rays (0, 1.5, 2.5 and 5.0±0.5[SD] Mrads). Blinded investigations were performed: Ultimate stress (Ult Stress, N= 16 replicates in each dose group). Each beam was loaded at its midpoint at a rate of 25 mm/min until failure, while its ends were supported 40 mms apart. Ult stress was calculated from 3-point bending theory using the load vs displacement curve and the cross-sectional area of the break (Power RA et al, submitted to J Bone and Joint Surg). Differential scanning calorimetry (DSC) was performed over the range -15 to +5 deg C. Samples were demineralized and small (7-10 mg) blocks were cut and sealed in stainless steel calorimetry capsules. The enthalpy (reflecting the normalised free water content) was calculated from the sample mass plus area under the heat capacity curve. Pyridinoline collagen (acid-insoluble) crosslinks (Pyrid, N=10) (Randall D et al, JBone and Min Res, 1996, in press) were determined from 5-mm 3 demineralised, freeze dried samples. Small and medium angle X-ray diffraction (XRD, N=5). Demineralised bone was sliced into thin

  8. Transportation impact analysis for shipment of irradiated N-reactor fuel and associated materials

    International Nuclear Information System (INIS)

    Daling, P.M.; Harris, M.S.

    1994-12-01

    An analysis of the radiological and nonradiological impacts of highway transportation of N-Reactor irradiated fuel (N-fuel) and associated materials is described in this report. N-fuel is proposed to be transported from its present locations in the 105-KE and 105-KW Basins, and possibly the PUREX Facility, to the 327 Building for characterization and testing. Each of these facilities is located on the Hanford Site, which is near Richland, Washington. The projected annual shipping quantity is 500 kgU/yr for 5 years for a total of 2500 kgU. It was assumed the irradiated fuel would be returned to the K- Basins following characterization, so the total amount of fuel shipped was assumed to be 5000 kgU. The shipping campaign may also include the transport and characterization of liquids, gases, and sludges from the storage basins, including fuel assembly and/or canister parts that may also be present in the basins. The impacts of transporting these other materials are bounded by the impacts of transporting 5000 kgU of N-fuel. This report was prepared to support an environmental assessment of the N-fuel characterization program. The RADTRAN 4 and GENII computer codes were used to evaluate the radiological impacts of the proposed shipping campaign. RADTRAN 4 was used to calculate the routine exposures and accident risks to workers and the general public from the N-fuel shipments. The GENII computer code was used to calculate the consequences of the maximum credible accident. The results indicate that the transportation of N-fuel in support of the characterization program should not cause excess radiological-induced latent cancer fatalities or traffic-related nonradiological accident fatalities. The consequences of the maximum credible accident are projected to be small and result in no excess latent cancer fatalities

  9. Comparison of Material Behavior of Matrix Graphite for HTGR Fuel Elements upon Irradiation: A literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel elements for the HTGRs (i.e., spherical fuel element in pebble-bed type core design and fuel compact in prismatic core design) consists of coated fuel particles dispersed and bonded in a closely packed array within a carbonaceous matrix. This matrix is generally made by mixing fully graphitized natural and needle- or pitchcoke originated powders admixed with a binder material (pitch or phenolic resin), The resulting resinated graphite powder mixture, when compacted, may influence a number of material properties as well as its behavior under neutron irradiation during reactor operation. In the fabrication routes of these two different fuel element forms, different consolidation methods are employed; a quasi-isostatic pressing method is generally adopted to make pebbles while fuel compacts are fabricated by uni-axial pressing mode. The result showed that the hardness values obtained from the two directions showed an anisotropic behavior: The values obtained from the perpendicular section showed much higher micro hardness (176.6±10.5MPa in average) than from the parallel section ((125.6±MPa in average). This anisotropic behavior was concluded to be related to the microstructure of the matrix graphite. This may imply that the uni-axial pressing method to make compacts influence the microstructure of the matrix and hence the material properties of the matrix graphite.

  10. Progress report on the accelerator production of tritium materials irradiation program

    International Nuclear Information System (INIS)

    Maloy, S.A.; Sommer, W.F.; Brown, R.D.; Roberts, J.E.

    1997-01-01

    The Accelerator Production of Tritium (APT) project is developing an accelerator and a spoliation neutron source capable of producing tritium through neutron capture on He-3. A high atomic weight target is used to produce neutrons that are then multiplied and moderated in a blanket prior to capture. Materials used in the target and blanket region of an APT facility will be subjected to several different and mixed particle radiation environments; high energy protons (1-2 GeV), protons in the 20 MeV range, high energy neutrons, and low energy neutrons, depending on position in the target and blanket. Flux levels exceed 10 14 /cm 2 s in some areas. The APT project is sponsoring an irradiation damage effects program that will generate the first data-base for materials exposed to high energy particles typical of spallation neutron sources. The program includes a number of candidate materials in small specimen and model component form and uses the Los Alamos Spallation Radiation Effects Facility (LASREF) at the 800 MeV, Los Alamos Neutron Science Center (LANSCE) accelerator

  11. Neutron-irradiation facilities at the Intense Pulsed Neutron Source-I for fusion magnet materials studies

    International Nuclear Information System (INIS)

    Brown, B.S.; Blewitt, T.H.

    1982-01-01

    The decommissioning of reactor-based neutron sources in the USA has led to the development of a new generation of neutron sources that employ high-energy accelerators. Among the accelerator-based neutron sources presently in operation, the highest-flux source is the Intense Pulsed Neutron Source (IPNS), a user facility at Argonne National Laboratory. Neutrons in this source are produced by the interaction of 400 to 500 MeV protons with either of two 238 U target systems. In the Radiation Effects Facility (REF), the 238 U target is surrounded by Pb for neutron generatjion and reflection. The REF has three separate irradiation thimbles. Two thimbles provide irradiation temperatures between that of liquid He and several hundred degrees centigrade. The third thimble operates at ambient temperature. The large irradiation volume, the neutron spectrum and flux, the ability to transfer samples without warm up, and the dedication of the facilities during the irradiation make this ideally suited for radiation damage studies on components for superconducting fusion magnets. Possible experiments for fusion magnet materials are discussed on cyclic irradiation and annealing of stabilizers in a high magnetic field, mechanical tests on organic insulation irradiated at 4 K, and superconductors measured in high fields after irradiation

  12. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

  13. Development of sterilization and storage processes of Xeno-skins by gamma irradiation for the Industrialization as dressing materials

    International Nuclear Information System (INIS)

    Choi, Jong Il; Lee, Ju Woon; Kim, Jae Hun; Song, Beom Seok; Kim, Jae Kyung; Park, Jong Hum; Sung, Nak Yun; Jo, Eu Ri; Kim, Jeong Soo

    2011-08-01

    Allografts have been increased in concerns as treatment methods to be use in replacement and reconstruction procedures of tissue. The major concern about tissue and bone transplantation is the risk of transferring bacteria, viruses from recipient to the donors, especially from different species. To avoid these risks, tissue and bone banks use many different sterilization methods such as a deep-frozenness, freeze-drying ethylene oxide-sterilization and gamma-irradiation. Among these methods, gamma-irradiation has been widely used to sterilize the tissue for allograft and xenograft safety. Gamma irradiation could be an effective method for sterilization of soft tissue for the development of xenografts. Also, this study suggests the potential of development as tissue graft with the improved mechanical property of the structural stability and protection of tissue materials by antioxidant on porcine skin irradiated for sterilization and storage process

  14. Effect of low dose pre-irradiation on DNA damage and genetic material damage caused by high dosage of cyclophosphamide

    International Nuclear Information System (INIS)

    Yu Hongsheng; Zhu Jingjuan; Shang Qingjun; Wang Zhuomin; Cui Fuxian

    2007-01-01

    Objective: To study the effect of low dose γ-rays pre-irradiation on the induction of DNA damage and genetic material damage in peripheral lymphocytes by high dosage of cyclophosphamide (CTX). Methods: Male Kunming strain mice were randomly divided into five groups: control group, sham-irradiated group, low dose irradiated group(LDR group), cyclophosphamide chemotherapy group(CTX group) and low dose irradiation combined with chemotherapy group(LDR + CTX group). After being feeded for one week, all the mice were implanted subcutaneously with S180 cells in the left groin (control group excluded). On days 8 and 11, groups of LDR and LDR + CTX were administered with 75 mGy of whole-body irradiation, 30 h later groups CTX and LDR + CTX were injected intraperitoneally 3.0 mg cyclophosphamide. All the mice were sacrificed on day 13. DNA damage of the peripheral lymphocytes was analyzed using single cell gel electrophoresis (SCGE). Genetic material damage was analyzed using micronucleus frequency(MNF) of polychromatoerythrocytes(PCE) in bone marrow. Results: (1) Compared with control group and sham-irradiated group, the DNA damage of peripheral lymphocytes in CTX group were increased significantly (P 0.05). Conclusions: (1) High- dosage of CTX chemotherapy can cause DNA damage in peripheral lymphocytes. 75 mGy y-irradiation before chemotherapy may have certain protective effect on DNA damage. (2) CTX has potent mutagenic effect, giving remarkable rise to MNF of PCE. 75 mGy γ-ray pre-irradiation has not obvious protection against genetic toxicity of high-dose CTX chemotherapy. (authors)

  15. 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.)

  16. Multi-MW accelerator target material properties under proton irradiation at Brookhaven National Laboratory linear isotope producer

    Science.gov (United States)

    Simos, N.; Ludewig, H.; Kirk, H.; Dooryhee, E.; Ghose, S.; Zhong, Z.; Zhong, H.; Makimura, S.; Yoshimura, K.; Bennett, J. R. J.; Kotsinas, G.; Kotsina, Z.; McDonald, K. T.

    2018-05-01

    The effects of proton beams irradiating materials considered for targets in high-power accelerator experiments have been studied using the Brookhaven National Laboratory's (BNL) 200 MeV proton linac. A wide array of materials and alloys covering a wide range of the atomic number (Z) are being scoped by the high-power accelerator community prompting the BNL studies to focus on materials representing each distinct range, i.e. low-Z, mid-Z and high-Z. The low range includes materials such as beryllium and graphite, the midrange alloys such as Ti-6Al-4V, gum metal and super-Invar and finally the high-Z range pure tungsten and tantalum. Of interest in assessing proton irradiation effects are (a) changes in physiomechanical properties which are important in maintaining high-power target functionality, (b) identification of possible limits of proton flux or fluence above which certain materials cease to maintain integrity, (c) the role of material operating temperature in inducing or maintaining radiation damage reversal, and (d) phase stability and microstructural changes. The paper presents excerpt results deduced from macroscopic and microscopic post-irradiation evaluation (PIE) following several irradiation campaigns conducted at the BNL 200 MeV linac and specifically at the isotope producer beam-line/target station. The microscopic PIE relied on high energy x-ray diffraction at the BNL NSLS X17B1 and NSLS II XPD beam lines. The studies reveal the dramatic effects of irradiation on phase stability in several of the materials, changes in physical properties and ductility loss as well as thermally induced radiation damage reversal in graphite and alloys such as super-Invar.

  17. Improving the thermal stability and electrical parameters of a liquid crystalline material 4-n-(nonyloxy) benzoic acid by using Li ion beam irradiation

    Science.gov (United States)

    Kumar, Satendra; Verma, Rohit; Dwivedi, Aanchal; Dhar, R.; Tripathi, Ambuj

    2018-05-01

    Li ion beam irradiation studies on a liquid crystalline material 4-n-(nonyloxy) benzoic acid (NOBA) have been carried out. The material has phase sequence of I-N-SmC-Cr. Thermodynamic studies demonstrate that an irradiation fluence of 1×1013 ions-cm-2 results in the increased thermal stability of the smectic C (SmC) phase of the material. Dielectric measurements illustrate that the transverse component of the dielectric permittivity and hence the dielectric anisotropy of the material in the nematic (N) and SmC phases are increased as compared to those of the pure material due to irradiation. UV-Visible spectrum of the irradiated material shows an additional peak along with the peak of the pure material. The observed change in the thermodynamic and electrical parameters is attributed to the conversion of some of the dimers of NOBA to monomers of NOBA due to irradiation.

  18. Irradiation tests on bitumen and bitumen coated materials; Essais d'irradiation de bitume et d'enrobes bitumineux

    Energy Technology Data Exchange (ETDEWEB)

    Tabardel-Brian, R.; Rodier, J.; Lefillatre, G. [Commissariat a l' Energie Atomique, Chusclan (France). Centre de Production de Plutonium de Marcoule

    1969-07-01

    The use of bitumen as a material for coating high-activity products calls for prior study of the resistance of bitumen to irradiation. After giving briefly the methods of preparation of bitumen- coated products, this report lists the equipment which has been used for carrying out the {beta} and {gamma} irradiations of these products, and gives the analytical results obtained as a function of the dose rates chosen and of the total integrated dose. Finally, some conclusions have been drawn concerning the best types of bitumen. It should be stressed that some bitumens apparently underwent no degradation whatsoever nor any volume increase, for a total integrated dose of 1.8 x 10{sup 10} rads. (authors) [French] Dans le cadre de l'enrobage par le bitume des produits de haute activite, il est necessaire de verifier au prealable, la tenue du bitume a l'irradiation. Apres un bref rappel de la preparation des enrobes bitumineux, le present rapport regroupe les moyens qui ont ete mis en oeuvre, pour effectuer leurs irradiations sous rayonnement {beta} et sous rayonnement {gamma}, ainsi que les resultats des analyses qui ont ete faites en fonction des debits de doses choisis et de la dose totale integree. Enfin, des conclusions ont ete tirees sur les types de bitume a retenir. On peut souligner que certains bitumes n'ont subi aucune degradation apparente ni augmentation de volume pour une dose totale integree de 1.8 x 10{sup 10} rads. (auteurs)

  19. A performance test of a capsule for a material irradiation in the OR holes of HANARO

    International Nuclear Information System (INIS)

    Cho, M. S.; Choo, K. N.; Shin, Y. T.; Sohn, J. M.; Park, S. J.; Kang, Y. H.; Kim, B. G.

    2008-01-01

    A test for a pressure drop and a vibration was performed to develop a material capsule for an irradiation at the OR hole in HANARO. It was analyzed before the test that a diameter of a material capsule for the OR holes should be more than 49mm by an evaluation of a flow rate and pressure drop in theory. According to this estimation, 3 kinds of mock-up capsules with a diameter of 52, 54, 56 mm were made and applied to a pressure drop test. As a result of the pressure drop test, the requirement for a pressure and a flow rate in HANARO was confirmed to be satisfied for the 3 kinds of diameters. The capsules with diameters of 54, 56mm were applied to a vibration test by taking into consideration a receptive capacity of the specimens. The capsule with a diameter of 56mm satisfied the requirement for an allowable limit of the vibration acceleration applied in HANARO. The heat transfer coefficient and the temperature on the surface of a capsule were estimated. As the temperature on the surface of the capsule was calculated to be 43.7 .deg. C, the ONB condition in HANARO was satisfied

  20. Irradiation study of different silicon materials for the CMS tracker upgrade

    CERN Document Server

    Erfle, Joachim; Hansen, Wolfgang; Garutti, Erika

    Around 2022, an upgrade of the LHC collider complex is planned to significantly increase the luminosity (the High Luminosity LHC, HL-LHC). This means that the experiments have to cope with a higher number of collisions per bunch crossing and survive in a radiation environment much harsher than that at the present LHC. Especially the tracking detectors have to be improved for the HL-LHC. The increased number of tracks requires an increase of the number of readout channels while the higher radiation makes new sensor materials necessary. Within CMS, a measurement campaign was initiated to study the performance of different silicon materials in a corresponding radiation environment. To simulate the expected radiation the samples were irradiated with neutrons and with protons with two different energies. Radiation damage can be divided in two categories. First, ionizing energy loss in the surface isolation layers of the sensor leads to a change of the concentration of charged states in the sensor surface and there...

  1. Ion-irradiation-induced damage in nuclear materials: Case study of a-SiO2 and MgO

    International Nuclear Information System (INIS)

    Bachiller-Perea, Diana

    2016-01-01

    One of the most important challenges in Physics today is the development of a clean, sustainable, and efficient energy source that can satisfy the needs of the actual and future society producing the minimum impact on the environment. For this purpose, a huge international research effort is being devoted to the study of new systems of energy production; in particular, Generation IV fission reactors and nuclear fusion reactors are being developed. The materials used in these reactors will be subjected to high levels of radiation, making necessary the study of their behavior under irradiation to achieve a successful development of these new technologies. In this thesis two materials have been studied: amorphous silica (a-SiO 2 ) and magnesium oxide (MgO). Both materials are insulating oxides with applications in the nuclear energy industry. High-energy ion irradiations have been carried out at different accelerator facilities to induce the irradiation damage in these two materials; then, the mechanisms of damage have been characterized using principally Ion Beam Analysis (IBA) techniques. One of the challenges of this thesis was to develop the Ion Beam Induced Luminescence or iono-luminescence (which is not a widely known IBA technique) and to apply it to the study of the mechanisms of irradiation damage in materials, proving the power of this technique. For this purpose, the iono-luminescence of three different types of silica (containing different amounts of OH groups) has been studied in detail and used to describe the creation and evolution of point defects under irradiation. In the case of MgO, the damage produced under 1.2 MeV Au + irradiation has been characterized using Rutherford backscattering spectrometry in channeling configuration and X-ray diffraction. Finally, the iono-luminescence of MgO under different irradiation conditions has also been studied.The results obtained in this thesis help to understand the irradiation-damage processes in materials

  2. Effect of LED-LCU light irradiance distribution on mechanical properties of resin based materials

    Energy Technology Data Exchange (ETDEWEB)

    Magalhães Filho, T.R.; Weig, K.M. [Faculdade de Odontologia, Universidade Federal Fluminense, Rua São Paulo 28, CEP 24020-150 Niterói (Brazil); Engenharia Metalúrgica e de Materiais (COPPE), Universidade Federal do Rio de Janeiro, CP 68505, CEP: 21941-972 Rio de Janeiro (Brazil); Costa, M.F. [Engenharia Metalúrgica e de Materiais (COPPE), Universidade Federal do Rio de Janeiro, CP 68505, CEP: 21941-972 Rio de Janeiro (Brazil); Werneck, M.M. [Engenharia Elétrica (COPPE), Universidade Federal do Rio de Janeiro, CP 68504, CEP: 21941-972 Rio de Janeiro (Brazil); Barthem, R.B. [Instituto de Física, Universidade Federal do Rio de Janeiro, CP 68528, CEP: 21941-972 Rio de Janeiro (Brazil); Costa Neto, C.A., E-mail: celio@metalmat.ufrj.br [Engenharia Metalúrgica e de Materiais (COPPE), Universidade Federal do Rio de Janeiro, CP 68505, CEP: 21941-972 Rio de Janeiro (Brazil)

    2016-06-01

    The objective of this study is to analyze the light power distribution along the tip end of the light guide of three LED-LCUs (Light Curing Units) and to evaluate its effect on the mechanical properties of a polymer based dental composite. Firstly, the light power distribution over the whole area of LED-LCU light guide surface was analyzed by three methods: visual projection observation, spectral measurement and optical spectral analysis (OSA). The light power distribution and the total irradiance were different for the three LEDs used, but the wavelength was within the camphorquinone absorption spectrum. The use of a blank sheet was quite on hand to make a qualitative analysis of a beam, and it is costless. Secondly, specimens of a hybrid composite with approximately 8 mm diameter and 2 mm thickness were produced and polymerized by 20 s exposition time to each LED-LCU. Thirdly, the elastic modulus (E) and hardness (HV) were measured throughout the irradiated area by instrumented micro-indentation test (IIT), allowing to correlate localized power and mechanical properties. Both E and HV showed to be very sensitive to local power and wavelength dependent, but they followed the beam power profile. It was also shown that the mechanical properties could be directly correlated to the curing process. Very steep differences in mechanical properties over very short distances may impair the material performance, since residual stresses can easily be built over it. - Highlights: • A resin based composite (RBC) was polymerized by three different Light Emitting Diodes. • Each LED had its beam profile visually, wavelength and power analyzed. • The effective polymerization power (EPP) varied from 28% to 52% of the total beam power. • Wavelength seems to be as relevant as power in the light curing process. • Mechanical properties depend on the simultaneous effect of wavelength and power.

  3. Effect of LED-LCU light irradiance distribution on mechanical properties of resin based materials

    International Nuclear Information System (INIS)

    Magalhães Filho, T.R.; Weig, K.M.; Costa, M.F.; Werneck, M.M.; Barthem, R.B.; Costa Neto, C.A.

    2016-01-01

    The objective of this study is to analyze the light power distribution along the tip end of the light guide of three LED-LCUs (Light Curing Units) and to evaluate its effect on the mechanical properties of a polymer based dental composite. Firstly, the light power distribution over the whole area of LED-LCU light guide surface was analyzed by three methods: visual projection observation, spectral measurement and optical spectral analysis (OSA). The light power distribution and the total irradiance were different for the three LEDs used, but the wavelength was within the camphorquinone absorption spectrum. The use of a blank sheet was quite on hand to make a qualitative analysis of a beam, and it is costless. Secondly, specimens of a hybrid composite with approximately 8 mm diameter and 2 mm thickness were produced and polymerized by 20 s exposition time to each LED-LCU. Thirdly, the elastic modulus (E) and hardness (HV) were measured throughout the irradiated area by instrumented micro-indentation test (IIT), allowing to correlate localized power and mechanical properties. Both E and HV showed to be very sensitive to local power and wavelength dependent, but they followed the beam power profile. It was also shown that the mechanical properties could be directly correlated to the curing process. Very steep differences in mechanical properties over very short distances may impair the material performance, since residual stresses can easily be built over it. - Highlights: • A resin based composite (RBC) was polymerized by three different Light Emitting Diodes. • Each LED had its beam profile visually, wavelength and power analyzed. • The effective polymerization power (EPP) varied from 28% to 52% of the total beam power. • Wavelength seems to be as relevant as power in the light curing process. • Mechanical properties depend on the simultaneous effect of wavelength and power.

  4. Self-shielding and burn-out effects in the irradiation of strongly-neutron-absorbing material

    International Nuclear Information System (INIS)

    Sekine, T.; Baba, H.

    1978-01-01

    Self-shielding and burn-out effects are discussed in the evaluation of radioisotopes formed by neutron irradiation of a strongly-neutron-absorbing material. A method of the evaluation of such effects is developed both for thermal and epithermal neutrons. Gadolinium oxide uniformly mixed with graphite powder was irradiated by reactor-neutrons together with pieces of a Co-Al alloy wire (the content of Co being 0.475%) as the neutron flux monitor. The configuration of the samples and flux monitors in each of two irradiations is illustrated. The yields of activities produced in the irradiated samples were determined by the γ-spectrometry with a Ge(Li) detector of a relative detection efficiency of 8%. Activities at the end of irradiation were estimated by corrections due to pile-up, self-absorption, detection efficiency, branching ratio, and decay of the activity. Results of the calculation are discussed in comparison with the observed yields of 153 Gd, 160 Tb, and 161 Tb for the case of neutron irradiation of disc-shaped targets of gadolinium oxide. (T.G.)

  5. A radio-crystallographic apparatus for the examination of irradiated materials; Appareil de radiocristallographie pour examens de materiaux irradies

    Energy Technology Data Exchange (ETDEWEB)

    Bloch, J [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    Description of a radio-crystallographic apparatus designed for the examination of irradiated metals. - The double diffraction method was used: a monochromator is placed in the beam of X-rays diffracted by the sample, thus enabling lead shielding to be interposed between the sample and the counter. In this way a very large proportion of the radiation due to the activity of the sample is eliminated. By this method substances with activities of the order of several tens of curies may be examined, and the relative precision obtained in measurements of parameters with this apparatus is equal to 10{sup -4} in the case of good crystals. (author) [French] Description d'un appareil de radiocristallographie destine a l'examen de metaux irradies. - Nous avons utilise la methode de double diffraction: un monochromateur est place dans le faisceau de rayons X diffracte par l'echantillon, ce qui permet d'interposer une protection de plomb entre l'echantillon et le compteur. On elimine ainsi dans une tres grande proportion le rayonnement du a la radioactivite de l'echantillon. On peut ainsi examiner des substances d'une activite de l'ordre de quelques dizaines de curies et la precision relative obtenue pour les mesures de parametres avec cet appareil est egale a 10{sup -4} dans le cas de bons cristaux. (auteur)

  6. Investigation of special capsule technologies for material in-pile irradiation test and development plan in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, M. S.; Son, J. M.; Kim, D. S.; Park, S. J.; Cho, Y. G.; Seo, C. K.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    In-pile test for several materials such as Zr alloy, stainless steel, Cr-Ni steel etc. which are used as structural material of the advanced reactor and KNGR(Korea Next Generation Reactor) like SMART, is necessary to produce the design data for developing new reactor materials. Advanced countries like USA, Europe and Japan etc. are not only performing the simple irradiation test for materials, but developing many kinds of special capsule to perform in-pile test having special purpose. For the special test items of fuel rod, fission products, total heat generation, swelling, deformation, sweep gas, temperature ramping and BOCA etc. are being actively concerned. There are capsules measuring creep, fatigue, crack growth, and controlling fluence etc. for special irradiation test of materials. In addition, the advanced countries are developing several instrument technologies suitable for the special capsules. In HANARO, non-instrumented, instrumented material capsules and non-instrumented fuel capsule have been developed and they have been utilized in the irradiation test for users, and creep capsule loading single specimen was made and is planned to test in the reactor soon. For some forthcoming years, special capsules not only measuring creep deformation with multi-specimens, fatigue, controlling fluence but crack propagation and gas sweep considering the requirements of users will be developed in HANARO.

  7. The effects of irradiance and exposure time on the surface roughness of bulk-fill composite resin restorative materials

    Science.gov (United States)

    Alkhudhairy, Fahad I.

    2018-01-01

    Objectives: To evaluate the surface roughness of 4 different bulk-fill resin-based composites cured using different irradiance levels. Methods: This in vitro study was performed in February 2017 to August 2017 at the College of Dentistry, King Saud University. Twenty-four specimens were prepared from each of the bulk-fill materials [Tetric N-Ceram (TNC), SonicFill (SF), Smart Dentin Replacement (SDR), and Filtek Bulk-Fill (FB)] using a brass metal mold, resulting in a total of 96 specimens, cured using a Bluephase N light curing unit. Half of the total number of specimens (N=48) were cured using high-power irradiance (1200 mW/cm2) for 20 seconds, while the remaining half (N=48) were cured using low power irradiance (650 mW/cm2) for 40 seconds. After 24 hours, baseline surface roughness of each specimen was analyzed using a profilometer, then polished using Sof-lex abrasive disks, and the surface roughness of all groups was assessed. Results: Post-polished SonicFill cured at high irradiance had the highest mean surface roughness (0.23±0.03), whereas pre-polished Smart Dentin Replacement (0.11±0.01) and SonicFill (0.11±0.02) cured at low irradiance had the lowest mean surface roughness. Conclusion: High curing irradiance (1,200 mW/cm2) had no positive influence on the surface roughness of Filtek Bulk Fill and Tetric N-Ceram bulk-fill RBCs compared with lower curing irradiance (650 mW/cm2). However, the difference of curing irradiance significantly affected the surface roughness in SDR and sonic fill RBCs. PMID:29436570

  8. 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

  9. Study of dielectric materials irradiated with electron beam by using the Pulsed Electro-Acoustic (PEA) method

    International Nuclear Information System (INIS)

    Nguyen, Xuan Truong

    2014-01-01

    Dielectric materials are frequently used as electrical insulators in spatial applications. Due to their dielectric nature, these dielectrics are likely to accumulate electric charges during their service. Under certain critical conditions, these internal or surface space charges can lead to an electrostatic surface discharge. To understand these phenomena, an experimental device has been developed in the laboratory. This device allows us to simulate the electronic irradiation conditions encountered in space. The aim of our study is to characterize the electrical behavior of insulating materials irradiated by electron beam, to investigate charge storage and transport phenomena and anticipate electrostatic discharges. In this work, the device based on the Pulsed Electro-Acoustic (PEA) technique has been chosen. It has been implanted in the irradiation chamber. It allows us to obtain the spatial distribution of charges injected between two periods of irradiation and during relaxation. However the PEA method offers a limited resolution and does not allow the detection of injected charges when they are too close to the surface. First, we performed a parameters signal processing analysis that we will call the spreading factor and the resolution factor. The preliminary study post-irradiation in air of experimental measurements showed that the resolution factor choice is important for the analysis and interpretation of the signal when the space charge is localized near the surface. Then, a comparison to the spreading parameter used in some deconvolution technique was established. In the second time, space charge distribution measurements in vacuum have been carried out on Poly Tetra Fluoro Ethylene (PTFE) films irradiated by an electron beam in the range [10-100] keV. Results from irradiation periods with increasing energies [10 keV → 100 keV] of the electron beam have been compared with results from irradiation periods with decreasing energies [100 keV → 10 keV]. In

  10. Development of mutation techniques for farm product ingredients and production of new raw materials by irradiation

    International Nuclear Information System (INIS)

    Morishita, T.

    2004-01-01

    An increase of mutation for anti-oxidization competence of buckwheat was investigated by gamma-ray irradiation. An individual and a family of the buckwheat which showed an excellent anti-oxidization competence was selected. Survival rate of irradiated buckwheat decreased at acute irradiation dose of >200 Gy. The divergence of anti-oxidization competence on the irradiated buckwheat showed large value in comparison with untreated buckwheat. The average value of anti-oxidization competence on the irradiated buckwheat was higher than the one on the untreated buckwheat. The relations between anti-oxidization competence and total irradiation dose, however, were not clear. The relations between the anti-oxidization competence and total polyphenol contents in the backwheat showed clearly a positive correlation. (M. Suetake)

  11. Chemical and physical change of packing materials for food induced by γ-ray irradiation

    International Nuclear Information System (INIS)

    Kawamura, Yoko; Sugita, Takiko; Yamada, Takashi

    1997-01-01

    28 kinds of antioxidants and UV-absorbents of synthetic resin with a kind of addition agent were stable by 30 kGy of irradiation. When polyethylene sheet incorporated with them were irradiated, amounts of many kinds of antioxidants largely decreased. They might be decomposed by reacting with radicals and peroxides in polymer produced by irradiation. Accordingly, they acted effectively to reduce deterioration of polymer. But UV-absorbent was not changed by irradiation. The addition agents of polyethylene sheet were not eluted by pseudo-food aqueous solvent independent on irradiation. However, on pseudo-fatty food solvent, amount of eluate decreased. So that, eluate of antioxidant and UV-absorbent in polyethylene was decreased by γ-ray irradiation which produced good effects on food hygiene. (S.Y.)

  12. Toxicological safety evaluation of biomolecules and materials transformed by gamma irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Il Jun; Jeon, Young Eun; Kang, Hyo Jin; Yun, Sung Bok

    2010-01-15

    In the bacterial reversion assay with S. typhimurium TA98, TA100, TA1535 and TA1537, gamma irradiated hyaluronic acid (10 and 50 kGy) did not induce a significant increase in the number of revertant colonies in the presence of S9 metabolic activation system. In chromosomal aberration tests with CHO cells, gamma irradiated hyaluronic acid (10 and 50 kGy) did not result in an increase in the frequency of chromosomal aberrations. In vivo mouse micronucleus assay, gamma irradiated hyaluronic acid (10 and 50 kGy) did not show an increase in the frequency of polychromatic erythrocytes with micronuclei. These results indicate that hyaluronic acids irradiated at 10 and 50 kGy did not show any genotoxic effects under these experimental conditions. In order to evaluate their possible subacute toxicity, the male and female of ICR mouse were given to methanol extract of 50 kGy irradiated red ginseng and 20 kGy irradiated water extract of mistletoe for three months. During the experimental periods, appearance, behavior, mortality, food and water consumption of rats fed the 50 kGy irradiated red ginseng and 20 kGy irradiated water extract of mistletoe were not affected compared to the non-irradiated control. Although minor changes in biochemical parameters were observed, they were not dose dependent and not affected by gamma irradiation. These results indicate that 50 kGy irradiated red ginseng and 20 kGy irradiated water extract of mistletoe did not show any toxic effects under these experimental conditions

  13. Toxicological safety evaluation of biomolecules and materials transformed by gamma irradiation

    International Nuclear Information System (INIS)

    Kang, Il Jun; Jeon, Young Eun; Kang, Hyo Jin; Yun, Sung Bok

    2010-01-01

    In the bacterial reversion assay with S. typhimurium TA98, TA100, TA1535 and TA1537, gamma irradiated hyaluronic acid (10 and 50 kGy) did not induce a significant increase in the number of revertant colonies in the presence of S9 metabolic activation system. In chromosomal aberration tests with CHO cells, gamma irradiated hyaluronic acid (10 and 50 kGy) did not result in an increase in the frequency of chromosomal aberrations. In vivo mouse micronucleus assay, gamma irradiated hyaluronic acid (10 and 50 kGy) did not show an increase in the frequency of polychromatic erythrocytes with micronuclei. These results indicate that hyaluronic acids irradiated at 10 and 50 kGy did not show any genotoxic effects under these experimental conditions. In order to evaluate their possible subacute toxicity, the male and female of ICR mouse were given to methanol extract of 50 kGy irradiated red ginseng and 20 kGy