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Sample records for candidate container materials

  1. An annotated history of container candidate material selection

    International Nuclear Information System (INIS)

    McCright, R.D.

    1988-07-01

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

  2. Candidate container materials for Yucca Mountain waste package designs

    International Nuclear Information System (INIS)

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

    1991-09-01

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

  3. Corrosion of candidate container materials by Yucca Mountain bacteria

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-09-01

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

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

    International Nuclear Information System (INIS)

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

    1988-06-01

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

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

    International Nuclear Information System (INIS)

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

    1988-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-04-01

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

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

    International Nuclear Information System (INIS)

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

    1988-04-01

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

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

    International Nuclear Information System (INIS)

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

    1988-06-01

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

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

    International Nuclear Information System (INIS)

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

    1988-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-06-01

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

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

    International Nuclear Information System (INIS)

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

    1995-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-06-01

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

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

    International Nuclear Information System (INIS)

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

    1988-06-01

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

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

    International Nuclear Information System (INIS)

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

    1988-08-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Korzan, M.A.

    1994-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

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

    International Nuclear Information System (INIS)

    Seni, C.; Mills, R.H.

    1994-01-01

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

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

    Science.gov (United States)

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

    2015-04-01

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

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

    International Nuclear Information System (INIS)

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

    1988-05-01

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

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

    International Nuclear Information System (INIS)

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

    1983-11-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Grover, L.K.

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-10-01

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

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

    Science.gov (United States)

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

    2014-12-10

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

  9. Localized corrosion and stress corrosion cracking of candidate materials for high-level radioactive waste disposal containers in U.S

    International Nuclear Information System (INIS)

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

    1989-01-01

    Three ion-based to nickel-based austenitic alloys and three copper-based alloys are being considered in the United States as candidate materials for the fabrication of high-level radioactive waste containers. The austenitic alloys are Types 304L and 316L stainless steels as well as the high-nickel material Alloy 825. The copper-based alloys are CDA 102 (oxygen-free copper) CDA 613 (Cu7Al), and CDA 715 (Cu-30Ni). Waste in the forms of spent fuel assemblies from reactors and borosilicate glass will be sent to a proposed repository at Yucca Mountain, Nevada. The decay of radionuclides will result in the generation of substantial heat and in gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including: undesirable phase transformations due to a lack of phase stability; atmospheric oxidation; general aqueous corrosion; pitting; crevice corrosion; intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This paper is an analysis of data from the literature relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of these alloys

  10. 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

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

    NARCIS (Netherlands)

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

    1999-01-01

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

  12. Localized corrosion and stress corrosion cracking of candidate materials for high-level radioactive waste disposal containers in the US: A literature review

    International Nuclear Information System (INIS)

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

    1988-01-01

    Container materials may undergo any of several modes of degradation in this environment, including: undesirable phase transformations due to lack of phase stability; atmospheric oxidation; general aqueous corrosion; pitting; crevice corrosion; intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This paper is an analysis of data from the literature relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of these alloys. Though all three austenitic candidates have demonstrated pitting and crevice corrosion in chloride-containing environments, Alloy 825 has the greatest resistance to these forms of localized attack. Both types 304L and 316L stainless steels are susceptible to SCC in acidic chloride media. In contrast, SCC has not been documented for Alloy 825 under comparable conditions. Gamma irradiation has been found to enhance SCC of Types 304 and 304L stainless steels, but it has no detectable effect on the resistance of Alloy 825 to SCC. Furthermore, while microbiologically induced corrosion effects have been observed for 300-series stainless steels, nickel-based alloys such as Alloy 825 seem to be immune to such problems. Of the copper-based alloys, CDA 715 has the best overall resistance to localized attack. Its resistance to pitting is comparable to that of CDA 613 and superior to that of CDA 102. Observed rates of dealloying in CDA 715 are less than those observed in CDA 613 by orders of magnitude. The resistance of CDA 715 to SCC in tarnishing ammonical environments is comparable to that of CDA 102 and superior to that of CDA 613. Its resistance to SCC in nontarnishing ammonical environments is comparable to that of CDA 613 and superior to that of CDA 102. 22 refs., 8 figs., 4 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-02-01

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

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

    International Nuclear Information System (INIS)

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

    1993-02-01

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

  15. Radioactive Material Containment Bags

    National Research Council Canada - National Science Library

    2000-01-01

    The audit was requested by Senator Joseph I. Lieberman based on allegations made by a contractor, Defense Apparel Services, about the Navy's actions on three contracts for radioactive material containment bags...

  16. Updated candidate list for engineered barrier materials

    International Nuclear Information System (INIS)

    McCright, R.D.

    1995-10-01

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

  17. Characterization of nanoparticles as candidate reference materials

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  18. Container for radioactive materials

    International Nuclear Information System (INIS)

    Housholder, W.R.; Greer, N.L.

    1976-01-01

    The improvement of the construction of containers for the transport of nuclear fuels is proposed where above all, the insulating mass suggested is important as it acts as a safeguard in case of an accident. The container consists of a metal casing in which there is a pressure boiler and a gamma-shielding device, spacers between the metal casing and the shielding device as well as an insulation filling the space between them. The insulating material is a water-in-resin emulsion which is hardened or cross-linked by peroxide and which can furthermore contain up to 50 wt.% solid silicious material such as vermuculite or chopped glass fibre. The construction and variations of the insulating mass composition are described in great detail. (HR) [de

  19. Material containment enclosure

    International Nuclear Information System (INIS)

    Carlson, D.O.

    1993-01-01

    An isolation enclosure and a group of isolation enclosures are described which are useful when a relatively large containment area is required. The enclosure is in the form of a ring having a section removed so that a technician may enter the center area of the ring. In a preferred embodiment, an access zone is located in the transparent wall of the enclosure and extends around the inner perimeter of the ring so that a technician can insert his hands into the enclosure to reach any point within. The inventive enclosures provide more containment area per unit area of floor space than conventional material isolation enclosures. 3 figures

  20. Consumer Products Containing Radioactive Materials

    Science.gov (United States)

    Fact Sheet Adopted: February 2010 Health Physics Society Specialists in Radiation Safety Consumer Products Containing Radioactive Materials Everything we encounter in our daily lives contains some radioactive material, ...

  1. Radioactive material transporting container

    International Nuclear Information System (INIS)

    Watabe, Yukio.

    1990-01-01

    As a supporting member of a sealing container for containing spent fuels, etc., a straight pipe or a cylinder has been used. However, upon dropping test, the supporting member is buckled toward the central axis of a transporting container and a shock absorber is crushed in the axial direction to prevent its pushing force to the outer side, which may possibly hinder normal shock moderating function. Then, at least more than one-half of the supporting member is protruded radially to the outer side of the sealing container beyond the fixed portion with the sealed container, so that the member has a portion extended in the radial outside of the transporting container with an angle greater than the angle formed between a line connecting the outer circumference at the bottom of an outer cylinder with the gravitational center of the transporting container and the central axis of the transporting container. As a result, buckling of the supporting member toward the central axis of the transporting container upon dropping test can be prevented and the deformation of the shock absorber is neither not prevented to exhibit normal shock absorbing effect. This can improve the reliability and reduce the amount of shock absorbers. (N.H.)

  2. Materials designed for containments

    International Nuclear Information System (INIS)

    Piehl, K.H.

    1976-01-01

    The present article points out that high-tensile fine-grained steels have been used successfully in the construction of reactor containments, spherical gasometers, and pressure vessels. It has been confirmed that their use requires safety measures concerning lay out and production. Viscosity properties of high-tensile, fine-grained steels can be improved significantly by means of electroslag remelting. The extent to which this improvement influences the heat-affected zone is being examined. (orig./RW) [de

  3. Container Materials, Fabrication And Robustness

    International Nuclear Information System (INIS)

    Dunn, K.; Louthan, M.; Rawls, G.; Sindelar, R.; Zapp, P.; Mcclard, J.

    2009-01-01

    The multi-barrier 3013 container used to package plutonium-bearing materials is robust and thereby highly resistant to identified degradation modes that might cause failure. The only viable degradation mechanisms identified by a panel of technical experts were pressurization within and corrosion of the containers. Evaluations of the container materials and the fabrication processes and resulting residual stresses suggest that the multi-layered containers will mitigate the potential for degradation of the outer container and prevent the release of the container contents to the environment. Additionally, the ongoing surveillance programs and laboratory studies should detect any incipient degradation of containers in the 3013 storage inventory before an outer container is compromised.

  4. Container for shipping dangerous material

    International Nuclear Information System (INIS)

    Blum, P.T.

    1987-01-01

    This container for shipping dangerous material is made by a cylindrical casing of austenitic stainless steel with rounded ends and walls of uniform thickness with welded joins, a tubular metal shock absorber fixed over each end of the casing, removable lugs fixed to the casing, optionally retainers for the material within the casing [fr

  5. Transport containers for radioactive material

    International Nuclear Information System (INIS)

    Doroszlai, P.; Ferroni, F.

    1984-01-01

    A cylindrical container for the transportation of radioactive reactor elements includes a top end, a bottom end and a pair of removable outwardly curved shock absorbers, each including a double-shelled construction having an internal shell with a convex intrados configuration and an external shell with a convex extrados configuration, the shock absorbers being filled with a low density energy-absorbing material and mounted at the top end and the bottom end of the container, respectively, and each of the shock absorbers having a toroidal configuration, and deformable tubes disposed within the shock absorbers and extending in the axial direction of the container

  6. Molten material-containing vessel

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Chanzy, Y.; Roland, V.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-04-01

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

  9. Immune reactivity of candidate reference materials

    NARCIS (Netherlands)

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

    2006-01-01

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

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

    Science.gov (United States)

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

    2018-03-01

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

  11. Immersion studies on candidate container alloys for the Tuff Repository

    International Nuclear Information System (INIS)

    Beavers, J.A.; Durr, C.L.

    1991-05-01

    Cortest Columbus Technologies (CC Technologies) is investigating the long-term performance of container materials used for high-level radioactive waste packages. This information is being developed for the Nuclear Regulatory Commission to aid in their assessment of the Department of Energy's application to construct a geologic repository for disposal of high-level radioactive waste. This report summarizes the results of exposure studies performed on two copper-base and two Fe-Cr-Ni alloys in simulated Tuff Repository conditions. Testing was performed at 90 degrees C in three environments; simulated J-13 well water, and two environments that simulated the chemical effects resulting from boiling and irradiation of the groundwater. Creviced specimens and U-bends were exposed to liquid, to vapor above the condensed phase, and to alternate immersion. A rod specimen was used to monitor corrosion at the vapor-liquid interface. The specimens were evaluated by electrochemical, gravimetric, and metallographic techniques following approximately 2000 hours of exposure. Results of the exposure tests indicated that all four alloys exhibited acceptable general corrosion rates in simulated J-13 well water. These rates decreased with time. Incipient pitting was observed under deposits on Alloy 825 and pitting was observed on both Alloy CDA 102 and Alloy CDA 715 in the simulated J-13 well water. No SCC was observed in U-bend specimens of any of the alloys in simulated J-13 well water. 33 refs., 48 figs., 23 tabs

  12. Tests of candidate materials for particle bed reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Erlstroem, M.; Pusch, R.

    1987-12-01

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

  14. Packet D: Fuel containing materials

    International Nuclear Information System (INIS)

    Tokarevskij, V.V.

    1999-01-01

    The tasks of the project 'D' are: increase of nuclear safety by fuel containing mass (FCM) characterisation, and development of a preliminary plan for FCM management which should be accomplished by FCM extraction

  15. Transport containers for radioactive material

    International Nuclear Information System (INIS)

    Bibby, D.

    1978-01-01

    A transport container for transporting irradiated nuclear fuel is described comprising a steel flask with detachable cover and having external heat exchange fins. The flask contains a solid annular shield comprised of discrete bodies of Pb or Fe bonded together by a solid matrix, for attenuating gamma rays and neutron emission. This may comprise lead shot bonded together by concrete or polyethylene, or alternatively iron shot bonded by concrete. (UK)

  16. Dry containment of radioactive materials

    International Nuclear Information System (INIS)

    Williams, C.E.

    1980-01-01

    A cask for the dry containment of radioactive fuel elements is described. The cask has a cover which contains valved drain and purge passageways. These passageways are sealed by after purge cover seals which are clamped over them and to the outer surface of the cover. The cover seals are tested by providing them with a pair of concentric ring seal elements squeezed between the cover seal and the outer surface of the cover and by forcing a gas under pressure into the annular region between the seal element

  17. Materials for high-level waste containment

    International Nuclear Information System (INIS)

    Marsh, G.P.

    1982-01-01

    The function of the high-level radioactive waste container in storage and of a container/overpack combination in disposal is considered. The consequent properties required from potential fabrication materials are discussed. The strategy adopted in selecting containment materials and the experimental programme underway to evaluate them are described. (U.K.)

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

    Science.gov (United States)

    Tuttle, JIm; Canavan, Ed; Jahromi, Amir

    2017-01-01

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

  19. Storage containers for radioactive material

    International Nuclear Information System (INIS)

    Cassidy, D.A.; Dates, L.R.; Groh, E.F.

    1981-01-01

    A radioactive material storage system is disclosed for use in the laboratory. This system is composed of the following: a flat base plate with a groove in one surface thereof and a hollow pedestal extending perpendicularly away from the other surface thereof; a sealing gasket in the groove, a cover having a filter therein and an outwardly extending flange which fits over the plate; the groove and the gasket, and a clamp for maintaining the cover and the plate are sealed together, whereby the plate and the cover and the clamp cooperate to provide a storage area for radioactive material readily accessible for use or inventory. Wall mounts are provided to prevent accidental formation of critical masses during storage

  20. Radiation Shielding Materials and Containers Incorporating Same

    Energy Technology Data Exchange (ETDEWEB)

    Mirsky, Steven M.; Krill, Stephen J.; and Murray, Alexander P.

    2005-11-01

    An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (''PYRUC'') shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

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

    International Nuclear Information System (INIS)

    Pan, Z.L.

    2010-01-01

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

  2. Radioactive materials transporting container and vehicles

    International Nuclear Information System (INIS)

    Reese, S.L.

    1980-01-01

    A container and vehicle therefor for transporting radioactive materials is provided. The container utilizes a removable system of heat conducting fins made of a light weight highly heat conductive metal, such as aluminum or aluminum alloys. This permits a substantial reduction in the weight of the container during transport, increases the heat dissipation capability of the container and substantially reduces the scrubbing operation after loading and before unloading the radioactive material from the container. The vehicle utilizes only a pair of horizontal side beams interconnecting a pair of yoke members to support the container and provide the necessary strength and safety with a minimum of weight

  3. SCC evaluation of candidate container alloys by DCB method

    International Nuclear Information System (INIS)

    Roy, A.K.; Freeman, D.C.; Lum, B.Y.; Spragge, M.K.

    1999-01-01

    The authors use a solid mechanics approach to investigate hydride formation and cracking in zirconium-niobium alloys used in the pressure tubes of CANDU nuclear reactors. In this approach, the forming hydride is assumed to be purely elastic and its volume dilation is accommodated by elasto-plastic deformation of the surrounding matrix material. The energetics of the hydride formation is revisited and the terminal solid solubility of hydrogen in solution is defined on the basis of the total elasto-plastic work done on the system by the forming hydride and the external loads. Hydrogen diffusion and probabilistic hydride formation coupled with the material deformation are modeled at a blunting crack tip under plane strain loading. A full transient finite element analysis allows for numerical monitoring of the development and expansion of the hydride zone as the externally applied loads increase. Using a Griffith fracture criterion for fracture limitiation, the reduced fracture resistance of the alloy can be predicted and the factors affecting fracture toughness quantified

  4. Space Environmental Effects on Candidate Solar Sail Materials

    Science.gov (United States)

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

    2004-01-01

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

  5. Candidate Materials Evaluation for Supercritical Water-Cooled Reactor

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  6. Behavior of candidate canister materials in deep ocean environments

    International Nuclear Information System (INIS)

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

    1977-04-01

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

  7. Asbestos-Containing Materials (ACM) and Demolition

    Science.gov (United States)

    There are specific federal regulatory requirements that require the identification of asbestos-containing materials (ACM) in many of the residential buildings that are being demolished or renovated by a municipality.

  8. Storage vessel for containing radiation contaminated material

    International Nuclear Information System (INIS)

    Ogawa, Kazuya.

    1995-01-01

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

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

    Science.gov (United States)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1988-02-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  12. Corrosion of candidate materials in Lake Rotokawa geothermal exposure

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-05-01

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

  13. Containing and discarding method for radiation contaminated materials and radiation contaminated material containing composite member

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1995-01-01

    A container for high level radiation contaminated materials is loaded in an outer container in a state of forming a gap between the outer container and a container wall, low level radiation contaminated materials are filled to the gap between the container of the radiation contaminated materials and the container wall, and then the outer container is sealed. In addition, the thickness of the layer of the low level radiation contaminated materials is made substantially uniform. Then, since radiation rays from the container of the radiation contaminated materials are decayed by the layer of the low level radiation contaminated materials at the periphery of the container and the level of the radiation rays emitted from the outer container is extremely reduced than in a case where the entire amount of high level radiation contaminated materials are filled, the level is suppressed to an extent somewhat higher than the level in the case where the entire amount of the low level radiation contaminated materials are filled. Accordingly, the management corresponds to that for the low level radiation contaminated materials, and the steps for the management and the entire volume thereof are reduced than in a case where the high level radiation contaminated materials and the low level radiation contaminated materials are sealed separately. (N.H.)

  14. Compatibility of candidate structural materials with static gallium

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  15. Surface segregation in binary alloy first wall candidate materials

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  16. Buried waste containment system materials. Final Report

    International Nuclear Information System (INIS)

    Weidner, J.R.; Shaw, P.G.

    1997-10-01

    This report describes the results of a test program to validate the application of a latex-modified cement formulation for use with the Buried Waste Containment System (BWCS) process during a proof of principle (POP) demonstration. The test program included three objectives. One objective was to validate the barrier material mix formulation to be used with the BWCS equipment. A basic mix formula for initial trials was supplied by the cement and latex vendors. The suitability of the material for BWCS application was verified by laboratory testing at the Idaho National Engineering and Environmental Laboratory (INEEL). A second objective was to determine if the POP BWCS material emplacement process adversely affected the barrier material properties. This objective was met by measuring and comparing properties of material prepared in the INEEL Materials Testing Laboratory (MTL) with identical properties of material produced by the BWCS field tests. These measurements included hydraulic conductivity to determine if the material met the US Environmental Protection Agency (EPA) requirements for barriers used for hazardous waste sites, petrographic analysis to allow an assessment of barrier material separation and segregation during emplacement, and a set of mechanical property tests typical of concrete characterization. The third objective was to measure the hydraulic properties of barrier material containing a stop-start joint to determine if such a feature would meet the EPA requirements for hazardous waste site barriers

  17. Treating agent for urea containing radioactive materials

    International Nuclear Information System (INIS)

    Ogawa, Hiroshi; Maki, Kentaro.

    1973-01-01

    Object: To add a coagulant into urea containing radioactive material to precipitate and remove the radioactive material in the urea. Structure: Iodosalt is added into urea and next, a mixed reagent in which silver ion or silver acetic ion and iron hydroxide precipitation or ferrite ion coexist is added therein. The urea is treated to have a sufficient alkaline, after which it is introduced into a basket type centrifuge formed with a filter layer in combination of an upper glass fiber layer and a lower active carbon layer. The treating agent can uniformly remove radioactive ion and radioactive chelate within urea containing inorganic salt and various metabolites. (Nakamura, S.)

  18. Mineralogical conversion of asbestos containing materials

    International Nuclear Information System (INIS)

    Pulsford, S.K.; Foltz, A.D.; Ek, R.B.

    1996-01-01

    The principal objective of the Technical Task Plan (TTP) is to demonstrate a thermal-chemical mineralogical asbestos conversion unit at the Hanford Site, which converts non-radiological asbestos containing materials (ACMs) into an asbestos-free material. The permanent thermal-chemical mineralogical conversion of ACMs to a non-toxic, non-hazardous, potentially marketable end product should not only significantly reduce the waste stream volumes but terminate the open-quotes cradle to graveclose quotes ownership liabilities

  19. Compatibility of ITER candidate structural materials with static gallium

    International Nuclear Information System (INIS)

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

    1993-12-01

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

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

    Science.gov (United States)

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

    2013-12-01

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

  1. Container for storage of environmental incompatible materials

    International Nuclear Information System (INIS)

    Ruggenthaler, P.T.

    1984-01-01

    The container consists of a cuboid chamber, closed to five sides, just as the cover made of concrete. Iron mountings for use with lifting gears are coupled with the armouring of the container. The cover is made in such a way that mountings are hidden by the recesses at its borders. Therefore it is possible to stick these boxes. Concrete employed for is enriched with sealing materials of synthetics, the box is painted too. Sensors on the outside ensure telemetering of closeness of the boxes. (J.K.) [de

  2. SULPHUR DIOXIDE LEACHING OF URANIUM CONTAINING MATERIAL

    Science.gov (United States)

    Thunaes, A.; Rabbits, F.T.; Hester, K.D.; Smith, H.W.

    1958-12-01

    A process is described for extracting uranlum from uranium containing material, such as a low grade pitchblende ore, or mill taillngs, where at least part of the uraniunn is in the +4 oxidation state. After comminuting and magnetically removing any entrained lron particles the general material is made up as an aqueous slurry containing added ferric and manganese salts and treated with sulfur dioxide and aeration to an extent sufficient to form a proportion of oxysulfur acids to give a pH of about 1 to 2 but insufficient to cause excessive removal of the sulfur dioxide gas. After separating from the solids, the leach solution is adjusted to a pH of about 1.25, then treated with metallic iron in the presence of a precipitant such as a soluble phosphate, arsonate, or fluoride.

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

    International Nuclear Information System (INIS)

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

    1984-11-01

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

  4. Reuse of Material Containing Natural Radionuclides - 12444

    Energy Technology Data Exchange (ETDEWEB)

    Metlyaev, E.G.; Novikova, N.J. [Burnasyan Federal Medical Biophysical Centre, Moscow (Russian Federation)

    2012-07-01

    Disposal of and use of wastes containing natural radioactive material (NORM) or technologically enhanced natural radioactive material (TENORM) with excessive natural background as a building material is very important in the supervision body activity. At the present time, the residents of Octyabrsky village are under resettlement. This village is located just near the Priargunsky mining and chemical combine (Ltd. 'PPGHO'), one of the oldest uranium mines in our country. The vacated wooden houses in the village are demolished and partly used as a building material. To address the issue of potential radiation hazard of the wooden beams originating from demolition of houses in Octyabrsky village, the contents of the natural radionuclides (K-40, Th-232, Ra-226, U- 238) are being determined in samples of the wooden beams of houses. The NORM contents in the wooden house samples are higher, on average, than their content in the reference sample of the fresh wood shavings, but the range of values is rather large. According to the classification of waste containing the natural radionuclides, its evaluation is based on the effective specific activity. At the effective specific activity lower 1.5 kBq/kg and gamma dose rate lower 70 μR/h, the material is not considered as waste and can be used in building by 1 - 3 classes depending upon A{sub eff} value. At 1.5 kBq/kg < A{sub eff} ≤ 4 kBq/kg (4 class), the wooden beams might be used for the purpose of the industrial building, if sum of ratios between the radionuclide specific activity and its specific activity of minimum significance is lower than unit. The material classified as the waste containing the natural radionuclides has A{sub eff} higher 1.5 kBq /kg, and its usage for the purpose of house-building and road construction is forbidden. As for the ash classification and its future usage, such usage is unreasonable, because, according to the provided material, more than 50% of ash samples are considered as

  5. Preliminary selection criteria for the Yucca Mountain Project waste package container material

    International Nuclear Information System (INIS)

    Halsey, W.G.

    1991-01-01

    The Department of Energy's Yucca Mountain Project (YMP) is evaluating a site at Yucca Mountain in Nevada for construction of a geologic repository for the storage of high-level nuclear waste. Lawrence Livermore National Laboratory's (LLNL) Nuclear Waste Management Project (NWMP) has the responsibility for design, testing, and performance analysis of the waste packages. The design is performed in an iterative manner in three sequential phases (conceptual design, advanced conceptual design, and license application design). An important input to the start of the advanced conceptual design is the selection of the material for the waste containers. The container material is referred to as the 'metal barrier' portion of the waste package, and is the responsibility of the Metal Barrier Selection and Testing task at LLNL. The selection will consist of several steps. First, preliminary, material-independent selection criteria will be established based on the performance goals for the container. Second, a variety of engineering materials will be evaluated against these criteria in a screening process to identify candidate materials. Third, information will be obtained on the performance of the candidate materials, and final selection criteria and quantitative weighting factors will be established based on the waste package design requirements. Finally, the candidate materials will be ranked against these criteria to determine whether they meet the mandated performance requirements, and to provide a comparative score to choose the material for advanced conceptual design activities. This document sets forth the preliminary container material selection criteria to be used in screening candidate materials. 5 refs

  6. Manufacturing method for radioactive material containing vessel

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  7. 77 FR 20886 - Proposed Information Collection (Advertising, Sales, and Enrollment Materials, and Candidate...

    Science.gov (United States)

    2012-04-06

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

  8. Closure for casks containing radioactive materials

    International Nuclear Information System (INIS)

    Hall, G.V.B.; Mallory, C.W.

    1990-01-01

    This patent describes an improved closure for covering and sealing an opening in a single cask for containing radioactive material, wherein the opening is characterized by a ledge. It comprises: an inner lid receivable within the opening and having a gasket means that is seatable over the ledge; an outer lid which is likewise receivable into the opening and securable therearound when the outer lid is rotated relative to the opening. The inner lid remaining stationary relative to the cask opening when the outer lid is rotated and having no torque applied thereto by the outer lid when the outer lid is rotated, and bolt means threadedly mounted through the outer lid for applying a compressive force between the inner and outer lids after the outer lid has been secured to the opening in order to depress the gasket means of the inner lid into sealing engagement with the ledge while avoiding the application of torsion between the gasket means and the ledge

  9. Containers for the transport of radioactive materials

    International Nuclear Information System (INIS)

    Bochard, C.

    1975-01-01

    The container for heat evolving radioactive materials has a metallic outer casing formed with outwardly projecting heat dissipating or cooling members, such as pins or fins, while each of its ends is formed with a flat flange which extends radially beyond the outer ends of the cooling members. A perforated wall extends between the flanges to define with same and with the periphery of the outer casing an annular space within which the cooling members are enclosed. This perforated wall is adapted to support a flexible covering sleeve the ends of which are clamped by inflatable seals between the periphery of the flanges and outer rings removably secured to the latter. Spraying means are provided within the aforesaid space to permit of projecting an uncontaminated liquid on the cooling members to cool the container before and/or while the latter is immersed in a loading and unloading pond with the sleeve mounted in position. The lower flange is provided with liquid collecting and evacuating means and compressed air may be injected into the said space to force the collected liquid outwardly. (auth)

  10. Liquid filter for liquids containing radioactive materials

    International Nuclear Information System (INIS)

    Rohleder, N.; Schwarz, F.

    1986-01-01

    A device for filtering radioactive liquids loaded with solids is described, which has a pressure-resistant housing with a lid and an incomer for the turbid liquid and a collecting space and drain for the filtrate at the bottom of the housing. A filter cartridge is present in this housing. Such a filtering device must be suitable for use in nuclear plants, must be easy to replace by remote control and must minimise the carrying over of radioactive particles. This problem should be solved by the filter cartridge consisting of a large number of horizontal filter plates stacked above one another, which carry a deep layer filter material acting in the sub-micron range. The turbid liquid runs into the centre of the stack of filter plates via a vertical central duct. The intermediate spaces between the filter places are connected to this central duct via the layer of filter material. The filter plates are sealed against one another on the outer circumference and have radial drain openings for the filtrate on the outside. The central duct is sealed at the lower end by a plate. When the filter cartridge is replaced, the radioactive waste in the filter cartridge remains safely enclosed and can be conditioned in suitable containers. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  12. Potential containment materials for liquid-lead and lead-bismuth eutectic spallation neutron source

    International Nuclear Information System (INIS)

    Park, J.J.; Butt, D.P.; Beard, C.A.

    1997-11-01

    Lead (Pb) and lead-bismuth eutectic (44Pb-56Bi) have been the two primary candidate liquid-metal target materials for the production of spallation neutrons. Selection of a container material for the liquid-metal target will greatly affect the lifetime and safety of the target subsystem. For the lead target, niobium-1 (wt%) zirconium (Nb-1Zr) is a candidate containment material for liquid lead, but its poor oxidation resistance has been a major concern. The oxidation rate of Nb-1Zr was studied based on the calculations of thickness loss due to oxidation. According to these calculations, it appeared that uncoated Nb-1Zr may be used for a one-year operation at 900 C at P O 2 = 1 x 10 -6 torr, but the same material may not be used in argon with 5-ppm oxygen. Coating technologies to reduce the oxidation of Nb-1Zr are reviewed, as are other candidate refractory metals such as molybdenum, tantalum, and tungsten. For the Pb-Bi target, three candidate containment materials are suggested based on a literature survey of the materials compatibility and proton irradiation tests: Croloy 2-1/4, modified 9Cr-1Mo, and 12Cr-1Mo (HT-9) steel. These materials seem to be used only if the lead-bismuth is thoroughly deoxidized and treated with zirconium and magnesium

  13. Manufacturing method for radioactive material containing vessel

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    Science.gov (United States)

    1980-03-10

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

  15. Low activation structural material candidates for fusion power plants

    International Nuclear Information System (INIS)

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

    1997-06-01

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

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

    International Nuclear Information System (INIS)

    Gray, L.W.

    1996-01-01

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

  17. Charging material into a container and closing the container

    International Nuclear Information System (INIS)

    Critchley, R.J.; Oldham, S.R.

    1989-01-01

    Radioactive waste material is passed from a cell through a port into a storage drum which has a mouth with a removable lid. A port door includes an electromagnet to hold the lid to the door and both are moved into the cell by a pneumatic cylinder to allow loading of the drum and after re-closure of door and lid swaging tools operate to deform the lid to form one or more lips which engage under the margin around mouth to resist separation of the lid from the drum. Seals are provided. The swaging tools, or a single swaging tool, can be rotatably mounted on the port door. (author)

  18. Selection criteria for container materials at the proposed Yucca Mountain high level nuclear waste repository

    International Nuclear Information System (INIS)

    Halsey, W.G.

    1989-11-01

    A geological repository has been proposed for the permanent disposal of the nation's high level nuclear waste at Yucca Mountain in the Nevada desert. The containers for this waste must remain intact for the unprecedented service lifetime of 1000 years. A combination of engineering, regulatory, and licensing requirements complicate the container material selection. In parallel to gathering information regarding the Yucca Mountain service environment and material performance data, a set of selection criteria have been established which compare candidate materials to the performance requirements, and allow a quantitative comparison of candidates. These criteria assign relative weighting to varied topic areas such as mechanical properties, corrosion resistance, fabricability, and cost. Considering the long service life of the waste containers, it is not surprising that the corrosion behavior of the material is a dominant factor. 7 refs

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

    Science.gov (United States)

    Johnston, J. D.

    1974-01-01

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

  20. 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

  1. VUV photoemission studies of candidate LHC vacuum chamber materials

    CERN Document Server

    Baglin, V; Collins, I R

    1998-01-01

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

  2. 30 CFR 56.16004 - Containers for hazardous materials.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Containers for hazardous materials. 56.16004 Section 56.16004 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND... Storage and Handling § 56.16004 Containers for hazardous materials. Containers holding hazardous materials...

  3. 30 CFR 57.16004 - Containers for hazardous materials.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Containers for hazardous materials. 57.16004 Section 57.16004 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND... Storage and Handling § 57.16004 Containers for hazardous materials. Containers holding hazardous materials...

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

    Science.gov (United States)

    Bovard, Francine S.; Cieslak, Wendy R.

    1987-09-01

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

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

    Science.gov (United States)

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

    2017-01-01

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

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

    Science.gov (United States)

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

    2015-07-28

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

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Fan, L.

    2015-07-01

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

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

    International Nuclear Information System (INIS)

    Freitas, M.C.

    1993-01-01

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

  10. Storage chamber for container of radiation-contaminated material

    International Nuclear Information System (INIS)

    Takakura, Masahide.

    1996-01-01

    The present invention concerns a storage chamber for containing radiation-contaminated materials in containing tubes and having cooling fluids circulated at the outer side of the containing tubes. The storage chamber comprises a gas supply means connected to the inside of the container tube for supplying a highly heat-conductive gas and a gas exhaustion means for discharging the gas present in the container tube. When containing vessels for radiation-contaminated materials are contained in the container tube, the gases present inside of the container tube is exhausted by means of the gas exhaustion means, and highly heat conductive gases are filled from the gas supply means to the space between the container tube and the containing vessels for the radiation-contaminated materials. When the temperature of the highly heat conductive gas is elevated due to the heat generation of the radiation-contaminated materials, the container tube is heated, and then cooled by the cooling fluid at the outer side of the container tube. In this case, the heat of the radiation-contaminated material-containing vessels is removed by the heat conduction by the highly heat conductive gas to reduce temperature gradient between the containing vessels and the containing tube. This can enhance the cooling effect. (T.M.)

  11. Transport and storage container for radioactive materials

    International Nuclear Information System (INIS)

    Engelstaedter, R.; Henning, E.; Storch, S.

    1987-01-01

    The container consists of a reinforced concrete wall, which is lined on the outsinde with a steel sheet and on the inside by a steel sheet several times thicker than this steel sheet. A front container opening can be closed by a reinforced concrete lid, which consists of a concrete core. This is surrounded by a lining, a covering and a lost shell. The cover projects beyond the reinforced concrete lid, so that bolts can be introduced in holes equally distributed over the extent of the cover, which can be turned in the threaded holes of threaded anchors, which are fixed on the steel sheet on the side towards the reinforced conrete wall. (orig./HP) [de

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

    CERN Document Server

    Bell, Z W

    2002-01-01

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

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

    Science.gov (United States)

    Nalbantoglu Yilmaz, Funda

    2017-01-01

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

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

    International Nuclear Information System (INIS)

    Nielsen, R.; Kuzay, T.M.

    1992-01-01

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

  15. Absorbent material for type a radioactive materials packaging containing liquids

    International Nuclear Information System (INIS)

    Saunders, G.A.

    1989-11-01

    The application of absorbent materials to the packaging and transport of liquid radioactive materials in Type A packages has not been reported in the literature. However, a significant body of research exists on absorbent materials for personal hygiene products such as diapers. Absorption capacity is dependent on both the absorbent material and the liquid being absorbed. Theoretical principles for capillary absorption in both the horizontal and the vertical plane indicate that small contact angle between the absorbent fibre and the liquid, and a small inter-fibre pore size are important. Some fluid parameters such as viscosity affect the rate of absorption but not the final absorption capacity. There appears to be little comparability between results obtained for the same absorbent and fluid using different test procedures. Test samples of materials from several classes of potential absorbents have been evaluated in this study, and shown to have a wide range of absorbent capacities. Foams, natural fibres, artificial fibres and granular materials are all potentially useful absorbents, with capacities ranging from as little as 0.86 to as much as 40.6 grams of distilled water per gram of absorbent. Two experimental procedures for evaluating the absorbent capacity of these materials have been detailed in this report, and found suitable for evaluating granular, fibrous or foam materials. Compression of the absorbent material reduces its capacity, but parameters such as relative humidity, pH, temperature, and viscosity appear to have little significant influence on capacity. When the materials were loaded to 50% of their one-minute absorbency, subsequent loss of the absorbed liquid was generally minimal. All of the absorbent materials rapidly lost their absorbed water through evaporation within twenty-four hours in still air at 21 degrees C and 50% relative humidity

  16. Sulfur containing nanoporous materials, nanoparticles, methods and applications

    Science.gov (United States)

    Archer, Lynden A.; Navaneedhakrishnan, Jayaprakash

    2018-01-30

    Sulfur containing nanoparticles that may be used within cathode electrodes within lithium ion batteries include in a first instance porous carbon shape materials (i.e., either nanoparticle shapes or "bulk" shapes that are subsequently ground to nanoparticle shapes) that are infused with a sulfur material. A synthetic route to these carbon and sulfur containing nanoparticles may use a template nanoparticle to form a hollow carbon shape shell, and subsequent dissolution of the template nanoparticle prior to infusion of the hollow carbon shape shell with a sulfur material. Sulfur infusion into other porous carbon shapes that are not hollow is also contemplated. A second type of sulfur containing nanoparticle includes a metal oxide material core upon which is located a shell layer that includes a vulcanized polymultiene polymer material and ion conducting polymer material. The foregoing sulfur containing nanoparticle materials provide the electrodes and lithium ion batteries with enhanced performance.

  17. Preliminary sensitivity analyses of corrosion models for BWIP [Basalt Waste Isolation Project] container materials

    International Nuclear Information System (INIS)

    Anantatmula, R.P.

    1984-01-01

    A preliminary sensitivity analysis was performed for the corrosion models developed for Basalt Waste Isolation Project container materials. The models describe corrosion behavior of the candidate container materials (low carbon steel and Fe9Cr1Mo), in various environments that are expected in the vicinity of the waste package, by separate equations. The present sensitivity analysis yields an uncertainty in total uniform corrosion on the basis of assumed uncertainties in the parameters comprising the corrosion equations. Based on the sample scenario and the preliminary corrosion models, the uncertainty in total uniform corrosion of low carbon steel and Fe9Cr1Mo for the 1000 yr containment period are 20% and 15%, respectively. For containment periods ≥ 1000 yr, the uncertainty in corrosion during the post-closure aqueous periods controls the uncertainty in total uniform corrosion for both low carbon steel and Fe9Cr1Mo. The key parameters controlling the corrosion behavior of candidate container materials are temperature, radiation, groundwater species, etc. Tests are planned in the Basalt Waste Isolation Project containment materials test program to determine in detail the sensitivity of corrosion to these parameters. We also plan to expand the sensitivity analysis to include sensitivity coefficients and other parameters in future studies. 6 refs., 3 figs., 9 tabs

  18. Self-closing shielded container for use with radioactive materials

    Science.gov (United States)

    Smith, J.E.

    A container for storage of radioactive material comprises a container body and a closure member. The closure member is coupled to the container body to enable the closure body to move automatically from a first position (e.g., closed) to a second position (open).

  19. Corrosion behaviour of container materials for geological disposal of high level radioactive waste

    International Nuclear Information System (INIS)

    Accary, A.

    1985-01-01

    The disposal of high level radioactive waste in geological formations, based on the multibarrier concept, may include the use of a container as one of the engineered barriers. In this report the requirements imposed on this container and the possible degradation processes are reviewed. Further on an overview is given of the research being carried out by various research centres in the European Community on the assessment of the corrosion behaviour of candidate container materials. The results obtained on a number of materials under various testing conditions are summarized and evaluated. As a result, three promising materials have been selected for a detailed joint testing programme. It concerns two highly corrosion resistant alloys, resp. Ti-Pd (0.2 Pd%) and Hastelloy C4 and one consumable material namely a low carbon steel. Finally the possibilities of modelling the corrosion phenomena are discussed

  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. Device for separating, purifying and recovering nuclear fuel material, impurities and materials from impurity-containing nuclear fuel materials or nuclear fuel containing material

    International Nuclear Information System (INIS)

    Sato, Ryuichi; Kamei, Yoshinobu; Watanabe, Tsuneo; Tanaka, Shigeru.

    1988-01-01

    Purpose: To separate, purify and recover nuclear fuel materials, impurities and materials with no formation of liquid wastes. Constitution: Oxidizing atmosphere gases are introduced from both ends of a heating furnace. Vessels containing impurity-containing nuclear fuel substances or nuclear fuel substance-containing material are continuously disposed movably from one end to the other of the heating furnace. Then, impurity oxides or material oxides selectively evaporated from the impurity-containing nuclear fuel substances or nuclear fuel substance-containing materials are entrained in the oxidizing atmosphere gas and the gases are led out externally from a discharge port opened at the intermediate portion of the heating furnace, filters are disposed to the exit to solidify and capture the nuclear fuel substances and traps are disposed behind the filters to solidify and capture the oxides by spontaneous air cooling or water cooling. (Sekiya, K.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-11-01

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

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

    Directory of Open Access Journals (Sweden)

    Zeynep Genc

    2017-04-01

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

  4. Evaluation of iron-base materials for waste package containers in a salt repository

    International Nuclear Information System (INIS)

    Westerman, R.E.; Nelson, J.L.; Kuhn, W.L.; Basham, S.G.; Moak, D.A.; Pitman, S.G.

    1983-11-01

    Design studies for high-level nuclear waste packages for salt repositories have identified low-carbon steel as a candidate material for containers. Among the requirements are strength, corrosion resistance, and fabricability. The studies of the corrosion resistance and structural stability of iron-base materials (particularly low-carbon steel) are treated in this paper. The materials have been exposed in brines that are characteristic of the potential sites for salt repositories. The effects of temperature, radiation level, oxygen level and other parameters are under investigation. The initial development of corrosion models for these environments is presented with discussion of the key mechanisms under consideration. 6 references, 5 figures

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    Science.gov (United States)

    Kiełbasa, Anna; Buszewski, Bogusław

    2017-08-01

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

  8. Complex-wide representation of material packaged in 3013 containers

    Energy Technology Data Exchange (ETDEWEB)

    Narlesky, Joshua E.; Peppers, Larry G.; Friday, Gary P.

    2009-06-01

    The DOE sites packaging plutonium oxide materials packaged according to Department of Energy 3013 Standard (DOE-STD-3013) are responsible for ensuring that the materials are represented by one or more samples in the Materials Identification and Surveillance (MIS) program. The sites categorized most of the materials into process groups, and the remaining materials were characterized, based on the prompt gamma analysis results. The sites issued documents to identify the relationships between the materials packaged in 3013 containers and representative materials in the MIS program. These “Represented” documents were then reviewed and concurred with by the MIS Working Group. However, these documents were developed uniquely at each site and were issued before completion of sample characterization, small-scale experiments, and prompt gamma analysis, which provided more detailed information about the chemical impurities and the behavior of the material in storage. Therefore, based on the most recent data, relationships between the materials packaged in 3013 containers and representative materials in the MIS program been revised. With the prompt gamma analysis completed for Hanford, Rocky Flats, and Savannah River Site 3013 containers, MIS items have been assigned to the 3013 containers for which representation is based on the prompt gamma analysis results. With the revised relationships and the prompt gamma analysis results, a Master “Represented” table has been compiled to document the linkages between each 3013 container packaged to date and its representative MIS items. This table provides an important link between the Integrated Surveillance Program database, which contains information about each 3013 container to the MIS items database, which contains the characterization, prompt gamma data, and storage behavior data from shelf-life experiments for the representative MIS items.

  9. Complex-wide representation of material packaged in 3013 containers

    International Nuclear Information System (INIS)

    Narlesky, Joshua E.; Peppers, Larry G.; Friday, Gary P.

    2009-01-01

    The DOE sites packaging plutonium oxide materials packaged according to Department of Energy 3013 Standard (DOE-STD-3013) are responsible for ensuring that the materials are represented by one or more samples in the Materials Identification and Surveillance (MIS) program. The sites categorized most of the materials into process groups, and the remaining materials were characterized, based on the prompt gamma analysis results. The sites issued documents to identify the relationships between the materials packaged in 3013 containers and representative materials in the MIS program. These 'Represented' documents were then reviewed and concurred with by the MIS Working Group. However, these documents were developed uniquely at each site and were issued before completion of sample characterization, small-scale experiments, and prompt gamma analysis, which provided more detailed information about the chemical impurities and the behavior of the material in storage. Therefore, based on the most recent data, relationships between the materials packaged in 3013 containers and representative materials in the MIS program have been revised. With the prompt gamma analysis completed for Hanford, Rocky Flats, and Savannah River Site 3013 containers, MIS items have been assigned to the 3013 containers for which representation is based on the prompt gamma analysis results. With the revised relationships and the prompt gamma analysis results, a Master 'Represented' table has been compiled to document the linkages between each 3013 container packaged to date and its representative MIS items. This table provides an important link between the Integrated Surveillance Program database, which contains information about each 3013 container to the MIS items database, which contains the characterization, prompt gamma data, and storage behavior data from shelf-life experiments for the representative MIS items

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  11. Experiments on container materials for Swiss high-level waste disposal projects. Part 2

    International Nuclear Information System (INIS)

    Simpson, J.P.

    1984-12-01

    The present concept for final disposal of high-level waste in Switzerland consists of a repository at a depth of 1000 to 1500 m in the crystalline bedrock of northern Switzerland. The waste will be placed in a container which is required to function as a high integrity barrier for at least 1000 years. This report is the second of a set of two dealing with the evaluation of potential materials for such containers. Four materials were identified for further evaluation in the first of these reports; they were cast steel, nodular cast iron, copper and Ti-Code 12. It was concluded that some testing was needed, in particular with respect to corrosion, in order to confirm these materials as candidate container materials. The experimental programme included: 1) corrosion tests on copper under gamma radiation; 2) immersion corrosion tests on the four candidate materials including welded specimens; 3) corrosion testing of the four materials in saturated bentonite; 4) constant strain rate testing of Ti-Code 12 and copper at 80 degrees C; 5) the behaviour of copper, Ti-Code 12 and Zircaloy-2 when immersed in liquid lead; 6) corrosion potential and galvanic current measurements on several material pairs. The standard test medium was natural mineral water from the Bad Saeckingen source. This water has a total dissolved solids content of approx. 3200 mg/l, about 1600 mg/l as chloride. The oxygen level was defined as 0.1 μg/g. In certain cases this medium was modified in order to test under more severe conditions. The results of the corrosion tests confirm in general the evaluation in the first part of the report. All of the materials are suitable for high-level waste containers: cast steel, nodular cast iron and copper as single layer containers, and Ti-Code 12 as an outer corrosion resistant layer. Copper could also be used under an outer steel layer, where it could arrest local penetration

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

    International Nuclear Information System (INIS)

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

    1987-03-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  14. Characterization of plutonium-containing materials and storage canisters

    International Nuclear Information System (INIS)

    Mason, R.E.

    1997-01-01

    Throughout the weapons complex, plutonium materials are stored in various containers. Some plutonium has been stored for 20 yr or more. The physical and chemical properties of the plutonium material and the containers that hold it are often not well characterized. The U.S. Department of Energy (DOE) 3013 standard sets criteria to which stored material must conform. The 3013 standard regulates materials that hold 50% or greater plutonium, and the other 50% is not specified and is usually unknown. The Materials Identification and Surveillance project is tasked to characterize representative materials and begin to characterize the other 50% and to show that materials can be brought into 3013 criteria conformance through thermal treatments

  15. Stress-corrosion-cracking studies on candidate container alloys for the Tuff Repository

    International Nuclear Information System (INIS)

    Beavers, J.A.; Durr, C.L.

    1992-05-01

    Cortest Columbus Technologies, Inc. (CC Technologies) investigated the long-term performance of container materials used for high-level waste package as part of the information needed by the Nuclear Regulatory Commission (NRC) to assess the Department of Energy's application to construct to geologic repository for high-level radioactive waste. At the direction of the NRC, the program focused on the Tuff Repository. This report summarizes the results of Stress-Corrosion-Cracking (SCC) studies performed in Tasks 3, 5, and 7 of the program. Two test techniques were used; U-bend exposures and Slow-Strain-Rate (SSR) tests. The testing was performed on two copper-base alloys (Alloy CDA 102 and Alloy CDA 175) and two Fe-Cr-Ni alloys (Alloy 304L and Alloy 825) in simulated J-13 groundwater and other simulated solutions for the Tuff Repository. These solutions were designed to simulate the effects of concentration and irradiation on the groundwater composition. All SCC testing on the Fe-Cr-Ni Alloys was performed on solution-annealed specimens and thus issues such as the effect of sensitization on SCC were not addressed

  16. Cyclodextrin-Containing Polymers: Versatile Platforms of Drug Delivery Materials

    Directory of Open Access Journals (Sweden)

    Jeremy D. Heidel

    2012-01-01

    Full Text Available Nanoparticles are being widely explored as potential therapeutics for numerous applications in medicine and have been shown to significantly improve the circulation, biodistribution, efficacy, and safety profiles of multiple classes of drugs. One leading class of nanoparticles involves the use of linear, cyclodextrin-containing polymers (CDPs. As is discussed in this paper, CDPs can incorporate therapeutic payloads into nanoparticles via covalent attachment of prodrug/drug molecules to the polymer (the basis of the Cyclosert platform or by noncovalent inclusion of cationic CDPs to anionic, nucleic acid payloads (the basis of the RONDEL platform. For each of these two approaches, we review the relevant molecular architecture and its rationale, discuss the physicochemical and biological properties of these nanoparticles, and detail the progress of leading drug candidates for each that have achieved clinical evaluation. Finally, we look ahead to potential future directions of investigation and product candidates based upon this technology.

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

    NARCIS (Netherlands)

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

    1991-01-01

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

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

    Science.gov (United States)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  20. 19 CFR 10.461 - Retail packaging materials and containers.

    Science.gov (United States)

    2010-04-01

    ... Free Trade Agreement Rules of Origin § 10.461 Retail packaging materials and containers. Packaging... classification set out in General Note 26(n), HTSUS. If the good is subject to a regional value content... non-originating materials, as the case may be, in calculating the regional value content of the good...

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

    Directory of Open Access Journals (Sweden)

    Nurhani Aryana

    2016-03-01

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

  2. Oxidation of a Silica-Containing Material in a Mach 0.3 Burner Rig

    Science.gov (United States)

    Nguyen, QuynhGiao N.; Cuy, Michael D.; Gray, Hugh R. (Technical Monitor)

    2002-01-01

    A primarily silica-containing material with traces of organic compounds, as well as aluminum and calcium additions, was exposed to a Mach 0.3 burner rig at atmospheric pressure using jet fuel. The sample was exposed for 5 continuous hours at 1370 C. Post exposure x-ray diffraction analyses indicate formation of cristobalite, quartz, NiO and Spinel (Al(Ni)CR2O4). The rig hardware is composed of a nickel-based superalloy with traces of Fe. These elements are indicated in the energy dispersive spectroscopy (EDS) results. This material was studied as a candidate for high temperature applications under an engine technology program.

  3. Effect of chloride concentration and pH on pitting corrosion of waste package container materials

    International Nuclear Information System (INIS)

    Roy, A.K.; Fleming, D.L.; Gordon, S.R.

    1996-12-01

    Electrochemical cyclic potentiodynamic polarization experiments were performed on several candidate waste package container materials to evaluate their susceptibility to pitting corrosion at 90 degrees C in aqueous environments relevant to the potential underground high-level nuclear waste repository. Results indicate that of all the materials tested, Alloy C-22 and Ti Grade-12 exhibited the maximum corrosion resistance, showing no pitting or observable corrosion in any environment tested. Efforts were also made to study the effect of chloride ion concentration and pH on the measured corrosion potential (Ecorr), critical pitting and protection potential values

  4. Test plan for buried waste containment system materials

    International Nuclear Information System (INIS)

    Weidner, J.; Shaw, P.

    1997-03-01

    The objectives of the FY 1997 barrier material work at the Idaho National Engineering and Environmental Laboratory are to (1) select a waste barrier material and verify that it is compatible with the Buried Waste Containment System Process, and (2) determine if, and how, the Buried Waste Containment System emplacement process affects the material properties and performance (on proof of principle scale). This test plan describes a set of measurements and procedures used to validate a waste barrier material for the Buried Waste Containment System. A latex modified proprietary cement manufactured by CTS Cement Manufacturing Company will be tested. Emplacement properties required for the Buried Waste Containment System process are: slump between 8 and 10 in., set time between 15 and 30 minutes, compressive strength at set of 20 psi minimum, and set temperature less than 100 degrees C. Durability properties include resistance to degradation from carbonate, sulfate, and waste-site soil leachates. A set of baseline barrier material properties will be determined to provide a data base for comparison with the barrier materials when tested in the field. The measurements include permeability, petrographic analysis to determine separation and/or segregation of mix components, and a set of mechanical properties. The measurements will be repeated on specimens from the field test material. The data will be used to determine if the Buried Waste Containment System equipment changes the material. The emplacement properties will be determined using standard laboratory procedures and instruments. Durability of the barrier material will be evaluated by determining the effect of carbonate, sulfate, and waste-site soil leachates on the compressive strength of the barrier material. The baseline properties will be determined using standard ASTM procedures. 9 refs., 1 fig., 2 tabs

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

    Science.gov (United States)

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

    2001-01-01

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

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

    DEFF Research Database (Denmark)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Susanna Tuning Sunanti; Samin; Supriyanto C

    2013-01-01

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

  8. 49 CFR 176.76 - Transport vehicles, freight containers, and portable tanks containing hazardous materials.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Transport vehicles, freight containers, and... TRANSPORTATION HAZARDOUS MATERIALS REGULATIONS CARRIAGE BY VESSEL General Handling and Stowage § 176.76 Transport... paragraphs (b) through (f) of this section, hazardous materials authorized to be transported by vessel may be...

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

    Science.gov (United States)

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

    2016-08-01

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

  10. Corrosion aspects of steel radioactive waste containers in cementitious materials

    International Nuclear Information System (INIS)

    Smart, Nick

    2012-01-01

    Nick Smart from Serco, UK, gave an overview of the effects of cementitious materials on the corrosion of steel during storage and disposal of various low- and intermediate-level radioactive wastes. Steel containers are often used as an overpack for the containment of radioactive wastes and are routinely stored in an open atmosphere. Since this is an aerobic and typically humid environment, the steel containers can start to corrode whilst in storage. Steel containers often come into contact with cementitious materials (e.g. grout encapsulants, backfill). An extensive account of different steel container designs and of steel corrosion mechanisms was provided. Steel corrosion rates under conditions buffered by cementitious materials have been evaluated experimentally. The main conclusion was that the cementitious environment generally facilitates the passivation of steel materials. Several general and localised corrosion mechanisms need to be considered when evaluating the performance of steel containers in cementitious environments, and environmental thresholds can be defined and used with this aim. In addition, the consequences of the generation of gaseous hydrogen by the corrosion of carbon steel under anoxic conditions must be taken into account. Discussion of the paper included: Is crevice corrosion really significant in cementitious systems? Crevice corrosion is unlikely in the cementitious backfill considered because it will tend to neutralise any acidic conditions in the crevice. What is the role of microbially-induced corrosion (MIC) in cementitious systems? Microbes are likely to be present in a disposal facility but their effect on corrosion is uncertain

  11. Assessment of materials for nuclear fuel immobilization containers

    Energy Technology Data Exchange (ETDEWEB)

    Nuttall, K; Urbanic, V F

    1981-09-01

    A wide range of engineering metals and alloys has been assessed for their suitability as container materials for irradiated nuclear fuel intended for permanent disposal in a deep, underground hard-rock vault. The expected range of service conditions in the disposal vault are discussed, as well as the material properties required for this application. An important requirement is that the container last at least 500 years without being breached. The assessment is treated in two parts. Part I concentrates on the physical and mechanical metallurgy, with special reference to strength, weldability, potential embrittlement mechanisms and some economic aspects. Part II discusses possible mechanisms of metallic corrosion for the various engineering alloys and the expected range of environmental conditions in the vault. Localized corrosion and delayed fracture processes are identified as being most likely to limit container lifetime. Hence an essential requirement is that such processes either be absent or proceed at an insignificant rate. Three groups of alloys are recommended for further consideration as possible container materials: AISI 300 series austenitic stainless steels, high nickel-base alloys and very dilute titanium-base alloys. Specific alloys from each group are indicated as having the optimum combination of required properties, including cost. For container designs where the outer container shell does not independently support the service loads, copper should also be considered. The final material selection will depend primarily on the enviromental conditions in the vault. 42 figures, 31 tables.

  12. Transport of bundles and equipment which contain radioactive material

    International Nuclear Information System (INIS)

    1987-01-01

    This norm settles down: 1) The requirements that should be completed in relation to safety precautions and protection against ionizing radiations during the transport radioactive material and/or equipment containing it, in order to avoid risks to the collective and the environment. 2) The basic information on procedures that will be completed in the event of happening accidents during the transport or the transit storage of radioactive material and/or equipment that contain it. 3) The measures of security and physical protection during the transport of radioactive material and/or equipment containing it. This norm is applied: 1) To all the ways of transport (by air, by ground and by ship, fluvial and marine) of radioactive material and/or equipment that contain it. 2) To all natural or legal, public or private person, devoted to install, produce, trade, market, import or export radioactive materials and/or equipment containing it, and that needs to transport them as main or secondary activity [es

  13. Nondestructive Examination Of Plutonium-Bearing Material Containers

    International Nuclear Information System (INIS)

    Yerger, L.; Mcclard, J.; Traver, L.; Grim, T.

    2010-01-01

    The first nondestructive examination (NDE) of 3013-type containers as part of the Department of Energy's (DOE's) Integrated Surveillance Program (ISP) was performed in February, 2005. Since that date 280 NDE surveillances on 255 containers have been conducted. These containers were packaged with plutonium-bearing materials at multiple DOE sites. The NDE surveillances were conducted at Hanford, Lawrence Livermore National Laboratory (LLNL), and Savannah River Site (SRS). These NDEs consisted of visual inspection, mass verification, radiological surveys, prompt gamma analysis, and radiography. The primary purpose of performing NDE surveillances is to determine if there has been a significant pressure buildup inside the inner 3013 container. This is done by measuring the lid deflection of the inner 3013 container using radiography images. These lid deflection measurements are converted to pressure measurements to determine if a container has a pressure of a 100 psig or greater. Making this determination is required by Surveillance and Monitoring Plan (S and MP). All 3013 containers are designed to withstand at least 699 psig as specified by DOE-STD-3013. To date, all containers evaluated have pressures under 50 psig. In addition, the radiography is useful in evaluating the contents of the 3013 container as well as determining the condition of the walls of the inner 3013 container and the convenience containers. The radiography has shown no signs of degradation of any container, but has revealed two packaging anomalies. Quantitative pressure measurements based on lid deflections, which give more information than the 'less than or greater than 100 psig' (pass/fail) data are also available for many containers. Statistical analyses of the pass/fail data combined with analysis of the quantitative data show that it is extremely unlikely that any container in the population of 3013 containers considered in this study (e.g., containers packaged according to the DOE-STD-3013

  14. 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)

  15. Surveillance of sealed containers with plutonium oxide materials (ms163)

    Science.gov (United States)

    Worl, Laura; Berg, John; Ford, Doris; Martinez, Max; McFarlan, Jim; Morris, John; Padilla, Dennis; Rau, Karen; Smith, Coleman; Veirs, Kirk; Hill, Dallas; Prenger, Coyne

    2000-07-01

    DOE is embarking upon a program to store large quantities of plutonium-bearing materials for up to fifty years. Materials destined for long-term storage include metals and oxides that are stabilized and packaged according to the DOE storage standard, where the packaging consists of two nested, welded, stainless steel containers. We have designed instrumented storage containers that mimic the inner storage can specified in the 3013 standard at both full- and small-scale capacities (2.4 liter and 0.005 liter, respectively), Figures 1 and 2. The containers are designed to maintain the volume to material mass ratio while allowing the gas composition and pressure to be monitored over time.

  16. Storage and transport containers for radioactive medical materials

    International Nuclear Information System (INIS)

    Suthanthiran, K.

    1989-01-01

    This patent describes a storage and transport container for small-diameter ribbon-like lengths of material including radioactive substances for use in medical treatments, comprising: an exterior shell for radiation shielding metal having top and bottom members of radiation shielding metal integral therewith; radiation shielding metal extending downward from the top of the container and forming a central cavity, the central cavity being separate from the exterior shell material of the container and extending downwardly a distance less than the height of the container; a plurality of small diameter carrier tubes located within the interior of the container and having one end of each tube opening through one side of the container and the other end of such tube opening through the opposite lateral side of the container with the central portion of each tube passing under the central cavity; and a plug of radiation shielding metal removably located in the top the central cavity for shielding the radiation from radiation sources located within the container

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  18. Effect of tablets containing probiotic candidate strains on gingival inflammation and composition of the salivary microbiome

    DEFF Research Database (Denmark)

    Keller, M K; Brandsborg, E; Holmstrøm, K

    2018-01-01

    The aim of the study was to investigate clinical and microbial effects of probiotic candidate strains in patients with moderate gingivitis. The null hypothesis was that the clinical measurements with treatment would not differ from placebo. 47 adult patients were enrolled in a randomised placebo...... were analysed with Next Generation Sequencing (NGS) and qPCR. In contrast to the placebo group, there was a significant reduction in BOP and amount of GCF (Pprobiotic test group when compared with baseline. The general PI was less affected although there was a tendency...... of decreased plaque levels in the probiotic group (P=0.05-0.09). The cytokines were unaffected by the intervention as well as the salivary microbiome. The Shannon index showed no significant differences between the groups or alterations over time. The occurrence of both probiotic strains increased in saliva...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  20. 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)

  1. System for indicating the level of material in a container

    International Nuclear Information System (INIS)

    Erb, T.L.

    1980-01-01

    In a radiation detecting system for controlling the level of material in a container, the first counter accumulates pulses generated by a geiger tube at a rate related to the level of material and a second counter accumulates clock pulses. A race condition is established between a NAND circuit indicating that the first counter has reached a predetermined total, and a NAND circuit indicating that the second counter has reached a second predetermined total representing a fixed counting interval. The first NAND circuit to respond to its predetermined total actuates a circuit to reset both counters and, if indicative of the material level being below a predetermined minimum, actuates an alarm or operates a control circuit to add material to the container. In the example shown, an additional NAND circuit responds to a different count in the first counter which count in the same time interval corresponds to a higher level, and when material is being added to the container, the race condition is between two NAND circuits. The effect of this is to provide a hysteresis effect preventing the circuit from 'hunting' around one level of material. (author)

  2. Corrosion-Resistant Container for Molten-Material Processing

    Science.gov (United States)

    Stern, Theodore G.; McNaul, Eric

    2010-01-01

    In a carbothermal process, gaseous methane is passed over molten regolith, which is heated past its melting point to a temperature in excess of 1,625 C. At this temperature, materials in contact with the molten regolith (or regolith simulant) corrode and lose their structural properties. As a result, fabricating a crucible to hold the molten material and providing a method of contact heating have been problematic. Alternative containment approaches use a large crucible and limit the heat zone of the material being processed, which is inefficient because of volume and mass constraints. Alternative heating approaches use non-contact heating, such as by laser or concentrated solar energy, which can be inefficient in transferring heat and thus require higher power heat sources to accomplish processing. The innovation is a combination of materials, with a substrate material having high structural strength and stiffness and high-temperature capability, and a coating material with a high corrosion resistance and high-temperature capability. The material developed is a molybdenum substrate with an iridium coating. Creating the containment crucible or heater jacket using this material combination requires only that the molybdenum, which is easily processed by conventional methods such as milling, electric discharge machining, or forming and brazing, be fabricated into an appropriate shape, and that the iridium coating be applied to any surfaces that may come in contact with the corrosive molten material. In one engineering application, the molybdenum was fashioned into a container for a heat pipe. Since only the end of the heat pipe is used to heat the regolith, the container has a narrowing end with a nipple in which the heat pipe is snugly fit, and the external area of this nipple, which contacts the regolith to transfer heat into it, is coated with iridium. At the time of this reporting, no single material has been found that can perform the functions of this combination

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

    International Nuclear Information System (INIS)

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

    1998-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

  5. Radiation exposure by man-modified materials containing natural radionuclides

    Energy Technology Data Exchange (ETDEWEB)

    Becker, D.E. [Technical Inspection Agency of Bavaria, Munich (Germany); Eder, E. [Government of Bavaria, Ministry for State Development and Environmental Affairs Development, Munich (Germany); Reichelt, A. [Technical Inspection Agency of Bavaria, Munich (Germany)

    1992-07-01

    More than one hundred materials, containing natural radioactive nuclides, are being investigated due to radiation exposure to people. This paper deals with thoriated gas mantles and shows that the radiation exposure by inhalation of radionuclides released while burning and exchange is not negligible. (author)

  6. 19 CFR 10.539 - Retail packaging materials and containers.

    Science.gov (United States)

    2010-04-01

    ...-Singapore Free Trade Agreement Rules of Origin § 10.539 Retail packaging materials and containers. Packaging... requirement. The United States importer of good C decides to use the build-down method, RVC=((AV−VNM)/AV... content requirement. In applying this method, the non-originating blister packages are taken into account...

  7. Ensuring the 50 year life of a fissile material container

    International Nuclear Information System (INIS)

    Glass, R.E.; Towne, T.L.

    1997-12-01

    Sandia was presented with an opportunity in 1993 to design containers for the long term storage and transport of fissile material. This program was undertaken at the direction of the US Department of Energy and in cooperation with Lawrence Livermore National Laboratory and Los Alamos National Laboratory which were tasked with developing the internal fixturing for the contents. The hardware is being supplied by Allied Signal Federal Manufacturing and Technologies, and the packaging will occur at Mason and Hangar Corporation's Pantex Plant. The unique challenge was to design a container that could be sealed with the fissile material contents; and, anytime during the next 50 years, the container could be transported with only the need for the pre-shipment leak test. This required not only a rigorous design capable of meeting the long term storage and transportation requirements, but also resulted in development of a surveillance program to ensure that the container continues to perform as designed over the 50-year life. This paper addresses the design of the container, the testing that was undertaken to demonstrate compliance with US radioactive materials transport regulations, and the surveillance program that has been initiated to ensure the 50-year performance

  8. An assessment of materials for nuclear fuel immobilization containers

    International Nuclear Information System (INIS)

    Nuttall, K.; Urbanic, V.F.

    1981-09-01

    A wide range of engineering metals and alloys was assessed for their suitability as container materials for irradiated nuclear fuel intended for permanent disposal in a deep, underground hard-rock vault. The container must last at least 500 years without being breached. Materials were assessed for their physical and mechanical metallurgy, weldability, potential embrittlement mechanisms, and economics. A study of the possible mechanisms of metallic corrosion for the various engineering alloys and the expected range of environmental conditons in the vault showed that localized corrosion and delayed fracture processes are the most likely to limit container lifetime. Thus such processes either must be absent or proceed at an insignificant rate. Three groups of alloys are recommended for further study: AISI 300 series austenitic stainless steels, high nickel-base alloys and very dilute titanium-base alloys. Specific alloys from each group are indicated as having the optimum combination of required properties, including cost. For container designs where the outer container shell does not independently support the service loads, copper should also be considered. The final material selection will depend primarily on the environmental conditions in the vault

  9. Hearth furnace for distilling powdered materials containing hydrocarbons

    Energy Technology Data Exchange (ETDEWEB)

    1937-06-21

    The present invention has for its object a hearth furnace particularly intended for the distillation of powdered material containing hydrocarbons. It consists of a fixed circular hearth above which are placed the moving scrapers intended to displace the material from the center toward the periphery. The material is poured by a central vertical pipe at the base of which is arranged a rotating ring for distributing the material on the hearth; this ring, which is fastened on the vertical axis of the drive, carries the radial arms to which are attached the scrapers arranged obliquely by the support on the arms and intended to displace the material on the hearth. The hearth is heated from below by means of forced circulation of gases produced in a fire-box and maintained at a convenient temperature by mixing with a part of the cold gases.

  10. Materials performance in off-gas systems containing iodine

    International Nuclear Information System (INIS)

    Beavers, J.A.; Berry, W.E.; Griess, J.C.

    1981-11-01

    During the reprocessing of spent reactor fuel elements, iodine is released to gas streams from which it is ultimately removed by conversion to nonvolatile iodic acid. Under some conditions iodine can produce severe corrosion in off-gas lines; in this study these conditions were established. Iron- and nickel-based alloys containing more than 6% molybdenum, such as Hastelloy G (7%), Inconel 625 (9%), and Hastelloy C-276 (16%), as well as titanium and zirconium, remained free of attack under all conditions tested. When the other materials, notably the austenitic stainless steels, were exposed to gas streams containing even only low concentrations of iodine and water vapors at 25 and 40 0 C, a highly corrosive, brownish-green liquid formed on their surfaces. In the complete absence of water vapor, the iodine-containing liquid did not form and all materials remained unaffected. The liquid that formed had a low pH (usually 2 inhibited attack

  11. Development of Latent Heat Storage Phase Change Material Containing Plaster

    Directory of Open Access Journals (Sweden)

    Diana BAJARE

    2016-05-01

    Full Text Available This paper reviews the development of latent heat storage Phase Change Material (PCM containing plaster as in passive application. Due to the phase change, these materials can store higher amounts of thermal energy than traditional building materials and can be used to add thermal inertia to lightweight constructions. It was shown that the use of PCMs have advantages stabilizing the room temperature variations during summer days, provided sufficient night ventilation is allowed. Another advantage of PCM usage is stabilized indoor temperature on the heating season. The goal of this study is to develop cement and lime based plaster containing microencapsulated PCM. The plaster is expected to be used for passive indoor applications and enhance the thermal properties of building envelope. The plaster was investigated under Scanning Electron Microscope and the mechanical, physical and thermal properties of created plaster samples were determined.

  12. Container materials for isolation of radioactive waste in salt

    International Nuclear Information System (INIS)

    Streicher, M.A.; Andrews, A.

    1987-10-01

    The workshop reviewed the extensive data on the corrosion resistance of low-carbon steel in simulated salt repository environments, determined whether these data were sufficient to recommend low-carbon steel for fabrication of the container, and assessed the suitability of other materials under consideration in the SRP. The panelists determined the need for testing and research programs, recommended experimental approaches, and recommended materials based on existing technology. On the first day of the workshop, presentations were made on waste package requirements; the expected corrosion environment; degradation processes, including a review of data from corrosion tests on carbon steel; and rationales for container design and materials, modeling studies, and planned future work. The second day was devoted to a panel caucus, presentation of workshop findings, and open discussion. 76 refs., 2 figs., 3 tabs

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

    International Nuclear Information System (INIS)

    Quigley, R.M.

    1984-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

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

    DEFF Research Database (Denmark)

    Sørensen, Søren N.; Stolpe, Mathias

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  19. Migration of TRU-containing slag into candidate refractories for waste management

    International Nuclear Information System (INIS)

    Seymour, W.C.

    1978-11-01

    Results of initial tests on transuranic migration into commercial refractories are reported. Five castable refractories and a plastic-bonded refractory were tested using the cup test method. Initial results indicate that castable materials will not survive conditions expected in the slagging pyrolysis unit; plastic refractories appear to have better performance. The major mechanism of migration into the tested refractories is pore penetration. Al 2 O 3 and Al 2 O 3 -Cr 2 O 3 phases appear resistant to chemical diffusion of transuranics. 2 figures, 16 tables

  20. Glucose from a cellulose-containing raw material

    Energy Technology Data Exchange (ETDEWEB)

    Feniksova, R V

    1977-01-01

    Glucose was produced by mixing the raw material with an enzyme and water, enzymic hydrolysis of raw material, and separation of the hydrolyzate. To increase the yield of glucose, the hydrolysis took place in 2 stages: at the 1st stage, the hydrolysis was continued until the mixture of glucose and oligosaccharides was obtained under the influence of an active enzyme (1 to 20 units/g of cellulose present in the raw material); at the 2nd stage, cellulase immobilized on an insoluble carrier (by covalent bonds) was used for the hydrolysis of oligosaccharides. The hydromodulus of the suspension of cellulose-containing raw material was maintained 5 to 30. A carrier nonhydrolyzable by cellulase, for instance a homogenious macroporous Aerosil gel, a porous glass, or porous ceramics, was used.

  1. Photometric determination of yttrium in zirconium-containing materials

    International Nuclear Information System (INIS)

    Barbina, T.M.; Polezhaev, Yu.M.

    1984-01-01

    Comparative evaluation of the effect of different ways of eliminating the zirconium interfering effect on the results of yttrium photometric determination with arsenazo 2 in artificial mixtures of Y 2 O 3 and ZrO 2 , containing 5 and 10 mol.% Y 2 O 3 , has been carried out. The effect of Zr is eliminated by means of its precipitation by ammonium solution in the form of hydroxide and using camouflaging with 25% sulfosalicylic acid. Both ways do not provide a correct enough result. The use of non-reagent thermohydrolytic Zr precipitation during the analysis of zirconium-containing materials permits to obtain correct and well-reproducible results

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

    Science.gov (United States)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Sucipta; Hadi Suntoko; Bunawas

    2007-01-01

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

  4. POLYMER COMPOSITES MODIFIED BY WASTE MATERIALS CONTAINING WOOD FIBRES

    Directory of Open Access Journals (Sweden)

    Bernardeta Dębska

    2016-11-01

    Full Text Available In recent years, the idea of sustainable development has become one of the most important require-ments of civilization. Development of sustainable construction involves the need for the introduction of innovative technologies and solutions that will combine beneficial economic effects with taking care of the health and comfort of users, reducing the negative impact of the materials on the environment. Composites obtained from the use of waste materials are part of these assumptions. These include modified epoxy mortar containing waste wood fibres, described in this article. The modification consists in the substitution of sand by crushed waste boards, previously used as underlays for panels, in quantities of 0%, 10%, 20%, 35% and 50% by weight, respectively. Composites containing up to 20% of the modifier which were characterized by low water absorption, and good mechanical properties, also retained them after the process of cyclic freezing and thawing.

  5. Disposal containers for radioactive waste materials and separation systems for radioactive waste materials

    International Nuclear Information System (INIS)

    Rubin, L.S.

    1986-01-01

    A separation system for dewatering radioactive waste materials includes a disposal container, drive structure for receiving the container, and means for releasably attaching the container to the drive structure. The separation structure disposed in the container adjacent the inner surface of the side wall structure retains solids while allowing passage of liquids. The inlet port structure in the container top wall is normally closed by first valve structure that is centrifugally actuated to open the inlet port and the discharge port structure at the container periphery receives liquid that passes through the separation structure and is normally closed by a second valve structure that is centrifugally actuated to open the discharge ports. The container also includes a coupling structure for releasable engagement with the centrifugal drive structure. The centrifugal force produced when the container is driven in rotation by the drive structure opens the valve structures, and radioactive waste material introduced into the container through the open inlet port is dewatered, and the waste is compacted. The ports are automatically closed by the valves when the container drum is not subjected to centrifugal force such that containment effectiveness is enhanced and exposure of personnel to radioactive materials is minimized. (author)

  6. Gamma motes for detection of radioactive materials in shipping containers

    International Nuclear Information System (INIS)

    Harold McHugh; William Quam; Stephan Weeks; Brendan Sever

    2007-01-01

    Shipping containers can be effectively monitored for radiological materials using gamma (and neutron) motes in distributed mesh networks. The mote platform is ideal for collecting data for integration into operational management systems required for efficiently and transparently monitoring international trade. Significant reductions in size and power requirements have been achieved for room-temperature cadmium zinc telluride (CZT) gamma detectors. Miniaturization of radio modules and microcontroller units are paving the way for low-power, deeply-embedded, wireless sensor distributed mesh networks

  7. ERG review of waste package container materials selection and corrosion

    International Nuclear Information System (INIS)

    Moak, D.P.; Perrin, J.S.

    1986-07-01

    The Engineering Review Group (ERG) was established by the Office of Nuclear Waste Isolation (ONWI) to help evaluate engineering-related issues in the US Department of Energy's nuclear waste repository program. The October 1984 meeting of the ERG reviewed the waste package container materials selection and corrosion. This report documents the ERG's comments and recommendations on these subjects and the ONWI response to the specific points raised by the ERG

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  9. 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

  10. Stowing of packages containing radioactive materials on conveyances

    International Nuclear Information System (INIS)

    Draulans, J.; Lafontaine, I.; Chevalier, G.; Gilles, P.; Jolys, J.C.; Pouard, M.

    1986-04-01

    The Commission of the European Communities has financed some research work carried out jointly by the ''Commissariat a l'Energie Atomique'' and the belgian company ''TRANSNUBEL'', in the field of stowing containers for radioactive materials on trucks. 2 reference type accidents are selected: . a front-end collision against a rigid barrier at an impact speed of 50 km/h . a side-on collision of an impacting vehicle at a speed of 25-35 km/h against a truck loaded with a container. A mathematical model has been developed by means of the CEA Trico code to compute a frontal impact in which the container (1.3 t weight), is stowed by means of 4 tie-down members, each for a nominal load of 2 t. Results indicate the stowing being insufficient and the attachment points too weak to keep the container on the platform. Real tests have been performed to verify these results. Tie-down members and chocks have been defined on the basis of static- and dynamic tests for being used in 8 crash tests. Different containers (low- and high center of gravity) and different ways of stowing have been tried out. An attempt is made to work a code of good practice for stowing, by means of tie-down members and chocks, packages on a truck platform. 19 refs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    CERN Document Server

    Cimino, R; Baglin, V

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

  16. Experiments on container materials for Swiss high-level waste disposal projects. Part IV

    International Nuclear Information System (INIS)

    Simpson, J.P.

    1989-12-01

    One concept for final disposal of high-level waste in switzerland consists of a repository at a depth of 1000 to 1500 m in the crystalline bedrock of Northern Switzerland. The waste will be placed in a container which will be required to function as a high integrity barrier for at least 100 years. This report is the fourth and last in the current series dealing with the evaluation of potential materials for such containers. Four materials were identified for further evaluation in the first of these reports: cast steel, nodular cast iron, copper and Ti-Code 12. This report deals with the problem of demonstrating that cast steel containers will not fail by stress corrosion cracking and with the problem of hydrogen produced by the reduction of water. The experimental results on pre-cracked specimens revealed no susceptibility of cast steel to stress corrosion cracking under model repository conditions. No crack growth was detected on compact DCB specimens exposed in aerobic and anaerobic groundwaters at 80 and 140 o C for 16-24 months. Cast steel remains a candidate material for high-level waste containers. As expected from thermodynamic considerations no hydrogen could be detected from copper immersed in model groundwaters at 50 o C. Hydrogen is evolved from corroding steel under anaerobic conditions. Hydrogen evolution due to corrosion of iron or steel in waste repositories has to be considered in any safety analysis; the amounts produced can be significant. Evidence todate suggests that both cast steel and copper are suitable container materials. Because the corrosion behaviour of both materials is sensitive to service conditions, in particular length of the aerobic phase, groundwater chemistry and temperature, further testing should be undertaken when a specific site has been identified. (author) 9 tabs., 11 figs., 25 refs

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

    International Nuclear Information System (INIS)

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

    1988-07-01

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

  18. Corrosion of container and infrastructure materials under clay repository conditions

    International Nuclear Information System (INIS)

    Debruyn, W.; Dresselaers, J.; Vermeiren, P.; Kelchtermans, J.; Tas, H.

    1991-01-01

    With regard to the disposal of high-level radioactive waste, it was recommended in a IAEA Technical Committee meeting to perform tests in realistic environments corresponding with normal and accidental conditions, to qualify and apply corrosion monitoring techniques for corrosion evaluation under real repository conditions and to develop corrosion and near-field evolution models. The actual Belgian experimental programme for the qualification of a container for long-term HLW storage in clay formations complies with these recommendations. The emphasis in the programme is indeed on in situ corrosion testing and monitoring and on in situ control of the near-field chemistry. Initial field experiments were performed in a near-surface clay quarry at Terhaegen. Based on a broad laboratory material screening programme and in agreement with the Commission of the European Communities, three reference materials were chosen for extensive in situ overpack testing. Ti/0.2 Pd and Hastelloy C-4 were chosen as reference corrosion resistant materials and a low-carbon steel as corrosion allowance reference material. This report summarizes progress made in the material qualification programme since the CEC contract of 1983-84. 57 Figs.; 15 Tabs.; 18 Refs

  19. Container materials in environments of corroded spent nuclear fuel

    Science.gov (United States)

    Huang, F. H.

    1996-07-01

    Efforts to remove corroded uranium metal fuel from the K Basins wet storage to long-term dry storage are underway. The multi-canister overpack (MCO) is used to load spent nuclear fuel for vacuum drying, staging, and hot conditioning; it will be used for interim dry storage until final disposition options are developed. Drying and conditioning of the corroded fuel will minimize the possibility of gas pressurization and runaway oxidation. During all phases of operations the MCO is subjected to radiation, temperature and pressure excursions, hydrogen, potential pyrophoric hazard, and corrosive environments. Material selection for the MCO applications is clearly vital for safe and efficient long-term interim storage. Austenitic stainless steels (SS) such as 304L SS or 316L SS appear to be suitable for the MCO. Of the two, Type 304L SS is recommended because it possesses good resistance to chemical corrosion, hydrogen embrittlement, and radiation-induced corrosive species. In addition, the material has adequate strength and ductility to withstand pressure and impact loading so that the containment boundary of the container is maintained under accident conditions without releasing radioactive materials.

  20. Processing Uranium-Bearing Materials Containing Coal and Loam

    Energy Technology Data Exchange (ETDEWEB)

    Civin, V; Prochazka, J [Research and Development Laboratory No. 3 of the Uranium Industry, Prague, Czechoslovakia (Czech Republic)

    1967-06-15

    Among the ores which are classified as low-grade in the CSSR are mixtures of coal and bentonitic loam of tertiary origin, containing approximately 0.1% U and with a moisture content at times well above 20-30%. The uranium is held mainly by the carbonaceous component. Conventional processing of these materials presents various difficulties which are not easily overcome. During leaching the pulp thickens and frequently becomes pasty, due to the presence of montmorillonites. Further complications arise from the high sorption capacity of the materials (again primarily due to montmorillonites) and poor sedimentation of the viscous pulps. In addition, the materials are highly refractory to the leaching agents. The paper presents experience gained in solving the problems of processing these ores. The following basic routes were explored: (1) separation of the carbonaceous and loamy components: The organic component appears to be the main activity carrier. Processing the concentrated material upon separation of the inactive or less active loam may not only remove the thixotropic behaviour but also substantially reduce the cost of the ore treatment; (2) 'liquifying' the pulps or preventing the thickening of the pulp by addition of suitable agents; (3) joint acid or carbonate processing of the materials in question with current ore types; (4) removal or suppression of thixotropic behaviour by thermal pretreatment of the material; and (5) application of the 'acid cure' method. The first method appears to be the most effective, but it presents considerable difficulties due to the extreme dispersion of the carbonaceous phase and further research is being carried out. Methods 2 and 3 proved to be unacceptable. Method 4, which includes roasting at 300-400{sup o}C, is now being operated on an industrial scale. The final method has also shown definite advantages for particular deposits of high montmorillonite content material. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-09-01

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

  2. Contained scanning electron microscope facility for examining radioactive materials

    International Nuclear Information System (INIS)

    Hsu, C.W.

    1986-03-01

    At the Savannah River Laboratory (SRL) radioactive solids are characterized with a scanning electron microscope (SEM) contained in a glove box. The system includes a research-grade Cambridge S-250 SEM, a Tracor Northern TN-5500 x-ray and image analyzer, and a Microspec wavelength-dispersive x-ray analyzer. The containment facility has a glove box train for mounting and coating samples, and for housing the SEM column, x-ray detectors, and vacuum pumps. The control consoles of the instruments are located outside the glove boxes. This facility has been actively used since October 1983 for high alpha-activity materials such as plutonium metal and plutonium oxide powders. Radioactive defense waste glasses and contaminated equipment have also been examined. During this period the facility had no safety-related incidents, and personnel radiation exposures were maintained at less than 100 mrems

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-01

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

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

    International Nuclear Information System (INIS)

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

    1987-10-01

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

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

    Science.gov (United States)

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

    2011-01-05

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

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

    International Nuclear Information System (INIS)

    Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

    2015-01-01

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

  7. Stress corrosion cracking tests on high-level-waste container materials in simulated tuff repository environments

    International Nuclear Information System (INIS)

    Abraham, T.; Jain, H.; Soo, P.

    1986-06-01

    Types 304L, 316L, and 321 austenitic stainless steel and Incoloy 825 are being considered as candidate container materials for emplacing high-level waste in a tuff repository. The stress corrosion cracking susceptibility of these materials under simulated tuff repository conditions was evaluated by using the notched C-ring method. The tests were conducted in boiling synthetic groundwater as well as in the steam/air phase above the boiling solutions. All specimens were in contact with crushed Topopah Spring tuff. The investigation showed that microcracks are frequently observed after testing as a result of stress corrosion cracking or intergranular attack. Results showing changes in water chemistry during test are also presented

  8. Shipping container for a bottle of radioactive material

    International Nuclear Information System (INIS)

    Soldan, D.W.

    1975-01-01

    A shipping container for a bottle of radioactive material comprises a can having a cavity therein for receiving the bottle, a screw cap for the can and the cavity, and an annular bottle retainer extending downwardly in the cavity from its upper end having an outwardly extending flange at its upper end clamped between the cap and the upper end of the cavity and an inwardly extending flange at its lower end receiving the neck of the bottle. The cap carries a sponge to absorb spillage from the bottle. (U.S.)

  9. Radioactive material dry-storage facility and radioactive material containing method

    International Nuclear Information System (INIS)

    Kanai, Hidetoshi; Kumagaya, Naomi; Ganda, Takao.

    1997-01-01

    The present invention provides a radioactive material dry storage facility which can unify the cooling efficiency of a containing tube and lower the pressure loss in a storage chamber. Namely, a cylindrical body surrounds a first containing tube situated on the side of an air discharge portion among a plurality of containing tubes and forms an annular channel extending axially between the cylindrical body and the first containing tube. An air flow channel partitioning member is disposed below a second containing tube situated closer to an air charging portion than the first containing tube. A first air flow channel is formed below the air channel partitioning member extending from the air charging portion to the annular channel. The second air channel is formed above the air channel partitioning member and extends from the air charging portion to the air discharge portion by way of a portion between the second containing tubes and the portion between the cylindrical body and the first containing tube. Then, low temperature air can be led from the air charging portion to the periphery of the first containing tube. The effect of cooling the first containing tube can be enhanced. The difference between the cooling efficiency between the second containing tube and the first containing tube is decreased. (I.S.)

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

    Science.gov (United States)

    Nettles, A. T

    2000-01-01

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

  11. Modification of clay-based waste containment materials

    International Nuclear Information System (INIS)

    Adu-Wusu, K.; Whang, J.M.; McDevitt, M.F.

    1997-01-01

    Bentonite clays are used extensively for waste containment barriers to help impede the flow of water in the subsurface because of their low permeability characteristics. However, they do little to prevent diffusion of contaminants, which is the major transport mechanism at low water flows. A more effective way of minimizing contaminant migration in the subsurface is to modify the bentonite clay with highly sorptive materials. Batch sorption studies were conducted to evaluate the sorptive capabilities of organo-clays and humic- and iron-based materials. These materials proved to be effective sorbents for the organic contaminants 1,2,4-trichlorobenzene, nitrobenzene, and aniline in water, humic acid, and methanol solution media. The sorption capacities were several orders of magnitude greater than that of unmodified bentonite clay. Modeling results indicate that with small amounts of these materials used as additives in clay barriers, contaminant flux through walls could be kept very small for 100 years or more. The cost of such levels of additives can be small compared to overall construction costs

  12. Performance of Cement Containing Laterite as Supplementary Cementing Material

    Directory of Open Access Journals (Sweden)

    Abbas Bukhari, Z. S.

    2013-03-01

    Full Text Available The utilization of different industrial waste, by-products or other materials such as ground granulated blast furnace slag, silica fume, fly ash, limestone, and kiln dust, etc. as supplemen- tary cementing materials has received considerable attention in recent years. A study has been conducted to look into the performance of laterite as Supplementary Cementing Materials (SCM. The study focuses on compressive strength performance of blended cement containing different percentage of laterite. The cement is replaced accordingly with percentage of 2 %, 5 %, 7 % and 10 % by weight. In addition, the effect of use of three chemically different laterites have been studied on physical performance of cement as in setting time, Le-Chatlier expansion, loss on ignition, insoluble residue, free lime and specifically compressive strength of cement cubes tested at the age of 3, 7, and 28 days. The results show that the strength of cement blended with laterite as SCM is enhanced. Key words: Portland cement, supplementary cementing materials (SCM, laterite, compressive strength KUI – 6/2013 Received January 4, 2012 Accepted February 11, 2013

  13. Corrosion of container materials for disposal of high-level radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Chun, K.S.; Park, H.S.; Yeon, J.W.; Ha, Y.K. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    In the corrosion aspect of container for the deep geological disposal of high-level radioactive waste, disposal concepts and the related container materials, which have been developed by advanced countries, have been reviewed. The disposal circumstances could be divided into the saturated and the unsaturated zones. The candidate materials in the countries, which consider the disposal in the unsaturated zone, are the corrosion resistant materials such as supper alloys and stainless steels, but those in the saturated zone is cupper, one of the corrosion allowable materials. By the results of the pitting corrosion test of sensitized stainless steels (such as 304, 304L, 316 and 316L), pitting potential is decreased with the degree of sensitization and the pitting corrosion resistance of 316L is higher than others. And so, the long-term corrosion experiment with 316L stainless steel specimens, sebsitized and non-sensitized, under the compacted bentonite and synthetic granitic groundwater has been being carried out. The results from the experiment for 12 months indicate that no evidence of pitting corrosion of the specimens has been observed but the crevice corrosion has occurred on the sensitized specimens even for 3 months. (author). 33 refs., 19 figs., 10 tabs.

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

    International Nuclear Information System (INIS)

    Malang, S.; Mattas, R.

    1994-06-01

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

  15. Detecting nuclear materials smuggling: performance evaluation of container inspection policies.

    Science.gov (United States)

    Gaukler, Gary M; Li, Chenhua; Ding, Yu; Chirayath, Sunil S

    2012-03-01

    In recent years, the United States, along with many other countries, has significantly increased its detection and defense mechanisms against terrorist attacks. A potential attack with a nuclear weapon, using nuclear materials smuggled into the country, has been identified as a particularly grave threat. The system for detecting illicit nuclear materials that is currently in place at U.S. ports of entry relies heavily on passive radiation detectors and a risk-scoring approach using the automated targeting system (ATS). In this article we analyze this existing inspection system and demonstrate its performance for several smuggling scenarios. We provide evidence that the current inspection system is inherently incapable of reliably detecting sophisticated smuggling attempts that use small quantities of well-shielded nuclear material. To counter the weaknesses of the current ATS-based inspection system, we propose two new inspection systems: the hardness control system (HCS) and the hybrid inspection system (HYB). The HCS uses radiography information to classify incoming containers based on their cargo content into "hard" or "soft" containers, which then go through different inspection treatment. The HYB combines the radiography information with the intelligence information from the ATS. We compare and contrast the relative performance of these two new inspection systems with the existing ATS-based system. Our studies indicate that the HCS and HYB policies outperform the ATS-based policy for a wide range of realistic smuggling scenarios. We also examine the impact of changes in adversary behavior on the new inspection systems and find that they effectively preclude strategic gaming behavior of the adversary. © 2011 Society for Risk Analysis.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Broc, D.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    Directory of Open Access Journals (Sweden)

    R. Cimino

    1999-06-01

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

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

    Science.gov (United States)

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

    2015-01-01

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

  1. Radiation levels on empty cylinders containing heel material

    Energy Technology Data Exchange (ETDEWEB)

    Shockley, C.W. [Martin Marietta Energy Systems, Inc., Paducah, KY (United States)

    1991-12-31

    Empty UF{sub 6} cylinders containing heel material were found to emit radiation levels in excess of 200 mr/hr, the maximum amount stated in ORO-651. The radiation levels were as high as 335 mr/hr for thick wall (48X and 48Y) cylinders and 1050 mr/hr for thin wall (48G and 48H) cylinders. The high readings were found only on the bottom of the cylinders. These radiation levels exceeded the maximum levels established in DOT 49 CFR, Part 173.441 for shipment of cylinders. Holding periods of four weeks for thick-wall cylinders and ten weeks for thin-wall cylinders were established to allow the radiation levels to decay prior to shipment.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  3. Analysis of wallboard containing a phase change material

    Science.gov (United States)

    Tomlinson, J. J.; Heberle, D. P.

    Phase change materials (PCMs) used on the interior of buildings hold the promise for improved thermal performance by reducing the energy requirements for space conditioning and by improving thermal comfort by reducing temperature swings inside the building. Efforts are underway to develop a gypsum wallboard containing a hydrocarbon PCM. With a phase change temperature in the room temperature range, the PCM wallboard adds substantially to the thermal mass of the building while serving the same architectural function as conventional wallboard. To determine the thermal and economic performance of this PCM wallboard, the Transient Systems Simulation Program (TRNSYS) was modified to accommodate walls that are covered with PCM plasterboard, and to apportion the direct beam solar radiation to interior surfaces of a building. The modified code was used to simulate the performance of conventional and direct-gain passive solar residential-sized buildings with and without PCM wallboard. Space heating energy savings were determined as a function of PCM wallboard characteristics. Thermal comfort improvements in buildings containing the PCM were qualified in terms of energy savings. The report concludes with a present worth economic analysis of these energy savings and arrives at system costs and economic payback based on current costs of PCMs under study for the wallboard application.

  4. Evaluation of calculational and material models for concrete containment structures

    International Nuclear Information System (INIS)

    Dunham, R.S.; Rashid, Y.R.; Yuan, K.A.

    1984-01-01

    A computer code utilizing an appropriate finite element, material and constitutive model has been under development as a part of a comprehensive effort by the Electric Power Research Institute (EPRI) to develop and validate a realistic methodology for the ultimate load analysis of concrete containment structures. A preliminary evaluation of the reinforced and prestressed concrete modeling capabilities recently implemented in the ABAQUS-EPGEN code has been completed. This effort focuses on using a state-of-the-art calculational model to predict the behavior of large-scale reinforced concrete slabs tested under uniaxial and biaxial tension to simulate the wall of a typical concrete containment structure under internal pressure. This paper gives comparisons between calculations and experimental measurements for a uniaxially-loaded specimen. The calculated strains compare well with the measured strains in the reinforcing steel; however, the calculations gave diffused cracking patterns that do not agree with the discrete cracking observed in the experiments. Recommendations for improvement of the calculational models are given. (orig.)

  5. The interaction of iodine with organic material in containment

    International Nuclear Information System (INIS)

    Wren, J.C.; Ball, J.M.; Glown, G.A.; Portmann, R.; Sanipelli, G.G.

    1996-01-01

    Organic impurities in containment water, originating from various painted structural surfaces and organic containment materials, could have a significant impact on iodine volatility following an accident. A research program at the Whiteshell Laboratories of AECL has been designed to determine the impact of organic impurities on iodine volatility under accident conditions. The program consists of experimental, literature and modelling studies on the radiolysis or organic compounds in the aqueous phase, thermal and radiolytic formation and decomposition of organic iodides, dissolution of organic solvents from various painted surfaces into the aqueous phase, and iodine deposition on painted surfaces. The experimental studies consist of bench-scale 'separate effects' tests as well as intermediate-scale 'integrated effects' in the Radioiodine Test facility. The studies have shown that organic impurities will be found in containment water, arising from the dissolution of organic compounds from various surface paints and that these compounds can potentially have a significant impact on iodine volatility following an accident. The main impact of surface paints will occur through aqueous-phase reactions of the organic compounds that they release to the aqueous phase. Under the radiation conditions expected during an accident, these compounds will react to reduce the pH and dissolved oxygen concentration, consequently increasing the formation of I 2 from I - that is present in the sump. It appears that the rates of these processes may be controlled by the dissolution kinetics of the organic compounds from the surface coatings. Moreover, the organic compounds may also react thermally and radiolytically with I 2 to form organic iodides in the aqueous phase. Our studies have shown that the formation of organic iodides from soluble organics such as ketones, alcohols and phenols may have more impact on the total iodine volatility than the formation of CH 3 I. (author) 13 figs., 2

  6. The interaction of iodine with organic material in containment

    Energy Technology Data Exchange (ETDEWEB)

    Wren, J C; Ball, J M; Glown, G A; Portmann, R; Sanipelli, G G [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-01

    Organic impurities in containment water, originating from various painted structural surfaces and organic containment materials, could have a significant impact on iodine volatility following an accident. A research program has been designed to determine the impact of organic impurities on iodine volatility under accident conditions. The program consists of experimental, literature and modelling studies on the radiolysis or organic compounds in the aqueous phase, thermal and radiolytic formation and decomposition of organic iodides, dissolution of organic solvents from various painted surfaces into the aqueous phase, and iodine deposition on painted surfaces. The experimental studies consist of bench-scale `separate effects` tests as well as intermediate-scale `integrated effects` in the Radioiodine Test facility. The studies have shown that organic impurities will be found in containment water, arising from the dissolution of organic compounds from various surface paints and that these compounds can potentially have a significant impact on iodine volatility following an accident. The main impact of surface paints will occur through aqueous-phase reactions of the organic compounds that they release to the aqueous phase. Under the radiation conditions expected during an accident, these compounds will react to reduce the pH and dissolved oxygen concentration, consequently increasing the formation of I{sub 2} from I{sup -} that is present in the sump. It appears that the rates of these processes may be controlled by the dissolution kinetics of the organic compounds from the surface coatings. Moreover, the organic compounds may also react thermally and radiolytically with I{sub 2} to form organic iodides in the aqueous phase. Our studies have shown that the formation of organic iodides from soluble organics such as ketones, alcohols and phenols may have more impact on the total iodine volatility than the formation of CH{sub 3}I. (author) 13 figs., 2 tabs., 19 refs.

  7. Fabricating Composite-Material Structures Containing SMA Ribbons

    Science.gov (United States)

    Turner, Travis L.; Cano, Roberto J.; Lach, Cynthia L.

    2003-01-01

    An improved method of designing and fabricating laminated composite-material (matrix/fiber) structures containing embedded shape-memory-alloy (SMA) actuators has been devised. Structures made by this method have repeatable, predictable properties, and fabrication processes can readily be automated. Such structures, denoted as shape-memory-alloy hybrid composite (SMAHC) structures, have been investigated for their potential to satisfy requirements to control the shapes or thermoelastic responses of themselves or of other structures into which they might be incorporated, or to control noise and vibrations. Much of the prior work on SMAHC structures has involved the use SMA wires embedded within matrices or within sleeves through parent structures. The disadvantages of using SMA wires as the embedded actuators include (1) complexity of fabrication procedures because of the relatively large numbers of actuators usually needed; (2) sensitivity to actuator/ matrix interface flaws because voids can be of significant size, relative to wires; (3) relatively high rates of breakage of actuators during curing of matrix materials because of sensitivity to stress concentrations at mechanical restraints; and (4) difficulty of achieving desirable overall volume fractions of SMA wires when trying to optimize the integration of the wires by placing them in selected layers only.

  8. Inelastic analysis acceptance criteria for radioactive material transportation containers

    International Nuclear Information System (INIS)

    Ammerman, D.J.; Ludwigsen, J.S.

    1993-01-01

    The design criteria currently used in the design of radioactive material (RAM) transportation containers are taken from the ASME Boiler and Pressure Vessel Code (ASME, 1992). These load-based criteria are ideally suited for pressure vessels where the loading is quasistatic and all stresses are in equilibrium with externally applied loads. For impact events, the use of load-based criteria is less supportable. Impact events tend to be energy controlled, and thus, energy-based acceptance criteria would appear to be more appropriate. Determination of an ideal design criteria depends on what behavior is desired. Currently there is not a design criteria for inelastic analysis for RAM nation packages that is accepted by the regulatory agencies. This lack of acceptance criteria is one of the major factors in limiting the use of inelastic analysis. In this paper inelastic analysis acceptance criteria based on stress and strain-energy density will be compared for two stainless steel test units subjected to impacts onto an unyielding target. Two different material models are considered for the inelastic analysis, a bilinear fit of the stress-strain curve and a power law hardening model that very closely follows the stress-strain curve. It is the purpose of this paper to stimulate discussion and research into the area of strain-energy density based inelastic analysis acceptance criteria

  9. Detection of nuclear material by photon activation inside cargo containers

    Science.gov (United States)

    Gmar, Mehdi; Berthoumieux, Eric; Boyer, Sébastien; Carrel, Frédérick; Doré, Diane; Giacri, Marie-Laure; Lainé, Frédéric; Poumarède, Bénédicte; Ridikas, Danas; Van Lauwe, Aymeric

    2006-05-01

    Photons with energies above 6 MeV can be used to detect small amounts of nuclear material inside large cargo containers. The method consists in using an intense beam of high-energy photons (bremsstrahlung radiation) in order to induce reactions of photofission on actinides. The measurement of delayed neutrons and delayed gammas emitted by fission products brings specific information on localization and quantification of the nuclear material. A simultaneous measurement of both of these delayed signals can overcome some important limitations due to matrix effects like heavy shielding and/or the presence of light elements as hydrogen. We have a long experience in the field of nuclear waste package characterization by photon interrogation and we have demonstrated that presently the detection limit can be less than one gram of actinide per ton of package. Recently we tried to extend our knowledge to assess the performance of this method for the detection of special nuclear materials in sea and air freights. This paper presents our first results based on experimental measurements carried out in the SAPHIR facility, which houses a linear electron accelerator with the energy range from 15 MeV to 30 MeV. Our experiments were also modeled using the full scale Monte Carlo techniques. In addition, and in a more general frame, due to the lack of consistent data on photonuclear reactions, we have been working on the development of a new photonuclear activation file (PAF), which includes cross sections for more than 600 isotopes including photofission fragment distributions and delayed neutron tables for actinides. Therefore, this work includes also some experimental results obtained at the ELSA electron accelerator, which is more adapted for precise basic nuclear data measurements.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-23

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Science.gov (United States)

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

    2017-06-29

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

  14. Fuel salt and container material studies for MOSART transforming system

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V.; Feynberg, O.; Merzlyakov, A.; Surenkov, A.; Zagnitko, A. [National Research Center, Kurchatov Institute, Moscow (Russian Federation); Afonichkin, V.; Bovet, A.; Khokhlov, V. [Institute of High Temperature Electrochemisty, Ekaterinburg (Russian Federation); Subbotin, V.; Gordeev, M.; Panov, A.; Toropov, A. [Institute of Technical Physics, Snezhinsk (Russian Federation)

    2013-07-01

    A study is under progress to examine the feasibility of single stream Molten Salt Actinide Recycling and Transmuting system without and with Th support (MOSART) fuelled with different compositions of actinide tri-fluorides (AnF{sub 3}) from used LWR fuel. New fast-spectrum design options with homogeneous core and fuel salts with high enough solubility for AnF{sub 3} are being examined because of new goals. The flexibility of single fluid MOSART concept with Th support is underlined, particularly, possibility of its operation in self-sustainable mode (Conversion Ratio: CR=1) using different loadings and make up. The paper summarizes the most current status of fuel salt and container material data for the MOSART concept received within ISTC-3749 and ROSATOM-MARS projects. Key physical and chemical properties of various fluoride fuel salts are reported. The issues like salt purification, the electroreduction of U(IV) to U(III) in LiF-ThF{sub 4} and the electroreduction of Yb(III) to Yb(II) in LiF-NaF are detailed.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  17. Combustion of crude oil sludge containing naturally occurring radioactive material

    International Nuclear Information System (INIS)

    Mohamad Puad Abu; Muhd Noor Muhd Yunus; Shamsuddin, A.H.; Sopian, K.

    2000-01-01

    The characteristics of crude oil sludge fi-om the crude oil terminal are very unique because it contains both heavy metals and also Naturally Occurring Radioactive Material (NORM). As a result, the Department of Environmental (DOE) and the Atomic Energy Licensing Board (AELB) considered it as Scheduled Wastes and Low Level Radioactive Waste (LLRW) respectively. As a Scheduled Wastes, there is no problem in dealing with the disposal of it since there already exist a National Center in Bukit Nanas to deal with this type of waste. However, the Center could not manage this waste due to the presence of NORM by which the policy regarding the disposal of this kind of waste has not been well established. This situation is unclear to certain parties, especially with respect to the relevant authorities having final jurisdiction over the issue as well as the best practical method of disposal of this kind of waste. Existing methods of treatment viewed both from literature and current practice include that of land farming, storing in plastic drum, re-injection into abandoned oil well, recovery, etc., found some problems. Due to its organic nature, very low level in radioactivity and the existence of a Scheduled Waste incineration facility in Bukit Nanas, there is a potential to treat this sludge by using thermal treatment technology. However, prior to having this suggestion to be put into practice, there are issues that need to be addressed. This paper attempts to discuss the potentials and the related issues of combusting crude oil sludge based on existing experimental data as well as mathematical modeling

  18. Development of Separation Materials Containing Palladium for Hydrogen Isotopes Separation

    International Nuclear Information System (INIS)

    Deng Xiaojun; Luo Deli; Qian Xiaojing

    2010-01-01

    Displacement chromatography (DC) is a ascendant technique for hydrogen isotopes separation. The performance of separation materials is a key factor to determine the separation effect of DC. At present,kinds of materials are researched, including palladium materials and non-palladium materials. It is hardly replaceable because of its excellent separation performance, although palladium is expensive. The theory of hydrogen isotopes separation using DC was introduced at a brief manner, while several palladium separation materials were expatiated in detail(Pd/K, Pd-Al 2 O 3 , Pd-Pt alloy). Development direction of separation materials for DC was forecasted elementarily. (authors)

  19. H1259 Container Foams: Performance Data on Aged Materials

    International Nuclear Information System (INIS)

    Linda Domeier

    2002-01-01

    Samples of the three cushioning foams used in the H1259 weapon storage container were obtained in 1997, 1998, 2000 and 2001 and tested for density, compression set and compressive strength using the same procedures specified for acceptance testing. Foams from six containers, all about 30 years old and located at Pantex, were evaluated. The bottom cushioning foam is a General Plastics polyurethane foam and the two side pads are rebonded polyurethane foams. All the tests were carried out at room temperature. When compared to the original acceptance requirements the foams were generally in-spec for density and compressive strength at 10% strain and were generally out-of-spec for compression set and compressive strength at 50% strain. Significant variability was noted in the performance of each foam sample and even more in the container-to-container foam performance. The container-to-container variability remains the major unknown in predicting the long-term suitability of these containers for continued use. The performance of the critical bottom cushion foams was generally more uniform and closer to the specified performance than that of the rebonded foams. It was judged that all the foams were adequate for continued use as storage container foams (not shipping) under controlled conditions to mitigate temperature extremes or high impact. This archived information is important in evaluations of the continued suitability for weapon storage use of the H1259 containers and other containers using the same foam cushions

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

    Science.gov (United States)

    Stewart, David A.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Groot, Joost de

    2012-01-01

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

  3. 19 CFR 10.601 - Retail packaging materials and containers.

    Science.gov (United States)

    2010-04-01

    ...-Central America-United States Free Trade Agreement Rules of Origin § 10.601 Retail packaging materials and...), HTSUS. (b) Effect on regional value content calculation. If the good is subject to a regional value... originating or non-originating materials, as the case may be, in calculating the regional value content of the...

  4. Development of neutron shielding concrete containing iron content materials

    Science.gov (United States)

    Sariyer, Demet; Küçer, Rahmi

    2018-02-01

    Concrete is one of the most important construction materials which widely used as a neutron shielding. Neutron shield is obtained of interaction with matter depends on neutron energy and the density of the shielding material. Shielding properties of concrete could be improved by changing its composition and density. High density materials such as iron or high atomic number elements are added to concrete to increase the radiation resistance property. In this study, shielding properties of concrete were investigated by adding iron, FeB, Fe2B, stainless - steel at different ratios into concrete. Neutron dose distributions and shield design was obtained by using FLUKA Monte Carlo code. The determined shield thicknesses vary depending on the densities of the mixture formed by the additional material and ratio. It is seen that a combination of iron rich materials is enhanced the neutron shielding of capabilities of concrete. Also, the thicknesses of shield are reduced.

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

    International Nuclear Information System (INIS)

    Cyster, L.A.

    2001-10-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. Characterization of Concrete Mixes Containing Phase Change Materials

    Science.gov (United States)

    Paksoy, H.; Kardas, G.; Konuklu, Y.; Cellat, K.; Tezcan, F.

    2017-10-01

    Phase change materials (PCM) can be used in passive building applications to achieve near zero energy building goals. For this purpose PCM can be added in building structures and materials in different forms. Direct incorporation, form stabilization and microencapsulation are different forms used for PCM integration in building materials. In addition to thermal properties of PCM itself, there are several other criteria that need to be fulfilled for the PCM enhanced building materials. Mechanical properties, corrosive effects, morphology and thermal buffering have to be determined for reliable and long-term applications in buildings. This paper aims to give an overview of characterization methods used to determine these properties in PCM added fresh concrete mixes. Thermal, compressive strength, corrosion, and microscopic test results for concrete mixes with PCM are discussed.

  8. Experimental analysis of plastic materials containing radionuclides for decontamination viability

    International Nuclear Information System (INIS)

    Tazaki, Kazue; Nakano, Mikio; Takehara, Teruaki; Ishigaki, Yasuhito; Nakagawa, Hideaki

    2015-01-01

    After the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident on 11 March, 2011, the high radioactive dosage was found in polluted water for agriculture use at Baba, Haramachi, Minami-Soma, Fukushima Prefecture, Japan. Field experiment for decontamination of water had been studied by using commercial plastic materials. The agricultural water comes from Tetsuzan dam is full of radioactive-contaminated water. Experimental analysis showed that the plastic materials can take up radioactive elements for several months soaked in the polluted agricultural water. The quantitative analyses using X-ray fluorescence analysis, Ge semiconductor and scanning electron microscopy equipped with energy dispersive X-ray spectrometer (SEM-EDS), revealed the detection of the radionuclides on the plastic materials with diatom and clays. The results suggest the adsorption of radionuclides on the surface of plastic materials due to FDNPP accident. The plastic materials associated with clays and diatoms could be stronger carriers of radionuclides in the polluted water. Adherence of diatoms to the plastic fiber in the water for 7 months suggested that some plastic materials were taking up heavy metals (Zn, Ba, Pb, Sb) with radioactive elements (Cs etc.). Mechanisms by which radioactive pollutants and microorganisms are adsorbed onto and desorbed from clays at aqueous interface can be understood by combining chemical analysis with electron microscopy observation. (author)

  9. New CSA guideline for the exemption or clearance from regulatory control of materials that contain, or potentially contain, nuclear substances

    International Nuclear Information System (INIS)

    Rhodes, M.; Kwong, A.

    2011-01-01

    The Canadian Standards Association (CSA) guideline N292.5, Guideline for the exemption or clearance from regulatory control of materials that contain, or potentially contain, nuclear substances, was recently developed to address a need for guidance on approaches for clearance of materials from facilities licensed by the Canadian Nuclear Safety Commission (CNSC) consistent with Canadian and international recommendations. This guideline is also applicable to determining if an activity associated with materials that contain nuclear substances is exempt from requiring a CNSC licence. The guideline summarizes the regulatory requirements associated with the exemption and clearance of materials and provides a graded approach to designing a survey based on the risk of residual contamination being present. (author)

  10. Polymers Containing 1, 3, 4-Oxadiazole Rings for Advanced Materials

    Directory of Open Access Journals (Sweden)

    Mariana-Dana Damaceanu

    2011-10-01

    Full Text Available This paper presents the synthesis, properties and potential applications of new polymers containing 1, 3, 4-oxadiazole rings, tacking into account the requirements of the modern technologies. Two classes of polymers containing oxadiazole rings were approached: polyamides and polyimides. All the polymers were characterized with respect to the identification of their chemical structure, solubility, molecular weights, film forming ability, thermal, dielectric and optical properties, and the behaviour of polyoxadiazole films upon irradiation with pulsed KrF laser. All the properties were discussed in correlation with their chemical structure and compared with those of related polymers.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-12-15

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

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

    International Nuclear Information System (INIS)

    Olsson, Siv; Karnland, Ola

    2009-12-01

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

  13. Guidelines for Assessment and Abatement of Asbestos-Containing Materials in Buildings.

    Science.gov (United States)

    Pielert, James H.; Mathey, Robert G.

    This report presents guidelines, based on available information, for the assessment and abatement of asbestos-containing materials in buildings. Section 1 provides background information on the history and use of asbestos-containing products in buildings, the characteristics of asbestos fibers, products and materials containing asbestos, and…

  14. Processing fissile material mixtures containing zirconium and/or carbon

    Science.gov (United States)

    Johnson, Michael Ernest; Maloney, Martin David

    2013-07-02

    A method of processing spent TRIZO-coated nuclear fuel may include adding fluoride to complex zirconium present in a dissolved TRIZO-coated fuel. Complexing the zirconium with fluoride may reduce or eliminate the potential for zirconium to interfere with the extraction of uranium and/or transuranics from fission materials in the spent nuclear fuel.

  15. Method of processing liquid wastes containing radioactive materials

    International Nuclear Information System (INIS)

    Matsumoto, Kaname; Shirai, Takamori; Nemoto, Kuniyoshi; Yoshikawa, Jun; Matsuda, Takeshi.

    1983-01-01

    Purpose: To reduce the number of solidification products by removing, particularly, Co-60 that is difficult to remove in a radioactive liquid wastes containing a water-soluble chelating agent, by adsorbing Co-60 to a specific chelating agent. Method: Liquid wastes containing radioactive cobalt and water-soluble chelating agent are passed through the layer of less water-soluble chelating agent that forms a complex compound with cobalt in an acidic pH region. Thus, the chelating compound of radioactive cobalt (particularly Co-60) is eliminated by adsorbing the same on a specific chelating agent layer. The chelating agent having Co-60 adsorbed thereon is discarded as it is through the cement- or asphalt-solidification process, whereby the number of solidification products to be generated can significantly be suppressed. (Moriyama, K.)

  16. Bismuth silicate glass containing heavy metal oxide as a promising radiation shielding material

    Science.gov (United States)

    Elalaily, Nagia A.; Abou-Hussien, Eman M.; Saad, Ebtisam A.

    2016-12-01

    Optical and FTIR spectroscopic measurements and electron paramagnetic resonance (EPR) properties have been utilized to investigate and characterize the given compositions of binary bismuth silicate glasses. In this work, it is aimed to study the possibility of using the prepared bismuth silicate glasses as a good shielding material for γ-rays in which adding bismuth oxide to silicate glasses causes distinguish increase in its density by an order of magnitude ranging from one to two more than mono divalent oxides. The good thermal stability and high density of the bismuth-based silicate glass encourage many studies to be undertaken to understand its radiation shielding efficiency. For this purpose a glass containing 20% bismuth oxide and 80% SiO2 was prepared using the melting-annealing technique. In addition the effects of adding some alkali heavy metal oxides to this glass, such as PbO, BaO or SrO, were also studied. EPR measurements show that the prepared glasses have good stability when exposed to γ-irradiation. The changes in the FTIR spectra due to the presence of metal oxides were referred to the different housing positions and physical properties of the respective divalent Sr2+, Ba2+ and Pb2+ ions. Calculations of optical band gap energies were presented for some selected glasses from the UV data to support the probability of using these glasses as a gamma radiation shielding material. The results showed stability of both optical and magnetic spectra of the studied glasses toward gamma irradiation, which validates their irradiation shielding behavior and suitability as the radiation shielding candidate materials.

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

    Science.gov (United States)

    1994-12-29

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

  18. Study of improving the thermal response of a construction material containing a phase change material

    Science.gov (United States)

    Laaouatni, A.; Martaj, N.; Bennacer, R.; Elomari, M.; El Ganaoui, M.

    2016-09-01

    The use of phase change materials (PCMs) for improving the thermal comfort in buildings has become an attractive application. This solution contributes to increasing the thermal inertia of the building envelope and reducing power consumption. A building element filled with a PCM and equipped with ventilation tubes is proposed, both for increasing inertia and contributing to refreshing building envelope. A numerical simulation is conducted by the finite element method in COMSOL Multiphysics, which aims to test the thermal behaviour of the developed solution. An experimental study is carried out on a concrete block containing a PCM with ventilation tubes. The objective is to see the effect of PCM coupled with ventilation on increasing the inertia of the block. The results show the ability of this new solution to ensure an important thermal inertia of a building.

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

    Energy Technology Data Exchange (ETDEWEB)

    Boss, D.E.

    1997-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-01

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

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

    Science.gov (United States)

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

    2015-09-01

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

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

    Science.gov (United States)

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

    2018-05-01

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

  3. 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

  4. Comparative study of biogenic and abiotic iron-containing materials

    Energy Technology Data Exchange (ETDEWEB)

    Cherkezova-Zheleva, Z., E-mail: zzhel@ic.bas.bg; Shopska, M., E-mail: shopska@ic.bas.bg; Paneva, D. [Bulgarian Academy of Sciences, Institute of Catalysis (Bulgaria); Kovacheva, D. [Bulgarian Academy of Sciences, Institute of General and Inorganic Chemistry (Bulgaria); Kadinov, G.; Mitov, I. [Bulgarian Academy of Sciences, Institute of Catalysis (Bulgaria)

    2016-12-15

    Series of iron-based biogenic materials prepared by cultivation of Leptothrix group of bacteria in different feeding media (Sphaerotilus-Leptothrix group of bacteria isolation medium, Adler, Lieske and silicon-iron-glucose-peptone) were studied. Control samples were obtained in the same conditions and procedures but the nutrition media were not infected with bacteria, i.e. they were sterile. Room and low temperature Mössbauer spectroscopy, powder X-ray diffraction (XRD), and infrared spectroscopy (IRS) were used to reveal the composition and physicochemical properties of biomass and respective control samples. Comparative analysis showed differences in their composition and dispersity of present phases. Sample composition included different ratio of nanodimensional iron oxyhydroxide and oxide phases. Relaxation phenomena such as superparamagnetism or collective magnetic excitation behaviour were registered for some of them. The experimental data showed that the biogenic materials were enriched in oxyhydroxides of high dispersion. Catalytic behaviour of a selected biomass and abiotic material were studied in the reaction of CO oxidation. In situ diffuse-reflectance (DR) IRS was used to monitor the phase transformations in the biomass and CO conversion.

  5. 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

  6. Electrochromic device containing metal oxide nanoparticles and ultraviolet blocking material

    Science.gov (United States)

    Garcia, Guillermo; Koo, Bonil; Gregoratto, Ivano; Basu, Sourav; Rosen, Evelyn; Holt, Jason; Thomsen, Scott

    2017-10-17

    An electrochromic device includes a nanostructured transition metal oxide bronze layer that includes one or more transition metal oxide and one or more dopant. The electrochromic device also includes nanoparticles containing one or more transparent conducting oxide (TCO), a solid state electrolyte, a counter electrode, and at least one protective layer to prevent degradation of the one or more nanostructured transition metal oxide bronze. The nanostructured transition metal oxide bronze selectively modulates transmittance of near-infrared (NIR) and visible radiation as a function of an applied voltage to the device.

  7. Method of forming capsules containing a precise amount of material

    Science.gov (United States)

    Grossman, M.W.; George, W.A.; Maya, J.

    1986-06-24

    A method of forming a sealed capsule containing a submilligram quantity of mercury or the like, the capsule being constructed from a hollow glass tube, by placing a globule or droplet of the mercury in the tube. The tube is then evacuated and sealed and is subsequently heated so as to vaporize the mercury and fill the tube therewith. The tube is then separated into separate sealed capsules by heating spaced locations along the tube with a coiled heating wire means to cause collapse spaced locations there along and thus enable separation of the tube into said capsules. 7 figs.

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

    International Nuclear Information System (INIS)

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

    1978-06-01

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

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

    Science.gov (United States)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-28

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

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

    Directory of Open Access Journals (Sweden)

    Herry Poernomo

    2010-12-01

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

  12. Remote automated material handling of radioactive waste containers

    International Nuclear Information System (INIS)

    Greager, T.M.

    1994-09-01

    To enhance personnel safety, improve productivity, and reduce costs, the design team incorporated a remote, automated stacker/retriever, automatic inspection, and automated guidance vehicle for material handling at the Enhanced Radioactive and Mixed Waste Storage Facility - Phase V (Phase V Storage Facility) on the Hanford Site in south-central Washington State. The Phase V Storage Facility, scheduled to begin operation in mid-1997, is the first low-cost facility of its kind to use this technology for handling drums. Since 1970, the Hanford Site's suspect transuranic (TRU) wastes and, more recently, mixed wastes (both low-level and TRU) have been accumulating in storage awaiting treatment and disposal. Currently, the Hanford Site is only capable of onsite disposal of radioactive low-level waste (LLW). Nonradioactive hazardous wastes must be shipped off site for treatment. The Waste Receiving and Processing (WRAP) facilities will provide the primary treatment capability for solid-waste storage at the Hanford Site. The Phase V Storage Facility, which accommodates 27,000 drum equivalents of contact-handled waste, will provide the following critical functions for the efficient operation of the WRAP facilities: (1) Shipping/Receiving; (2) Head Space Gas Sampling; (3) Inventory Control; (4) Storage; (5) Automated/Manual Material Handling

  13. Effects of materials containing antimicrobial compounds on food hygiene.

    Science.gov (United States)

    Møretrø, Trond; Langsrud, Solveig

    2011-07-01

    Surfaces with microorganisms may transfer unwanted microorganisms to food through cross-contamination during processing and preparation. A high hygienic status of surfaces that come in contact with food is important in order to reduce the risk of cross-contamination. During the last decade, products containing antimicrobial compounds, such as cutting boards, knives, countertops, kitchen utensils, refrigerators, and conveyor belts, have been introduced to the market, claiming hygienic effects. Such products are often referred to as "treated articles." Here we review various aspects related to treated articles intended for use during preparation and processing of food. Regulatory issues and methods to assess antibacterial effects are covered. Different concepts for treated articles as well as their antibacterial activity are reviewed. The effects of products with antimicrobials on food hygiene and safety are discussed. Copyright ©, International Association for Food Protection

  14. On material and energy sources of formation of fuel-containing materials during Chernobyl NPP UNIT 4 accident

    Directory of Open Access Journals (Sweden)

    O. V. Mikhailov

    2016-12-01

    Full Text Available Results of detailed analysis of material substance of lava-like fuel-containing materials sources (FCM and clusters with high uranium concentration were presented. Material and energy balance are aggregated in a process model for optimal composition of sacrificial materials and FCM. Quantitative estimate is given for spent nuclear fuel’ afterheat in a number of other heat energy sources in reactor vault. Conclusion was made that upon condition of 50 % heat loss, remained amount of “useful” heat would be sufficient for proceeding of blast furnace version of fuel-containing materials.

  15. Packaging material and flexible medical tubing containing thermally exfoliated graphite oxide

    Science.gov (United States)

    Prud'homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor)

    2011-01-01

    A packaging material or flexible medical tubing containing a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 m.sup.2/g to 2600 m.sup.2/g.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    Directory of Open Access Journals (Sweden)

    T. Zaman

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

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

    International Nuclear Information System (INIS)

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

    2006-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-03-15

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  7. Detection of tiny amounts of fissile materials in large-sized containers with radioactive waste

    Science.gov (United States)

    Batyaev, V. F.; Skliarov, S. V.

    2018-01-01

    The paper is devoted to non-destructive control of tiny amounts of fissile materials in large-sized containers filled with radioactive waste (RAW). The aim of this work is to model an active neutron interrogation facility for detection of fissile ma-terials inside NZK type containers with RAW and determine the minimal detectable mass of U-235 as a function of various param-eters: matrix type, nonuniformity of container filling, neutron gen-erator parameters (flux, pulse frequency, pulse duration), meas-urement time. As a result the dependence of minimal detectable mass on fissile materials location inside container is shown. Nonu-niformity of the thermal neutron flux inside a container is the main reason of the space-heterogeneity of minimal detectable mass in-side a large-sized container. Our experiments with tiny amounts of uranium-235 (<1 g) confirm the detection of fissile materials in NZK containers by using active neutron interrogation technique.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  9. Technique of stowing packages containing radioactive materials during maritime transportation

    International Nuclear Information System (INIS)

    Ringot, G.; Chevalier, G.; Tomachevsky, E.; Draulans, J.; Lafontaine, I.

    1989-01-01

    The Mont Louis accident (August 25, 1984 - North Sea), in which uraniumhexafluoride packages were involved, alarmed a large number of European competent authorities, including the Commission of European Communities. The latter sponsored in 1986-1987 a bibliographic data collection to obtain a first view on the problem. (C.E.C contracts n degree 86-B-7015-11-004-17 and 86-B-7015-11-005-17). The collected data supply the necessary basis for further work, aiming to increase the safety of transporting radioactive material by ship. The study collected the different deceleration values, used by the transport companies and defined the accident conditions to be considered. This work can serve as a basis for later research to end with the proposal of a code of good practice for stowing. The research-work has been carried out jointly by C.E.A.-France, I.P.S.N. at Fontenay-aux-Roses and by Transnubel S.A. Brussels Belgium. The preliminary research included two main tasks: a statistical analysis, a bibliographic study of ship accidents

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-08-01

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

  12. Impact of container material on the development of Aedes aegypti larvae at different temperatures.

    Science.gov (United States)

    Kumar, Gaurav; Singh, R K; Pande, Veena; Dhiman, R C

    2016-01-01

    Aedes aegypti, the primary vector of dengue generally breeds in intradomestic and peridomestic containers made up of different materials, i.e. plastic, iron, rubber, earthen material etc. The material of container is likely to affect the temperature of water in container with variation in environmental temperature. The present study was aimed to determine the effect of different container materials on larval development of Ae. aegypti at different temperatures. Newly hatched I instar larvae (2-4 h old) were used in the study and experiments were conducted using three different containers made up of plastic, iron and earthen material. Three replicates for each type of container at 22, 26, 30, 34, 38, 40, and 42°C were placed in environmental chamber for the development of larvae. At temperatures >22°C, 50% pupation was completed in earthen pot within 4.3±0.6 to 6.3±0.6 days followed by plastic containers (5±0 to 8±0 days) and iron containers (6±0 to 9±0 days). Developmental time for 50% pupation in the three containers differed significantly (p containers (p containers resulted in significant variations in the developmental period of larvae. More than 35°C temperature of water was found inimical for pupal development. The results revealed the variation in temperature of water in different types of containers depending on the material of container, affecting duration of larval development. As the larval development was faster in earthen pot as compared to plastic and iron containers, community should be discouraged for storing the water in earthen pots. However, in view of containers of different materials used by the community in different temperature zones in the country, further studies are required for devising area-specific preventive measures for Aedes breeding.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    Czech Academy of Sciences Publication Activity Database

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    Directory of Open Access Journals (Sweden)

    Donald C. Aduba

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-30

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  1. 12 CFR 503.2 - Exemptions of records containing investigatory material compiled for law enforcement purposes.

    Science.gov (United States)

    2010-01-01

    ... 12 Banks and Banking 5 2010-01-01 2010-01-01 false Exemptions of records containing investigatory material compiled for law enforcement purposes. 503.2 Section 503.2 Banks and Banking OFFICE OF THRIFT SUPERVISION, DEPARTMENT OF THE TREASURY PRIVACY ACT § 503.2 Exemptions of records containing investigatory material compiled for law enforcement...

  2. Researches of real observation geometry in monitoring fuel-containing materials' subcriticality

    International Nuclear Information System (INIS)

    Vysotskij, E.D.; Shevchenko, V.G.; Shevchenko, M.V.

    2002-01-01

    The effectiveness of fuel-containing materials monitoring is discussed in the part related to the feasibilities of researches and realization of optimal geometry (detectors - source) of survey of neutron activity dynamics in nuclearly hazardous areas with clusters of fuel-containing materials concentrated in the premises 305/2

  3. Risk ranking of LANL nuclear material storage containers for repackaging prioritization.

    Science.gov (United States)

    Smith, Paul H; Jordan, Hans; Hoffman, Jenifer A; Eller, P Gary; Balkey, Simon

    2007-05-01

    Safe handling and storage of nuclear material at U.S. Department of Energy facilities relies on the use of robust containers to prevent container breaches and subsequent worker contamination and uptake. The U.S. Department of Energy has no uniform requirements for packaging and storage of nuclear materials other than those declared excess and packaged to DOE-STD-3013-2000. This report describes a methodology for prioritizing a large inventory of nuclear material containers so that the highest risk containers are repackaged first. The methodology utilizes expert judgment to assign respirable fractions and reactivity factors to accountable levels of nuclear material at Los Alamos National Laboratory. A relative risk factor is assigned to each nuclear material container based on a calculated dose to a worker due to a failed container barrier and a calculated probability of container failure based on material reactivity and container age. This risk-based methodology is being applied at LANL to repackage the highest risk materials first and, thus, accelerate the reduction of risk to nuclear material handlers.

  4. Comparison of solute-binding properties of plastic materials used as pharmaceutical product containers.

    Science.gov (United States)

    Jenke, Dennis; Couch, Tom; Gillum, Amy

    2010-01-01

    Material/water equilibrium binding constants (E(b)) were determined for 11 organic solutes and 2 plastic materials commonly used in pharmaceutical product containers (plasticized polyvinyl chloride and polyolefin). In general, solute binding by the plasticized polyvinyl chloride material was greater, by nearly an order of magnitude, than the binding by the polyolefin (on an equal weight basis). The utilization of the binding constants to facilitate container compatibility assessments (e.g., drug loss by container binding) for drug-containing products is discussed.

  5. Study of mechanoactivation of tungsten-molybdenum containing raw material in gas-jet mill

    International Nuclear Information System (INIS)

    Agnokov, T.Sh.; Gorobets, L.Zh.; Martynenko, V.P.; Fedorov, Yu.P.; Krakhmaleva, M.T.; Sokolova, L.A.

    1988-01-01

    Investigation is aimed at intensifying autoclave-soda leaching of tungsten-molybdenum-containing raw material. Connection of reactivity and physicochemical properties of crushed tungsten-molybdenum-containing products under different gas-jet crushing parameters is investigated. Optimal technological indices of hydrometallurgical reprocessing of tungsten-molybdenum-containing raw materials and products processed by gas-jet technique are given. The results obtained point out to perspectiveness of applying gas-jet technique of thermomechanical processing for intensifying and increasing the quality of tungsten- and molybdenum-containing raw materials and products of hydrometallurgical production

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

    Science.gov (United States)

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

    2016-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

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

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

    Science.gov (United States)

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

    2009-04-01

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

  9. The compatibility of candidate first wall metallic materials with impure helium

    International Nuclear Information System (INIS)

    Noda, T.; Okada, M.; Watanabe, R.

    1979-01-01

    The compatibilities of SUS 316 stainless steels, Nimonic PE 16, Nb-1% Zr, V-25% Mo, Mo, and TZM with the commercial grade helium (> 99.995%) and the helium containing oxygen of 13 vpm at temperatures from 873 to 1273 K were studied. SUS 316 and PE 16 were internally oxidized above 1100 K. The marked depletion of Cr and Mn in SUS 316 specimens was observed in the commercial grade helium above around 1100 K. Nb-1% Zr and V-25% Mo extremely absorbed oxygen and nitrogen from the helium gases and were deteriorated in the range of test temperatures. Mo and TZM appeared not to be affected by the exposure to the commercial grade helium at temperature up to 1273 K. However, Mo and TZM lost ductility at room temperature after exposure to helium above 1100 and 900 K respectively. (orig.)

  10. Recommended welding criteria for use in the fabrication of shipping containers for radioactive materials

    International Nuclear Information System (INIS)

    Monroe, R.E.; Woo, H.H.; Sears, R.G.

    1984-03-01

    Welding and related operations are evaluated to assess the controls required to prevent weld-related failure of shipping containers used for transportation of radioactive materials. The report includes (1) recommended criteria for controlling welding as applied to shipping containers and (2) a discussion of modifications of the recommended industry Codes as applied to shipping containers. 13 references, 2 tables

  11. 19 CFR 10.540 - Packing materials and containers for shipment.

    Science.gov (United States)

    2010-04-01

    ...-Singapore Free Trade Agreement Rules of Origin § 10.540 Packing materials and containers for shipment. (a... the United States. Accordingly, in applying either the build-down or build-up method for determining... shipping container which it purchased from Company B in Singapore. The shipping container is originating...

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-11-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-04-01

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

  16. 49 CFR 176.170 - Transport of Class 1 (explosive) materials in freight containers.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Transport of Class 1 (explosive) materials in... REGULATIONS CARRIAGE BY VESSEL Detailed Requirements for Class 1 (Explosive) Materials Cargo Transport Units and Shipborne Barges § 176.170 Transport of Class 1 (explosive) materials in freight containers. (a...

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

    Science.gov (United States)

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

    2008-05-01

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

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

    Science.gov (United States)

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

    2005-08-01

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

  19. 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

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

    International Nuclear Information System (INIS)

    Agarwala, R.P.

    1989-01-01

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

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

    Science.gov (United States)

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

    1978-01-01

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

  2. 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.)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  4. Double container system for the transport and storage of radioactive materials

    International Nuclear Information System (INIS)

    Popp, F.W.; Pontani, B.; Ernst, E.

    1987-01-01

    The double container system consists of an inner storage container made of steel for the gastight inclusion of the radioactive material to be stored and an outer shielding container which ensures the necessary shielding and mechanical safety in handling and transport. A neutron moderator layer of material containing hydrogen, preferably polyethylene, is present in the annular gap between the outer shielding container and the inner storage container. In order to achieve good shielding with simultaneous very good heat conduction from the inside to the outside, the moderator layer consists of individual polyethylene rings stacked above one another. There is an H profile ring made of heat conducting metal material between each two polyethylene rings. The legs of the H profile ring surround the sides of the two polyethylene rings for fixing it. (orig.) [de

  5. Accumulation of organic compounds leached from plastic materials used in biopharmaceutical process containers.

    Science.gov (United States)

    Jenke, Dennis R; Zietlow, David; Garber, Mary Jo; Sadain, Salma; Reiber, Duane; Terbush, William

    2007-01-01

    Plastic materials are widely used in medical items, such as solution containers, transfusion sets, transfer tubing, and devices. An emerging trend in the biotechnology industry is the utilization of plastic containers to prepare, transport, and store an assortment of solutions including buffers, media, and in-process and finished product. The direct contact of such containers with the product at one or more points in its lifetime raises the possibility that container leachables may accumulate in the finished product. The interaction between several commercially available container materials and numerous model test solutions (representative of buffers and media used in biopharmaceutical applications) was investigated. This paper summarizes the identification of leachables associated with the container materials and documents the levels to which targeted leachables accumulate in the test solutions under defined storage conditions.

  6. Immobilization of INEL low-level radioactive wastes in ceramic containment materials

    International Nuclear Information System (INIS)

    Seymour, W.C.; Kelsey, P.V.

    1978-11-01

    INEL low-level radioactive wastes have an overall chemical composition that lends itself to self-containment in a ceramic-based material. Fewer chemical additives would be needed to process the wastes than to process high-level wastes or use a mixture containment method. The resulting forms of waste material could include a basalt-type glass or glass ceramic and a ceramic-type brick. Expected leach resistance is discussed in relationshp to data found in the literature for these materials and appears encouraging. An overview of possible processing steps for the ceramic materials is presented

  7. Effect of controlled potential on SCC of nuclear waste package container materials

    International Nuclear Information System (INIS)

    Lum, B. Y.; Roy, A. K.; Spragge, M. K.

    1999-01-01

    The slow-strain-rate (SSR) test technique was used to evaluate the susceptibility of Titanium (Ti) Gr-7 (UNS R52400) and Ti Gr-12 (UNS R53400) to stress corrosion cracking (SCC). Ti Gr-7 and Ti Gr-12 are two candidate container materials for the multi-barrier package for nuclear waste. The tests were done in a deaerated 90 C acidic brine (pH ∼ 2.7) containing 5 weight percent (wt%) sodium chloride (NaCl) using a strain rate of 3.3 x 10 -6 sec -1 . Before being tested in the acidic brine, specimens of each alloy were pulled inside the test chamber in the dry condition at ambient temperature. Then while in the test solution, specimens were strained under different cathodic (negative) controlled electrochemical potentials. These controlled potentials were selected based on the corrosion potential measured in the test solution before the specimens were strained. Results indicate that the times to failure (TTF) for Ti Gr-12 were much shorter than those for Ti Gr-7. Furthermore, as the applied potential became more cathodic, Ti Gr-12 showed reduced ductility in terms of percent reduction in area (%RA) and true fracture stress (σ f ). In addition, TTF and percent elongation (%El) reached the minimum values when Ti Gr-12 was tested under an impressed potential of -1162 mV. However, for Ti Gr-7, all these ductility parameters were not significantly influenced by the changes in applied potential. In general, the results of hydrogen analysis by secondary ion mass spectrometry (SIMS) showed increased hydrogen concentration at more cathodic controlled potentials. Optical microscopy and scanning electron microscopy (SEM) were used to evaluate the morphology of cracking both at the primary fracture face and the secondary cracks along the gage section of the broken tensile specimen. Transgranular secondary cracks were observed in both alloys possibly resulting from the formation of brittle titanium hydrides due to cathodic charging. The primary fracture face was characterized

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

    CERN Document Server

    Baglin, V; Gröbner, Oswald

    1998-01-01

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

  9. Containment

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    The primary mission of the Containment Group is to ensure that underground nuclear tests are satisfactorily contained. The main goal is the development of sound technical bases for containment-related methodology. Major areas of activity include siting, geologic description, emplacement hole stemming, and phenomenological predictions. Performance results of sanded gypsum concrete plugs on the Jefferson, Panamint, Cornucopia, Labquark, and Bodie events are given. Activities are also described in the following areas: computational capabilities site description, predictive modeling, and cavity-pressure measurement. Containment publications are listed. 8 references

  10. Determination of Fire Enviroment in Stacked Cargo Containers with Radioactive Materials Packages

    Energy Technology Data Exchange (ETDEWEB)

    Arviso, M.; Bobbe, J.G.; Dukart, R.D.; Koski, J.A.

    1999-05-01

    Results from a Fire Test with a three-by-three stack of standard 6 m long International Standards Organization shipping containers containing combustible fuels and empty radioactive materials packages are reported and discussed. The stack is intended to simulate fire conditions that could occur during on-deck stowage on container cargo ships. The fire is initated by locating the container stack adjacent to a 9.8 x 6 m pool fire. Temperatures of both cargoes (empty and simulated radioactive materials packages) and containers are recorded and reported. Observations on the duration, intensity and spread of the fire are discussed. Based on the results, models for simulation of fire exposure of radioactive materials packages in such fires are suggested.

  11. Use of the radiochemical method for identification of seized nuclear material - container

    International Nuclear Information System (INIS)

    Rosskopfova, O.; Matel, L.; Rajec, P.; Dobias, M.

    2003-01-01

    Analysis of the nuclear material of the seized illicit trafficking container was performed in the laboratory LARCHA. The container was professionally taken apart in the hot cell of HUMA-LAB APEKO. The compact core of container with mass of 36.00 kg was separated and analysed in the LARCHA laboratory. The results of alpha spectrometry proved that the core of container was made of depleted uranium. A suspicious container was seized, that was believed to be used for transportation of radioactive and nuclear materials. Container, according to Slovak Law, was transported to a Civil defence and police started with expertise analysis. On the basis of previous results, it was assumed that the core of the container was made of a nuclear material. The container was cut inside a hot cell to separate parts at HUMA-Lab Kosice and individual parts were prepared for analyses. From shape and type of the detained container it was assumed that the core of the container was made of depleted uranium and suspected cut pieces were sent to the accredited laboratory LARCHA Bratislava for isotope analysis. Samples from different parts of the container were analysed by using HPGe gamma. The results of gamma spectrometry proved the presence of Co-60, Cs-137 and U-235 with values 11.8 Bq, 5.6 Bq a 3.8 Bq respectively. On the basis of measured results it was calculated mass content (%) of uranium isotopes in samples. It was concluded that: - container is composed of different parts, which are made from steel, lead and uranium; - the wipe tests proved the presence of radionuclides of Cs-137 and Co-60; - the core of the container is made of depleted uranium; - chosen methods, which were used proved that are suitable for analysis and identification of nuclear materials. (authors)

  12. Containers for short-term storage of nuclear materials at the Los Alamos plutonium facility

    International Nuclear Information System (INIS)

    Hagan, R.; Gladson, J.

    1997-01-01

    The Los Alamos Plutonium Facility for the past 18 yr has stored nuclear samples for archiving and in support of nuclear materials research and processing programs. In the past several years, a small number of storage containers have been found in a deteriorated condition. A failed plutonium container can cause personnel contamination exposure and expensive physical area decontamination. Containers are stored in a physically secure radiation area vault, making close inspection costly in the form of personnel radiation exposure and work time. A moderate number of these containers are used in support of plutonium processing and must withstand daily handling abuse. A 2-yr evaluation of failed containers and those that have shown no deterioration has been conducted. Based on that study, a program was established to formalize our packing methods and materials and standardize the size and shape of containers that are used for short-term use. A standardized set of containers was designed, evaluated, tested, and procured for use in the facility. This paper reviews our vault storage problems, shows some failed containers, and presents our planned solutions to provide safe and secure containment of nuclear materials

  13. Waste handling and REACH : Recycling of materials containing SVHCs: daily practice challenges

    NARCIS (Netherlands)

    Janssen MPM; van Broekhuizen FA; MSP; M&V

    2017-01-01

    To achieve a circular economy it is essential to recycle substances, materials and products created by that economy. Recycling, however, becomes more difficult when said materials and products contain substances that are so hazardous that their use is restricted. This is the case with any substance

  14. 40 CFR 1612.3 - Published reports and material contained in the public incident investigation dockets.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Published reports and material... Published reports and material contained in the public incident investigation dockets. (a) Demands for published investigation reports should be directed to the Office of Congressional and Public Affairs, U.S...

  15. 19 CFR 10.602 - Packing materials and containers for shipment.

    Science.gov (United States)

    2010-04-01

    ...-Central America-United States Free Trade Agreement Rules of Origin § 10.602 Packing materials and... regional value content calculation. Packing materials and containers for shipment, as defined in § 10.593(m) of this subpart, are to be disregarded in determining the regional value content of a good imported...

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

    International Nuclear Information System (INIS)

    Raske, D.T.; Wang, Z.

    1992-01-01

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

  17. Standard Guide for Unrestricted Disposition of Bulk Materials Containing Residual Amounts of Radioactivity

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This guide covers the techniques for obtaining approval for release of materials encountered in decontamination and decommissioning (D&D) from restricted use. This would be addressed in the decommissioning plan (E 1281). It applies to materials that do not meet any of the requirements for regulatory control because of radioactivity content. Fig. 1 shows the logic diagram for determining the materials that could be considered for release. Materials that negotiate this logic tree are referred to as “candidate for release based on dose.” 1.2 The objective of this guide is to provide a methodology for distinguishing between material that must be carefully isolated to prevent human contact from that that can be recycled or otherwise disposed of. It applies to material in which the radioactivity is dispersed more or less uniformly throughout the volume of the material (termed residual in bulk form) as opposed to surface contaminated objects. 1.3 Surface contaminated objects are materials externally co...

  18. Direct vitrification of plutonium-containing materials (PCM`s) with the glass material oxidation and dissolution system (GMODS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W. Beahm, E.C.; Parker, G.W.; Rudolph, J.C.; Haas, P.A.; Malling, G.F.; Elam, K.; Ott, L.

    1995-10-30

    The end of the cold war has resulted in excess PCMs from nuclear weapons and associated production facilities. Consequently, the US government has undertaken studies to determine how best to manage and dispose of this excess material. The issues include (a) ensurance of domestic health, environment, and safety in handling, storage, and disposition, (b) international arms control agreements with Russia and other countries, and (c) economics. One major set of options is to convert the PCMs into glass for storage or disposal. The chemically inert characteristics of glasses make them a desirable chemical form for storage or disposal of radioactive materials. A glass may contain only plutonium, or it may contain plutonium along with other radioactive materials and nonradioactive materials. GMODS is a new process for the direct conversion of PCMs (i.e., plutonium metal, scrap, and residues) to glass. The plutonium content of these materials varies from a fraction of a percent to pure plutonium. GMODS has the capability to also convert other metals, ceramics, and amorphous solids to glass, destroy organics, and convert chloride-containing materials into a low-chloride glass and a secondary clean chloride salt strewn. This report is the initial study of GMODS for vitrification of PCMs as input to ongoing studies of plutonium management options. Several tasks were completed: initial analysis of process thermodynamics, initial flowsheet analysis, identification of equipment options, proof-of-principle experiments, and identification of uncertainties.

  19. Design and construction of thermal desorption measurement system for tritium contained materials

    International Nuclear Information System (INIS)

    Hara, M.; Hatano, Y.; Calderoni, P.; Shimada, M.

    2014-01-01

    The dual-mode thermal desorption analysis system was designed and built in Idaho National Laboratory (INL) to examine the evolution of the hydrogen isotope gas from materials. The system is equipped with a mass spectrometer for stable hydrogen isotopes and an ionization chamber for tritium components. The performance of the system built was tested with using tritium contained materials. The evolution of tritiated gas species from contaminated materials was measured successfully by using the system. (author)

  20. Calculations of radiation damage in target, container and window materials for spallation neutron sources

    International Nuclear Information System (INIS)

    Wechsler, M.S.; Mansur, L.K.

    1996-01-01

    Radiation damage in target, container, and window materials for spallation neutron sources is am important factor in the design of target stations for accelerator-driver transmutation technologies. Calculations are described that use the LAHET and SPECTER codes to obtain displacement and helium production rates in tungsten, 316 stainless steel, and Inconel 718, which are major target, container, and window materials, respectively. Results are compared for the three materials, based on neutron spectra for NSNS and ATW spallation neutron sources, where the neutron fluxes are normalized to give the same flux of neutrons of all energies

  1. Piezoelectric ceramic material, containing PbNb2O6, K2Nb2O6

    International Nuclear Information System (INIS)

    Fesenko, E.G.; Filip'ev, V.S.; Razumovskaya, O.N.; Cherner, Ya.E.; Rudkovskaya, L.M.; Zav'yalov, V.P.; Molchanova, R.A.; Kryshtop, V.G.; Panich, A.E.; Servuli, V.A.

    1984-01-01

    A new piezoelectric ceramic material including PbNb 2 O 6 , K 2 Nb 2 O 6 is prepared. Above the new material contains Nb 2 O 5 . The invention relates to piezotechnique. The principal advantage of this material for acoustic converters is high anisotropy of piezoelectric properties as well as high Curie temperature (T C =539-553 deg C). The composition containing 93.96 mole% PbNb 2 O 6 ; 2.48 mole% K 2 Nb 2 O 6 and 3.56 mole% Nb 2 O 5 has optimum content of parameters

  2. Electroactive mesoporous yttria stabilized zirconia containing platinum or nickel oxide nanoclusters: a new class of solid oxide fuel cell electrode materials

    Energy Technology Data Exchange (ETDEWEB)

    Mamak, M.; Coombs, N.; Ozin, G.A. [Toronto Univ., ON (Canada). Dept. of Chemistry

    2001-02-01

    The electroactivity of surfactant-templated mesoporous yttria stabilized zirconia, containing nanoclusters of platinum or nickel oxide, is explored by alternating current (AC) complex impedance spectroscopy. The observed oxygen ion and mixed oxygen ion-electron charge-transport behavior for these materials, compared to the sintered-densified non-porous crystalline versions, is ascribed to the unique integration of mesoporosity and nanocrystallinity within the binary and ternary solid solution microstructure. These attributes inspire interest in this new class of materials as candidates for the development of improved performance solid oxide fuel cell electrodes. (orig.)

  3. Detection of tiny amounts of fissile materials in large-sized containers with radioactive waste

    Directory of Open Access Journals (Sweden)

    Batyaev V.F.

    2018-01-01

    Full Text Available The paper is devoted to non-destructive control of tiny amounts of fissile materials in large-sized containers filled with radioactive waste (RAW. The aim of this work is to model an active neutron interrogation facility for detection of fissile ma-terials inside NZK type containers with RAW and determine the minimal detectable mass of U-235 as a function of various param-eters: matrix type, nonuniformity of container filling, neutron gen-erator parameters (flux, pulse frequency, pulse duration, meas-urement time. As a result the dependence of minimal detectable mass on fissile materials location inside container is shown. Nonu-niformity of the thermal neutron flux inside a container is the main reason of the space-heterogeneity of minimal detectable mass in-side a large-sized container. Our experiments with tiny amounts of uranium-235 (<1 g confirm the detection of fissile materials in NZK containers by using active neutron interrogation technique.

  4. MULCHES AND OTHER COVER MATERIALS TO REDUCE WEED GROWTH IN CONTAINER-GROWN NURSERY STOCK.

    Science.gov (United States)

    Rys, F; Van Wesemael, D; Van Haecke, D; Mechant, E; Gobin, B

    2014-01-01

    Due to the recent EU-wide implementation of Integrated Pest Management (IPM), alternative methods to reduce weed growth in container-grown nursery stock are needed to cut back the use of herbicides. Covering the upper layer of the substrate is known as a potential method to prevent or reduce weed growth in plant containers. As a high variety of mulches and other cover materials are on the market, however, it is no longer clear for growers which cover material is most efficient for use in containers. Therefore, we examined the effect on weed growth of different mulches and other cover materials, including Pinus maritima, P. sylvestris, Bio-Top Basic, Bio-Top Excellent, coco chips fine, hemp fibres, straw pellets, coco disk 180LD and jute disk. Cover materials were applied immediately after repotting of Ligustrum ovalifolium or planting of Fagus sylvatica. At regular times, both weed growth and side effects (e.g., plant growth, water status of the substrate, occurrence of mushrooms, foraging of birds, complete cover of the substrate and fixation) were assessed. All examined mulches or other cover materials were able to reduce weed growth on the containers during the whole growing season. Weed suppression was even better than that of a chemical treated control. Although all materials showed some side effects, the impact on plant growth is most important to the grower and depends not only on material characteristics (e.g., biodegradation, nutrient leaching and N-immobilisation) but also on container size and climatic conditions. In conclusion, mulches and other cover materials can be a valuable tool within IPM to lower herbicide use. To enable a deliberate choice of which cover material is best used in a specific situation more research is needed on lifespan and stability as well as on economic characteristics of the materials.

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

    Directory of Open Access Journals (Sweden)

    Rubenilson Caldas Valois

    2014-12-01

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

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

    Science.gov (United States)

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

    2017-03-01

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

  7. Program to develop analytical tools for environmental and safety assessment of nuclear material shipping container systems

    International Nuclear Information System (INIS)

    Butler, T.A.

    1978-11-01

    This paper describes a program for developing analytical techniques to evaluate the response of nuclear material shipping containers to severe accidents. Both lumped-mass and finite element techniques are employed to predict shipping container and shipping container-carrier response to impact. The general impact problem is computationally expensive because of its nonlinear, three-dimensional nature. This expense is minimized by using approximate models to parametrically identify critical cases before more exact analyses are performed. The computer codes developed for solving the problem are being experimentally substantiated with test data from full-scale and scale-model container drop tests. 6 figures, 1 table

  8. Preparation and multi-properties determination of radium-containing rocklike material

    Science.gov (United States)

    Hong, Changshou; Li, Xiangyang; Zhao, Guoyan; Jiang, Fuliang; Li, Ming; Zhang, Shuai; Wang, Hong; Liu, Kaixuan

    2018-02-01

    The radium-containing rocklike material were fabricated using distilled water, ordinary Portland cement and additives mixed aggregates and admixtures according to certain proportion. The physico-mechanical properties as well as radioactive properties of the prepared rocklike material were measured. Moreover, the properties of typical granite sample were also investigated. It is found on one hand, similarities exist in physical and mechanical properties between the rocklike material and the granite sample, this confirms the validity of the proposed method; on the other hand, the rocklike material generally performs more remarkable radioactive properties compared with the granite sample, while radon diffusive properties in both materials are essentially matching. This study will provide a novel way to prepare reliable radium-containing samples for radon study of underground uranium mine.

  9. Degradation mode survey candidate titanium-base alloys for Yucca Mountain project waste package materials. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Gdowski, G.E.

    1997-12-01

    The Yucca Mountain Site Characterization Project (YMP) is evaluating materials from which to fabricate high-level nuclear waste containers (hereafter called waste packages) for the potential repository at Yucca Mountain, Nevada. Because of their very good corrosion resistance in aqueous environments titanium alloys are considered for container materials. Consideration of titanium alloys is understandable since about one-third (in 1978) of all titanium produced is used in applications where corrosion resistance is of primary importance. Consequently, there is a considerable amount of data which demonstrates that titanium alloys, in general, but particularly the commercial purity and dilute {alpha} grades, are highly corrosion resistant. This report will discuss the corrosion characteristics of Ti Gr 2, 7, 12, and 16. The more highly alloyed titanium alloys which were developed by adding a small Pd content to higher strength Ti alloys in order to give them better corrosion resistance will not be considered in this report. These alloys are all two phase ({alpha} and {beta}) alloys. The palladium addition while making these alloys more corrosion resistant does not give them the corrosion resistance of the single phase {alpha} and near-{alpha} (Ti Gr 12) alloys.

  10. The adhesion characteristics of protective coating materials for the containment structure in nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Sang-Kook; Shin, Jae-Chul

    2003-01-01

    Protective coating materials used in the containment structures should be durable for the designed 30 to 40 year lifetime of a nuclear power plant. At the present, these materials have not yet been developed. Therefore it is very important to keep the durability of the protective coating materials through persistent maintenance, and in order to achieve this, understanding the adhesion characteristics of the coating materials is of utmost importance. Therefore, this study attempts to find any methods for durability maintenance of these protective coating materials. To accomplish these aims, this study applied an experimental deterioration environment condition relevant to Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB), categorized as of Design Basis Accident (DBA), onto steel liner plate specimens covered with protective coating materials. Adhesion tests were performed on these deteriorated coating materials to characterize the physical properties and through these tests, the quantitative adhesion characteristics according to the history of deterioration environment were found

  11. Effects of a range of machined and ground surface finishes on the simulated reactor helium corrosion of several candidate structural materials

    International Nuclear Information System (INIS)

    Thompson, L.D.

    1981-02-01

    This report discusses the corrosion behavior of several candidate reactor structural alloys in a simulated advanced high-temperature gas-cooled reactor (HTGR) environment over a range of lathe-machined and centerless-ground surface finishes. The helium environment contained 50 Pa H 2 /5 Pa CO/5 Pa CH 4 / 2 O (500 μatm H 2 /50 μatm CO/50 μatm CH 4 / 2 O) at 900 0 C for a total exposure of 3000 h. The test alloys included two vacuum-cast superalloys (IN 100 and IN 713LC); a centrifugally cast austenitic alloy (HK 40); three wrought high-temperature alloys (Alloy 800H, Hastelloy X, and Inconel 617); and a nickel-base oxide-dispersion-strengthened alloy (Inconel MA 754). Surface finish variations did not affect the simulated advanced-HTGR corrosion behavior of these materials. Under these conditions, the availability of reactant gaseous impurities controls the kinetics of the observed gas-metal interactions. Variations in the near-surface activities and mobilities of reactive solute elements, such as chromium, which might be expected to be affected by changes in surface finish, do not seem to greatly influence corrosion in this simulated advanced HTGR environment. 18 figures, 4 tables

  12. A Review of Removable Surface Contamination on Radioactive Materials Transportation Containers

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, Jr, W. E.; Watson, E. C.; Murphy, D. W.; Harrer, B. J.; Harty, R.; Aldrich, J. M.

    1981-05-01

    This report contains the results of a study sponsored by the U.S. Nuclear Regulatory Commission (NRC) of removable surface contamination on radioactive materials transportation containers. The purpose of the study is to provide information to the NRC during their review of existing regulations. Data was obtained from both industry and literature on three major topics: 1) radiation doses, 2) economic costs, and 3) contamination frequencies. Containers for four categories of radioactive materials are considered including radiopharmaceuticals, industrial sources, nuclear fuel cycle materials, and low-level radioactive waste. Assumptions made in this study use current information to obtain realistic yet conservative estimates of radiation dose and economic costs. Collective and individual radiation doses are presented for each container category on a per container basis. Total doses, to workers and the public, are also presented for spent fuel cask and low-level waste drum decontamination. Estimates of the additional economic costs incurred by lowering current limits by factors of 10 and 100 are presented. Current contamination levels for each category of container are estimated from the data collected. The information contained in this report is designed to be useful to the NRC in preparing their recommendations for new regulations.

  13. Methods of pretreating comminuted cellulosic material with carbonate-containing solutions

    Energy Technology Data Exchange (ETDEWEB)

    Francis, Raymond

    2012-11-06

    Methods of pretreating comminuted cellulosic material with an acidic solution and then a carbonate-containing solution to produce a pretreated cellulosic material are provided. The pretreated material may then be further treated in a pulping process, for example, a soda-anthraquinone pulping process, to produce a cellulose pulp. The pretreatment solutions may be extracted from the pretreated cellulose material and selectively re-used, for example, with acid or alkali addition, for the pretreatment solutions. The resulting cellulose pulp is characterized by having reduced lignin content and increased yield compared to prior art treatment processes.

  14. Anthropogenic materials and products containing natural radionuclides. Pt. 1a. Radiation properties of raw materials and waste materials. A literature study

    International Nuclear Information System (INIS)

    Reichelt, A.; Roehrer, J.; Lehmann, K.H.

    1995-12-01

    Cased on the literature study, the publication presents relevant data on raw materials and wastes containing natural radionuclides. The study is part 1a of the project on ''Anthropogenic materials and waste materials containing natural radionuclides''. Part 1 of the project gives data and information on about 100 different materials and wastes or products for household or industrial applications which contain significant amounts of natural radioactivity. In addition, part 1 presents for some of these materials information on their applications, consumption, radioactivity and resulting radiation doses. The raw materials and waste materials on the list in part 1 are characterised in this 1a report. Wherever appropriate, two or more materials are dealt with in one chapter, as e.g. felspar and felspar sands (pegmatite), talcum, and soapstone. The wastes are dealt with in the chapters discussing the relevant raw materials. The information given is as derived from the literature and does not include comments or evaluation by the authors of this report. Whenever the literature study did not yield information on radiological aspects of a material on the list, an appropriate notice is given. (Orig./DG) [de

  15. Novel Carbon (C)-Boron (B)-Nitrogen (N)-Containing H2 Storage Materials

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Shih-Yuan [Boston College, Chestnut Hill, MA (United States); Giustra, Zachary X. [Boston College, Chestnut Hill, MA (United States); Autrey, Tom [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dixon, David A. [Univ. of Alabama, Tuscaloosa, AL (United States); Osenar, Paul [Protonex Technology Corporation, Southborough, MA (United States)

    2017-09-20

    The following summarizes the research conducted for DOE project DE-EE0005658 “Novel Carbon(C)-Boron(B)-Nitrogen(N)-Containing H2 Storage Materials”. This work focused in part on the continued study of two materials identified from the preceding project DE-FG360GO18143 (“Hydrogen Storage by Novel CBN Heterocycle Materials”) as lead candidates to meet the DOE technical targets for either vehicular or non-automotive hydrogen storage applications. Specifically, a room-temperature liquid, 3-methyl-1,2-cyclopentane (B), and a high H2 capacity solid, 1,2-BN-cyclohexane (J), were selected for further characterization and performance optimization. In addition to these compounds, the current project also aimed to prepare several new materials predicted to be disposed towards direct reversibility of H2 release and uptake, a feature deemed critical to achieving efficient recycling of spent fuel end products. To assist in the rational design of these and other next-generation materials, this project undertook to investigate the mechanism of hydrogen release from established compounds (mainly B and J) using a combined experimental/computational approach. Among this project’s signature accomplishments, the preliminary synthetic route to B was optimized for production on decagram scale. With such quantities of material available, its performance in powering an actual 30 W proton exchange membrane (PEM) fuel cell stack was tested and found to be identical to that of facility H2. Despite this positive proof-of-concept achievement, however, further consideration of neat B as a potential hydrogen storage material was abandoned due to evidence of thermal instability. Specifically, mass spectrometry-coupled thermogravimetric analysis (TGA-MS) revealed significant H2 release from B to initiate at 50 °C, well below the 60 °C minimum threshold set by the DOE. This result prompted a more extensive investigation in the decomposition mechanism of B vis-à-vis that of J, which

  16. To the problem of structural materials serviceability in nitrogen-hydrogen-containing environments

    International Nuclear Information System (INIS)

    Bichuya, A.L.

    1982-01-01

    The analysis of the factors which affect high-temperature serviceability of structural materials in nitrogen-hydrogen-containing environments, in particular in ammonia, has been carried out on the basis of the published and own experimental data. It is shown that the observed reduction of serviceability of structural materials, under the effect of high temperatures and nitrogen-hydrogen-containing environments, can occur as a result of corrosion failure connected with nitriding, and also hydrogen embrittlement appearing as a result of the penetration of hydrogen formed during adsorbed gaseous phase dissociation on the metal being deformed. The suggested scheme of high-temperature metal fracture under the effect of nitrogen-hydrogen-containing environments, that in contrast to the previous ones includes the factor of hydrogen ebrittlement, allows to give a real estimation of structional materials serviceability under product service conditions

  17. Preliminary Study of Applying Phase Change Materials (PCM) for Containment Passive Cooling

    International Nuclear Information System (INIS)

    Ko, A Reum; Lee, Jeong Ik; Yoon, Ho Joon

    2016-01-01

    Most of Pressurized Water Reactor (PWR) containments use fan cooler systems and containment spray systems. However, the importance of passive safety system has increased after the Fukushima accident. As the main passive safety system, Passive Containment Cooling System (PCCS), which utilizes natural phenomena to remove the heat released from the reactor, is suggested in the advanced pressurized water reactor (APWR). To increase the efficiency of passive cooling, additional passive containment cooling method using Phase Change Material (PCM) is suggested in this paper. For containment using PCMs, there are many advantages. Phase Change Material (PCM) is proposed as an additional passive containment cooling method to increase the efficiency of passive cooling in this paper. To apply proper PCMs to containment, commercially available PCMs were screened while reviewing thermophysical properties data and suggested selection criteria. A sensitivity study was also carried out to identify the effect of potential installation location of PCM using the CAP code. The pressure of containment in most cases showed slightly higher than that of the initial case. For the temperature of steam and water and humidity, similar results with the initial case were showed in most cases

  18. Preliminary Study of Applying Phase Change Materials (PCM) for Containment Passive Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Ko, A Reum; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [KUSTAR, Abu Dhabi (United Arab Emirates)

    2016-05-15

    Most of Pressurized Water Reactor (PWR) containments use fan cooler systems and containment spray systems. However, the importance of passive safety system has increased after the Fukushima accident. As the main passive safety system, Passive Containment Cooling System (PCCS), which utilizes natural phenomena to remove the heat released from the reactor, is suggested in the advanced pressurized water reactor (APWR). To increase the efficiency of passive cooling, additional passive containment cooling method using Phase Change Material (PCM) is suggested in this paper. For containment using PCMs, there are many advantages. Phase Change Material (PCM) is proposed as an additional passive containment cooling method to increase the efficiency of passive cooling in this paper. To apply proper PCMs to containment, commercially available PCMs were screened while reviewing thermophysical properties data and suggested selection criteria. A sensitivity study was also carried out to identify the effect of potential installation location of PCM using the CAP code. The pressure of containment in most cases showed slightly higher than that of the initial case. For the temperature of steam and water and humidity, similar results with the initial case were showed in most cases.

  19. Effect of Eu magnetism on the electronic properties of the candidate Dirac material EuMnBi2

    Science.gov (United States)

    May, Andrew F.; McGuire, Michael A.; Sales, Brian C.

    2014-08-01

    The crystal structure and physical properties of the layered material EuMnBi2 have been characterized by measurements on single crystals. EuMnBi2 is isostructural with the Dirac material SrMnBi2 based on single-crystal x-ray diffraction, crystallizing in the I4/mmm space group (No. 139). Magnetic susceptibility measurements suggest antiferromagnetic (AFM) ordering of moments on divalent Eu ions near TN=22 K. For low fields, the ordered Eu moments are aligned along the c axis, and a spin flop is observed near 5.4 T at 5 K. The moment is not saturated in an applied field of 13 T at 5 K, which is uncommon for compounds containing Eu2+. The magnetic behavior suggests an anisotropy enhancement via interaction between Eu and the Mn moments that appear to be ordered antiferromagnetically below ≈310 K. A large increase in the magnetoresistance is observed across the spin flop, with absolute magnetoresistance reaching ≈650% at 5 K and 12 T. Hall effect measurements reveal a decrease in the carrier density below TN, which implies a manipulation of the Fermi surface by magnetism on the sites surrounding the Bi square nets that lead to Dirac cones in this family of materials.

  20. Nuclear power plant containment metallic pressure boundary materials and plans for collecting and presenting their properties

    International Nuclear Information System (INIS)

    Oland, C.B.

    1995-04-01

    A program is being conducted at the Oak Ridge National Laboratory (ORNL to assist the Nuclear Regulatory Commission (NRC)) in their assessment of the effects of degradation (primarily corrosion) on the structural capacity and leaktight integrity of metal containments and steel liners of reinforced concrete structures in nuclear power plants. One of the program objectives is to characterize and quantify manifestations of corrosion on the properties of steels used to construct containment pressure boundary components. This report describes a plan for use in collecting and presenting data and information on ferrous alloys permitted for use in construction of pressure retaining components in concrete and metal containments. Discussions about various degradation mechanisms that could potentially affect the mechanical properties of these materials are also included. Conclusions and recommendations presented in this report will be used to guide the collection of data and information that will be used to prepare a material properties data base for containment steels

  1. Automated nuclear material recovery and decontamination of large steel dynamic experiment containers

    International Nuclear Information System (INIS)

    Dennison, D.K.; Gallant, D.A.; Nelson, D.C.; Stovall, L.A.; Wedman, D.E.

    1999-01-01

    A key mission of the Los Alamos National Laboratory (LANL) is to reduce the global nuclear danger through stockpile stewardship efforts that ensure the safety and reliability of nuclear weapons. In support of this mission LANL performs dynamic experiments on special nuclear materials (SNM) within large steel containers. Once these experiments are complete, these containers must be processed to recover residual SNM and to decontaminate the containers to below low level waste (LLW) disposal limits which are much less restrictive for disposal purposes than transuranic (TRU) waste limits. The purpose of this paper is to describe automation efforts being developed by LANL for improving the efficiency, increasing worker safety, and reducing worker exposure during the material cleanout and recovery activities performed on these containers

  2. Numerical Simulation on Dense Packing of Granular Materials by Container Oscillation

    OpenAIRE

    Jun Liu; Dongxu You

    2013-01-01

    The packing of granular materials is a basic and important problem in geomechanics. An approach, which generates dense packing of spheres confined in cylindrical and cuboidal containers in three steps, is introduced in this work. A loose packing structure is first generated by means of a reference lattice method. Then a dense packing structure is obtained in a container by simulating dropping of particles under gravitational forces. Furthermore, a scheme that makes the bottom boundary fluctua...

  3. Hydrolysis of cellulose-containing materials by cellulase of the Trichoderma lignorum OM 534 fungus

    Energy Technology Data Exchange (ETDEWEB)

    Romanov, S L; Lobanok, A G

    1977-01-01

    Of the cellulose containing materials, hydrocellulose was most easily degraded while lignocellulose was hardest to break down with cellulase from T. lignorum grown on lactose or cellulose. Grinding and heat treatment (at 200/sup 0/) of lignocellulose enhanced its enzymic degradability. Hydrolysis was highest by cellulase from lactose-cultured Trichoderma. The hydrolysis products contained glucose, galactose, xylose, and mannose. Filtrates from T. lignorum grown on a lignocellulose were enzymically active after purification.

  4. 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...

  5. Heat-processing method and facility for helium-containing metal material

    International Nuclear Information System (INIS)

    Kato, Takahiko; Kodama, Hideyo; Matsumoto, Toshimi; Aono, Yasuhisa; Nagata, Tetsuya; Hattori, Shigeo; Kaneda, Jun-ya; Ono, Shigeki.

    1996-01-01

    Electric current is supplied to an objective portion of a He-containing metal material to be applied with heat processing without causing melting, to decrease the He content of the portion. Subsequently, the defect portion of the tissues of the He-containing metal is modified by heating the portion with melting. Since electric current can be supplied to the metal material in a state where the metal material is heated and the temperature thereof is elevated, an effect of further reducing the He content can be obtained. Further, if the current supply and/or the heating relative to the metal material is performed in a vacuum or inert gas atmosphere, an effect of reducing the degradation of the surface of the objective portion to be supplied with electric current can be obtained. (T.M.)

  6. Alkaline degradation of organic materials contained in TRU wastes under repository conditions

    International Nuclear Information System (INIS)

    Otsuka, Yoshiki; Banba, Tsunetaka

    2007-09-01

    Alkaline degradation tests for 9 organic materials were conducted under the conditions of TRU waste disposal: anaerobic alkaline conditions. The tests were carried out at 90degC for 91 days. The sample materials for the tests were selected from the standpoint of constituent organic materials of TRU wastes. It has been found that cellulose and plastic solidified products are degraded relatively easily and that rubbers are difficult to degrade. It could be presumed that the alkaline degradation of organic materials occurs starting from the functional group in the material. Therefore, the degree of degradation difficulty is expected to be dependent on the kinds of functional group contained in the organic material. (author)

  7. Dose and radon measurements inside houses containing ash as building material

    International Nuclear Information System (INIS)

    Bodnar, R.; Lendvai, Z.; Somlai, J.; Nemeth, C.

    1996-01-01

    Radon concentration and external dose have been measured in dwellings that contain by-products of coal burning for building materials. The concentrations of 40 K, 232 Th, 238 U and 226 Ra have been determined in the materials. The date are analyzed according to indices frequently used for decision of utilizing the by-products. The observed daily fluctuation of the radon concentration in dwellings might exceed a factor of 5. (author)

  8. Comparative analysis of non-destructive methods to control fissile materials in large-size containers

    Directory of Open Access Journals (Sweden)

    Batyaev V.F.

    2017-01-01

    Full Text Available The analysis of various non-destructive methods to control fissile materials (FM in large-size containers filled with radioactive waste (RAW has been carried out. The difficulty of applying passive gamma-neutron monitoring FM in large containers filled with concreted RAW is shown. Selection of an active non-destructive assay technique depends on the container contents; and in case of a concrete or iron matrix with very low activity and low activity RAW the neutron radiation method appears to be more preferable as compared with the photonuclear one.

  9. Comparative analysis of non-destructive methods to control fissile materials in large-size containers

    Science.gov (United States)

    Batyaev, V. F.; Sklyarov, S. V.

    2017-09-01

    The analysis of various non-destructive methods to control fissile materials (FM) in large-size containers filled with radioactive waste (RAW) has been carried out. The difficulty of applying passive gamma-neutron monitoring FM in large containers filled with concreted RAW is shown. Selection of an active non-destructive assay technique depends on the container contents; and in case of a concrete or iron matrix with very low activity and low activity RAW the neutron radiation method appears to be more preferable as compared with the photonuclear one. Note to the reader: the pdf file has been changed on September 22, 2017.

  10. Development of assessment methods for transport and storage containers with higher content of metallic recycling material

    International Nuclear Information System (INIS)

    Zencker, U.; Qiao Linan; Droste, B.

    2004-01-01

    The mechanical behaviour of transport and storage containers made of ductile cast iron melted with higher content of metallic recycling material from decommissioning and dismantling of nuclear installations is investigated. With drop tests of cubic container-like models, the influence of different real targets on the stresses in the cask body and the fracture behaviour is examined. A test stand foundation is suggested, which can be manufactured simply and improves the reproducibility of the test results strongly. The test objects are partially equipped with artificial cracklike defects. Dynamic fracture mechanics analyses of these defects were performed by means of finite element calculations to uncover safety margins. Numerous test results show depending on the requirements that containers for final disposal can be built by means of a ductile cast iron with fracture toughness more than half under the lower bound value for the licensed material qualities yet. The application limits of the material are determined also by the opportunities of the safety assessment methods. This project supports the application of brittle fracture safe transport and storage packages for radioactive materials as recommended in App. VI of the Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material (IAEA No. TS-G-1.1)

  11. Development of assessment methods for transport and storage containers with higher content of metallic recycling material

    Energy Technology Data Exchange (ETDEWEB)

    Zencker, U.; Qiao Linan; Droste, B. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

    2004-07-01

    The mechanical behaviour of transport and storage containers made of ductile cast iron melted with higher content of metallic recycling material from decommissioning and dismantling of nuclear installations is investigated. With drop tests of cubic container-like models, the influence of different real targets on the stresses in the cask body and the fracture behaviour is examined. A test stand foundation is suggested, which can be manufactured simply and improves the reproducibility of the test results strongly. The test objects are partially equipped with artificial cracklike defects. Dynamic fracture mechanics analyses of these defects were performed by means of finite element calculations to uncover safety margins. Numerous test results show depending on the requirements that containers for final disposal can be built by means of a ductile cast iron with fracture toughness more than half under the lower bound value for the licensed material qualities yet. The application limits of the material are determined also by the opportunities of the safety assessment methods. This project supports the application of brittle fracture safe transport and storage packages for radioactive materials as recommended in App. VI of the Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material (IAEA No. TS-G-1.1).

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

    Science.gov (United States)

    Place, Benjamin J

    2017-05-01

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

  13. Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

    International Nuclear Information System (INIS)

    Byun, T.S.; Farrell, K.; Lee, E.H.; Mansur, L.K.; Maloy, S.A.; James, M.R.; Johnson, W.R.

    2002-01-01

    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 0.54-2.53 dpa at 30-100 deg. C. Tensile testing was performed at room temperature (20 deg. C) and 164 deg. C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 deg. C decreased the strength by 13-18% and the ductility by 8-36%. 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. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability

  14. Neutron data error estimate of criticality calculations for lattice in shielding containers with metal fissionable materials

    International Nuclear Information System (INIS)

    Vasil'ev, A.P.; Krepkij, A.S.; Lukin, A.V.; Mikhal'kova, A.G.; Orlov, A.I.; Perezhogin, V.D.; Samojlova, L.Yu.; Sokolov, Yu.A.; Terekhin, V.A.; Chernukhin, Yu.I.

    1991-01-01

    Critical mass experiments were performed using assemblies which simulated one-dimensional lattice consisting of shielding containers with metal fissile materials. Calculations of the criticality of the above assemblies were carried out using the KLAN program with the BAS neutron constants. Errors in the calculations of the criticality for one-, two-, and three-dimensional lattices are estimated. 3 refs.; 1 tab

  15. Modeling property evolution of container materials used in nuclear waste storage

    Science.gov (United States)

    Li, Dongsheng; Garmestani, Hamid; Khaleel, Moe; Sun, Xin

    2010-03-01

    Container materials under irradiation for a long time will raise high energy in the structure to generate critical structural damage. This study investigated what kind of mesoscale microstructure will be more resistant to radiation damage. Mechanical properties evolution during irradiation was modeled using statistical continuum mechanics. Preliminary results also showed how to achieve the desired microstructure with higher resistance to radiation.

  16. ENGINEERING CONTROL PRACTICES FOR REDUCING EMISSIONS DURING DRILLING OF ASBESTOS-CONTAINING FLOORING MATERIALS

    Science.gov (United States)

    This report describes the implementation and testing of control measures to reduce airborne asbestos generated by the drilling of asbestos-containing flooring materials, an OSHA Class III asbestos maintenance activity. Bosch 11224 and 11222 rotary drills were fitted with shrouds ...

  17. Improved process for producing a fermentation product from a lignocellulose-containing material

    DEFF Research Database (Denmark)

    2017-01-01

    The present invention relates to the production of hydrolyzates from a lignocellulose-containing material, and to fermentation of the hydrolyzates. More specifically, the present invention relates to the detoxification of phenolic inhibitors and toxins formed during the processing of lignocellulose...

  18. A process for producing a fermentation product from a lignocellulose-containing material

    DEFF Research Database (Denmark)

    2016-01-01

    The present invention relates to the production of hydrolyzates from a lignocellulose-containing material, and to fermentation of the hydrolyzates. More specifically, the present invention relates to the detoxification of phenolic inhibitors and toxins formed during the processing of lignocellulose...

  19. 76 FR 41178 - Pesticides; Policies Concerning Products Containing Nanoscale Materials; Opportunity for Public...

    Science.gov (United States)

    2011-07-13

    ... Pesticides; Policies Concerning Products Containing Nanoscale Materials; Opportunity for Public Comment; Extension of Comment Period AGENCY: Environmental Protection Agency (EPA). ACTION: Proposed policy statement; extension of comment period. SUMMARY: EPA issued a proposed policy statement in the Federal Register of June...

  20. 19 CFR 10.462 - Packing materials and containers for shipment.

    Science.gov (United States)

    2010-04-01

    ... Free Trade Agreement Rules of Origin § 10.462 Packing materials and containers for shipment. (a... disregarded in determining the regional value content of a good imported into the United States. Accordingly, in applying either the build-down or build-up method for determining the regional value content of...

  1. Inherent N,O-containing carbon frameworks as electrode materials for high-performance supercapacitors.

    Science.gov (United States)

    Hu, Fangyuan; Wang, Jinyan; Hu, Shui; Li, Linfei; Wang, Gang; Qiu, Jieshan; Jian, Xigao

    2016-09-15

    N,O-Containing micropore-dominated materials have been developed successfully via temperature-dependent cross-linking of 4,4'-(dioxo-diphenyl-2,3,6,7-tetraazaanthracenediyl)dibenzonitrile (DPDN) monomers. By employing a molecular engineering strategy, we have designed and synthesized a series of porous heteroatom-containing carbon frameworks (PHCFs), in which nitrogen and oxygen heteroatoms are distributed homogeneously throughout the whole framework at the atomic level, which can ensure the stability of its electrical properties. The as-made PHCFs@550 exhibits a high specific capacitance of 378 F g -1 , with an excellent long cycling life, including excellent cycling stability (capacitance retention of ca. 120% over 20 000 cycles). Moreover, the successful preparation of PHCFs provides new insights for the fabrication of nitrogen and oxygen-containing electrode materials from readily available components via a facile route.

  2. Steam reforming as an alternative technique for treatment of oil sludge containing naturally occurring radioactive material

    International Nuclear Information System (INIS)

    Norasalwa Zakaria; Muhd Noor Muhd Yunus; Mohd Khairi Muhd Said; Mohamad Azman Che Mat Isa; Mohd Puad Abu

    2004-01-01

    Steam reforming treatment system is an innovative technology that holds a potential to treat mixed waste containing radioactive material. The system is utilizing the thermal heat of the superheated steam at 500 degree C to produce combustible gases and integrates it with ash melting at 1400 degree C for final destruction. In this system, liquids are evaporated, organics are converted into a hydrogen-rich gas, chlorinated compounds are converted in hydrochloric acid, and reactive chemicals in the waste containing radionuclide and heavy metals are converted into the stable product through ash melting dioxins and furans are not formed, but instead are destroyed in the reducing environment of the system. No secondary pollutants are produced from the system that requires subsequent treatment. The system is divided into three development stages, and currently the project is progressing at development stage 1. This project is an entailment of a concentrated effort to solve oil sludge containing radioactive material treatment issue. (Author)

  3. Conversion of plutonium-containing materials into borosilicate glass using the glass material oxidation and dissolution system

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Beahm, E.C.; Parker, G.W.

    1996-01-01

    The end of the cold war has resulted in excess plutonium-containing materials (PCMs) in multiple chemical forms. Major problems are associated with the long-term management of these materials: safeguards and nonproliferation issues; health, environment, and safety concerns; waste management requirements; and high storage costs. These issues can be addressed by conversion of the PCMs to glass: however, conventional glass processes require oxide-like feed materials. Conversion of PCMs to oxide-like materials followed by vitrification is a complex and expensive process. A new vitrification process has been invented, the Glass Material Oxidation and Dissolution System (GMODS) to allow direct conversion of PCMs to glass. GMODS directly converts metals, ceramics, and amorphous solids to glass; oxidizes organics with the residue converted to glass; and converts chlorides to borosilicate glass and a secondary sodium chloride stream. Laboratory work has demonstrated the conversion of cerium (a plutonium surrogate), uranium (a plutonium surrogate), Zircaloy, stainless steel, multiple oxides, and other materials to glass. Equipment options have been identified for processing rates between 1 and 100,000 t/y. Significant work, including a pilot plant, is required to develop GMODS for applications at an industrial scale

  4. Leaching assessment of road materials containing primary lead and zinc slags.

    Science.gov (United States)

    Barna, R; Moszkowicz, P; Gervais, C

    2004-01-01

    Characterisation of the leaching behaviour of waste-containing materials is a crucial step in the environmental assessment for reuse scenarios. In our research we applied the multi-step European methodology ENV 12-920 to the leaching assessment of road materials containing metallurgical slag. A Zn slag from an imperial smelting furnace (ISF) and a Pb slag from a lead blast furnace (LBF) are investigated. The two slags contain up to 11.2 wt% of lead and 3.5 wt% of zinc and were introduced as a partial substitute for sand in two road materials, namely sand-cement and sand-bitumen. At the laboratory scale, a leaching assessment was performed first through batch equilibrium leaching tests. Second, the release rate of the contaminants was evaluated using saturated leaching tests on monolithic material. Third, laboratory tests were conducted on monolithic samples under intermittent wetting conditions. Pilot-scale tests were conducted for field testing of intermittent wetting conditions. The results show that the release of Pb and Zn from the materials in a saturated scenario was controlled by the pH of the leachates. For the intermittent wetting conditions, an additional factor, blocking of the pores by precipitation during the drying phase is proposed. Pilot-scale leaching behaviour only partially matched with the laboratory-scale test results: new mass transfer mechanisms and adapted laboratory leaching tests are discussed.

  5. TEMPERATURE PREDICTION IN 3013 CONTAINERS IN K AREA MATERIAL STORAGE (KAMS) FACILITY USING REGRESSION METHODS

    International Nuclear Information System (INIS)

    Gupta, N

    2008-01-01

    3013 containers are designed in accordance with the DOE-STD-3013-2004. These containers are qualified to store plutonium (Pu) bearing materials such as PuO2 for 50 years. DOT shipping packages such as the 9975 are used to store the 3013 containers in the K-Area Material Storage (KAMS) facility at Savannah River Site (SRS). DOE-STD-3013-2004 requires that a comprehensive surveillance program be set up to ensure that the 3013 container design parameters are not violated during the long term storage. To ensure structural integrity of the 3013 containers, thermal analyses using finite element models were performed to predict the contents and component temperatures for different but well defined parameters such as storage ambient temperature, PuO 2 density, fill heights, weights, and thermal loading. Interpolation is normally used to calculate temperatures if the actual parameter values are different from the analyzed values. A statistical analysis technique using regression methods is proposed to develop simple polynomial relations to predict temperatures for the actual parameter values found in the containers. The analysis shows that regression analysis is a powerful tool to develop simple relations to assess component temperatures

  6. Eudragit ® FS 30 D polymeric films containing chondroitin sulfate as candidates for use in coating seeking modified delivery of drugs

    Directory of Open Access Journals (Sweden)

    Camila Borges dos Reis

    Full Text Available ABSTRACT Polymeric films associating different concentrations of Eudragit(r FS 30 D (EFS and chondroitin sulfate (CS were produced by casting for the development of a new target-specific site material. Formed films kept a final polymer mass of 4% (w/v in the following proportions: EFS 100:00 CS (control, EFS 95:05 CS, EFS 90:10 CS and EFS 80:20 CS. They were analyzed for physical and chemical characteristics using Fourier transform infrared spectroscopy (FTIR, scanning electron microscopy (SEM and Raman spectroscopy. Furthermore, they were characterized by their water vapor permeability and degree of hydration at different conditions simulating the gastrointestinal tract. No chemical interactions were observed between CS and EFS, suggesting only a physical interaction between them in the different combinations tested. The results suggest that EFS and CS, when combined, may form films that are candidates for coating processes seeking a modified drug delivery, especially due to the synergism between pH dependency and specific biodegradability properties by the colonic microbiota. EFS 90:10 CS proved to be the most suitable for this purpose considering hydration and permeability characteristics of different associations analyzed.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  8. Development of a container for the transportation and storage of plutonium bearing materials

    International Nuclear Information System (INIS)

    Ammerman, D.; Geinitz, R.; Thorp, D.; Rivera, M.

    1998-03-01

    There is a large backlog of plutonium contaminated materials at the Rocky Flats Environmental Technology Site near Denver, Colorado, USA. The clean-up of this site requires this material to be packaged in such a way as to allow for efficient transportation to other sites or to a permanent geologic repository. Prior to off-site shipment of the material, it may be stored on-site for a period of time. For this reason, it is desirable to have a container capable of meeting the requirements for storage as well as the requirements for transportation. Most of the off-site transportation is envisioned to take place using the TRUPACT-II Type B package, with the Waste Isolation Pilot Plant (WIPP) as the destination. Prior to the development of this new container, the TRUPACT-II had a limit of 325 FGE (fissile gram equivalents) of plutonium due to criticality control concerns. Because of the relatively high plutonium content in the material to be transported, transporting 325 FGE per TRUPACT-II is uneconomical. Thus, the purpose of the new containers is to provide criticality control to increase the allowed TRUPACT-II payload and to provide a safe method for on-site storage prior to transport. This paper will describe the analysis and testing used to demonstrate that the Pipe Overpack Container provides safe on-site storage of plutonium bearing materials in unhardened buildings and provides criticality control during transportation within the TRUPACT-II. Analyses included worst-case criticality analyses, analyses of fork-lift time impacts, and analyses of roof structure collapse onto the container. Testing included dynamic crush tests, bare pipe impact tests, a 30-minute totally engulfing pool-fire test, and multiple package impact tests in end-on and side-on orientations

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  10. Modeling of the solution interaction properties of plastic materials used in pharmaceutical product container systems.

    Science.gov (United States)

    Jenke, Dennis; Couch, Tom; Gillum, Amy; Sadain, Salma

    2009-01-01

    Material/water equilibrium binding constants (Eb) were determined for 14 organic solutes and 17 plastic raw materials that could be used in pharmaceutical product container systems. Correlations between the measured binding constants and the organic solute's octanol/water and hexane/water partition coefficients were obtained. In general, while the materials examined exhibited a wide range of binding characteristics, the tested materials by and large fell within two broad classes: (1) those that were octanol-like in their binding characteristics, and (2) those that were hexane-like. Materials of the same class (e.g., polypropylenes) generally had binding models that were very similar. Rank ordering of the materials in terms of their magnitude of drug binding (least binding to most binding) was as follows: polypropylene < polyethylene < polyamide < styrene-ethylene-butylene-styrene < copolyester ether elastomer approximately equal to amine-terminated poly fatty acid amide polymer. The utilization of the developed models to estimate drug loss via sorption by the container is discussed.

  11. Preliminary analysis of the creep behaviour of nuclear fuel-waste container materials

    International Nuclear Information System (INIS)

    Dutton, R.; Leitch, B.W.; Crosthwaite, J.L.; Kasprick, G.R.

    1996-12-01

    In the Canadian Nuclear Fuel Waste Management Program, it is proposed that nuclear fuel waste be placed in a durable container and disposed of in a deep underground vault. Consideration of various disposal-container designs has identified either titanium or copper as the material suitable for constructing the container shell. As part of the R and D program to examine the structural integrity of the container, creep tests are being conducted on commercially pure titanium and oxygen-free copper. This report presents the preliminary data obtained. It also describes the evaluation of various constitutive equations to represent the creep curves, thus providing the basis for extrapolation of the creep behaviour over the design lifetime of the container. In this regard, a specific focus is placed on equations derived from the 0-Projection Concept. Recognizing that the container lifetime will be determined by the onset of tertiary creep leading to creep rupture, we present the results of the metallographic examination of creep damage. This shows that the tertiary stage in titanium is associated with the formation of transgranular cavities within the region of localized necking of the creep specimens. In contrast, creep damage in copper is in the form of intergranular cavities uniformly distributed throughout the gauge length. These results are analyzed within the context of the extant literature, and their implications for future container design are discussed. (author)

  12. Ultrasonic Fingerprinting of Structural Materials: Spent Nuclear Fuel Containers Case-Study

    Science.gov (United States)

    Sednev, D.; Lider, A.; Demyanuk, D.; Kroening, M.; Salchak, Y.

    Nowadays, NDT is mainly focused on safety purposes, but it seems possible to apply those methods to provide national and IAEA safeguards. The containment of spent fuel in storage casks could be dramatically improved in case of development of so-called "smart" spent fuel storage and transfer casks. Such casks would have tamper indicating and monitoring/tracking features integrated directly into the cask design. The microstructure of the containers material as well as of the dedicated weld seam is applied to the lid and the cask body and provides a unique fingerprint of the full container, which can be reproducibly scanned by using an appropriate technique. The echo-sounder technique, which is the most commonly used method for material inspection, was chosen for this project. The main measuring parameter is acoustic noise, reflected from material's artefacts. The purpose is to obtain structural fingerprinting. Reference measurement and additional measurement results were compared. Obtained results have verified the appliance of structural fingerprint and the chosen control method. The successful authentication demonstrates the levels of the feature points' compliance exceeding the given threshold which differs considerably from the percentage of the concurrent points during authentication from other points. Since reproduction or doubling of the proposed unique identification characteristics is impossible at the current state science and technology, application of this technique is considered to identify the interference into the nuclear materials displacement with high accuracy.

  13. Container material for the disposal of highly radioactive wastes: corrosion chemistry aspects

    International Nuclear Information System (INIS)

    Grauer, R.

    1984-08-01

    Prior to disposal in crystalline formations it is planned to enclose vitrified highly radioactive waste from nuclear power plants in metallic containers ensuring their isolation from the groundwater for at least 1,000 years. Appropriate metals can be either thermodynamically stable in the repository environment (such as copper), passive materials with very low corrosion rates (titanium, nickel alloys), or metals such as cast iron or unalloyed cast steels which, although they corrode, can be used in sections thick enough to allow for this corrosion. The first part of the report presents the essentials of corrosion science in order to enable even a non-specialist to follow the considerations and arguments necessary to choose the material and design the container against corrosion. Following this, the principles of the long-term extrapolation of corrosion behaviour are discussed. The second part summarizes and comments upon the literature search carried out to identify published results relevant to corrosion in a repository environment. Results of archeaological studies are included wherever possible. Not only the general corrosion behaviour but also localized corrosion and stress corrosion cracking are considered, and the influence of hydrogen on the material behaviour is discussed. Taking the corrosion behaviour as criterion, the author suggests the use either of copper or of cast iron or steel as an appropriate container material. The report concludes with proposals for further studies. (Auth.)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  16. Discrimination of high-Z materials in concrete-filled containers using muon scattering tomography

    Science.gov (United States)

    Frazão, L.; Velthuis, J.; Thomay, C.; Steer, C.

    2016-07-01

    An analysis method of identifying materials using muon scattering tomography is presented, which uses previous knowledge of the position of high-Z objects inside a container and distinguishes them from similar materials. In particular, simulations were performed in order to distinguish a block of Uranium from blocks of Lead and Tungsten of the same size, inside a concrete-filled drum. The results show that, knowing the shape and position from previous analysis, it is possible to distinguish 5 × 5 × 5 cm3 blocks of these materials with about 4h of muon exposure, down to 2 × 2 × 2 cm3 blocks with 70h of data using multivariate analysis (MVA). MVA uses several variables, but it does not benefit the discrimination over a simpler method using only the scatter angles. This indicates that the majority of discrimination is provided by the angular information. Momentum information is shown to provide no benefits in material discrimination.

  17. Thermodynamics of a natural draught-cooled container deposit for radioactive materials

    International Nuclear Information System (INIS)

    Schoenfeld, R.; Dorst, H.J.

    1982-01-01

    This article interprets a calculation model for the determination of heat dissipation and temperature distribution of cooling air within an intermediate container deposit cooled with air by natural draught for self-heating materials. The mathematical modelling of the physical processes in the flowing medium 'air' takes place stationarily and one-dimensionelly in flow direction. The calculation model is based on known procedures for the convective heat transmission, for the description of channel flow, and for the determination of pressure losses. It provides the thermodynamic characteristic data which are required for an engineering and safety design of a container deposit. (orig.) [de

  18. Physicochemical and bioactive properties of innovative resin-based materials containing functional halloysite-nanotubes fillers.

    Science.gov (United States)

    Degrazia, Felipe Weidenbach; Leitune, Vicente Castelo Branco; Takimi, Antonio Shigueaki; Collares, Fabrício Mezzomo; Sauro, Salvatore

    2016-09-01

    This study aimed to assess the degree of conversion, microhardness, solvent degradation, contact angle, surface free energy and bioactivity (e.g., mineral precipitation) of experimental resin-based materials containing, pure or triclosan-encapsulated, aluminosilicate-(halloysite) nanotubes. An experimental resin blend was prepared using bis-GMA/TEGDMA, 75/25wt% (control). Halloysite nanotubes (HNT) doped with or without triclosan (TCN) were first analyzed using transmission electron microscopy (TEM). HNT or HNT/TCN fillers were incorporated into the resin blend at different concentrations (5, 10, and 20wt%). Seven experimental resins were created and the degree of conversion, microhardness, solvent degradation and contact angle were assessed. Bioactive mineral precipitation induced by the experimental resins was evaluated through Raman spectroscopy and SEM-EDX. TEM showed a clear presence of TCN particles inside the tubular lumen and along the outer surfaces of the halloysite nanotubes. The degree of conversion, surface free energy, microhardness, and mineral deposition of polymers increased with higher amount of HNTs. Conversely, the higher the amount (20wt%) of TCN-loaded HNTs the lower the microhardness of the experimental resins. The incorporation of pure or TCN-loaded aluminosilicate-(halloysite) nanotubes into resin-based materials increase the bioactivity of such experimental restorative materials and promotes mineral deposition. Therefore, innovative resin-based materials containing functional halloysite-nanotube fillers may represent a valuable alternative for therapeutic minimally invasive treatments. Copyright © 2016 The Academy of Dental Materials. Published by Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  20. Micromechanical Analysis of Crack Closure Mechanism for Intelligent Material Containing TiNi Fibers

    Science.gov (United States)

    Araki, Shigetoshi; Ono, Hiroyuki; Saito, Kenji

    In our previous study, the micromechanical modeling of an intelligent material containing TiNi fibers was performed and the stress intensity factor KI at the tip of the crack in the material was expressed in terms of the magnitude of the shape memory shrinkage of the fibers and the thermal expansion strain in the material. In this study, the value of KI at the tip of the crack in the TiNi/epoxy material is calculated numerically by using analytical expressions obtained in our first report. As a result, we find that the KI value decreases with increasing shrink strain of the fibers, and this tendency agrees with that of the experimental result obtained by Shimamoto etal.(Trans. Jpn. Soc. Mech. Eng., Vol. 65, No. 634 (1999), pp. 1282-1286). Moreover, there exists an optimal value of the shrink strain of the fibers to make the KI value zero. The change in KI with temperature during the heating process from the reference temperature to the inverse austenitic finishing temperature of TiNi fiber is also consistent with the experimental result. These results can be explained by the changes in the shrink strain, the thermal expansion strain, and the elastic moduli of TiNi fiber with temperature. These results may be useful in designing intelligent materials containing TiNi fibers from the viewpoint of crack closure.

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

    Science.gov (United States)

    Miles, A. M.

    1982-01-01

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

  2. The possibilities of the microwave utilization of wastes on the example of materials containing the asbestos

    Directory of Open Access Journals (Sweden)

    M. Pigiel

    2010-04-01

    Full Text Available The presented paper introduce some of the results of the investigations in the utilization of the materials containing asbestos in the existingin Wroclaw University of Technology Institute’s of Technology of Machines and the Automation Foundry and Automation Group themicrowave reactor. In the reactor’s heating chamber there is possible to recycle from 3 up to 5 kg of the batch at once. The temperaturewith which is possible to receive in it is approx. 1400 oC. The time of it’s achievement (in dependence from utilized material can take outfrom 25 up to 40 minutes.

  3. Stress corrosion cracking of the tubing materials for nuclear steam generators in an environment containing lead

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Kim, Uh Chul; Lee, Eun Hee; Hwang, Seong Sik

    2004-01-01

    Steam generator tube materials show a high susceptibility to stress corrosion cracking (SCC) in an environment containing lead species and some nuclear power plants currently have degradation problems associated with lead-induced stress corrosion cracking in a caustic solution. Effects of an applied potential on SCC is tested for middle-annealed Alloy 600 specimens since their corrosion potential can be changed when lead oxide coexists with other oxidizing species like copper oxide in the sludge. In addition, all the steam generator tubing materials used for nuclear power plants being operated and currently under construction in Korea are tested in a caustic solution with lead oxide. (author)

  4. Nupack, the new ASME code for radioactive material transportation packaging containments

    International Nuclear Information System (INIS)

    Turula, P.

    1998-01-01

    The American Society of Mechanical Engineers (ASME) has added a new division to the nuclear construction section of its Boiler and Pressure Vessel Code (B and PVC). This Division, commonly referred to as Nupack, has been written to provide a consistent set of technical requirements for containment vessels of transportation packagings for high-level radioactive materials. This paper provides an introduction to Nupack, discusses some of its technical provisions, and describes how it can be used for the design and construction of packaging components. Nupack's general provisions and design requirements are emphasized, while treatment of materials, fabrication and inspection is left for another paper

  5. Study of zwitterionic sulfopropylbetaine containing reactive siloxanes for application in antibacterial materials.

    Science.gov (United States)

    Chen, Shiguo; Chen, Shaojun; Jiang, Song; Mo, Yangmiao; Luo, Junxuan; Tang, Jiaoning; Ge, Zaochuan

    2011-07-01

    Antibacterial agents receive a great deal of attention around the world due to the interesting academic problems of how to combat bacteria and of the beneficial health, social and economic effects of successful agents. Scientists are actively developing new antibacterial agents for biomaterial applications. This paper reports the novel antibacterial agent siloxane sulfopropylbetaine (SSPB), which contains reactive alkoxysilane groups. The structure and properties of SSPB were systematically investigated, with the results showing that SSPB contains both quaternary ammonium compounds and reactive siloxane groups. SSPB has good antibacterial activity against both Escherichia coli (E. coli, 8099) and Staphylococcus aureus (S. aureus, ATCC 6538). The minimal inhibition concentration is 70 μmol/ml SSPB against both E. coli and S. aureus. In addition, the SSPB antibacterial agent can be used in both weak acid and weak alkaline environments, functioning within the wide pH range of 4.0-9.0. The SSPB-modified glass surface killed 99.96% of both S. aureus and E. coli organisms within 24 h. No significant decrease was observed in this antibacterial activity after 20 washes. Moreover, SSPB does not induce a skin reaction and is nontoxic to animals. Thus, SSPB is an ideal candidate for future applications as a safe, environmentally friendly antibacterial agent. Copyright © 2011 Elsevier B.V. All rights reserved.

  6. Device for manufacturing methane or synthetic gas from materials containing carbon using a nuclear reactor

    International Nuclear Information System (INIS)

    Jaeger, W.

    1984-01-01

    This invention concerns a device for manufacturing methane or synthetic gas from materials containing carbon using a nuclear reactor, where part of the carbon is gasified with hydration and the remaining carbon is converted to synthetic gas by adding steam. This synthetic gas consists mainly of H 2 , CO, CO 2 and CH 4 and can be converted to methane in so-called methanising using a nickel catalyst. The hydrogen gasifier is situated in the first of two helium circuits of a high temperature reactor, and the splitting furnace is situated in the second helium circuit, where part of the methane produced is split into hydrogen at high temperature, which is used for the hydrating splitting of another part of the material containing carbon. (orig./RB) [de

  7. Anthropogenic materials and products containing natural radionuclides. Pt. 1. Survey of the major exposure pathways

    International Nuclear Information System (INIS)

    Becker, D.E.; Reichelt, A.

    1991-06-01

    Knowledge of the possible exposure pathways permits to perform an overall assessment of the radiation doses and qualities affecting the population, as well as their inter-relations: A catalogue was established of products, raw materials and waste materials containing natural radioactivity that are processed, produced or dumped in Bavaria and that contribute above negligible level to the radiation exposure of the population and to occupational radiation doses. A literature study rounds up the information on anthropogenic sources containing natural radioactivity and thus representing a radiation source generally to be considered for assessments. Some of these sources are discussed in more detail, indicating their radiological significance for the population and the environment in Bavaria. (Orig./DG) [de

  8. Hybrid silica luminescent materials based on lanthanide-containing lyotropic liquid crystal with polarized emission

    Energy Technology Data Exchange (ETDEWEB)

    Selivanova, N.M., E-mail: natsel@mail.ru [Kazan National Research Technological University, 68 Karl Marx Str., Kazan 420015 (Russian Federation); Vandyukov, A.E.; Gubaidullin, A.T. [A.E. Arbuzov Institute of Organic and Physical Chemistry of the Kazan Scientific Center of the Russian Academy of Sciences, 8 Acad. Arbuzov Str., Kazan 420088 (Russian Federation); Galyametdinov, Y.G. [Kazan National Research Technological University, 68 Karl Marx Str., Kazan 420015 (Russian Federation)

    2014-11-14

    This paper represents the template method for synthesis of hybrid silica films based on Ln-containing lyotropic liquid crystal and characterized by efficient luminescence. Luminescence films were prepared in situ by the sol–gel processes. Lyotropic liquid crystal (LLC) mesophases C{sub 12}H{sub 25}O(CH{sub 2}CH{sub 2}O){sub 10}H/Ln(NO{sub 3}){sub 3}·6H{sub 2}O/H{sub 2}O containing Ln (III) ions (Dy, Tb, Eu) were used as template. Polarized optical microscopy, X-ray powder diffraction, and FT-IR-spectroscopy were used for characterization of liquid crystal mesophases and hybrid films. The morphology of composite films was studied by the atomic force microscopy method (AFM). The optical properties of the resulting materials were evaluated. It was found that hybrid silica films demonstrate significant increase of their lifetime in comparison with an LLC system. New effects of linearly polarized emission revealed for Ln-containing hybrid silica films. Polarization in lanthanide-containing hybrid composites indicates that silica precursor causes orientation of emitting ions. - Highlights: • We suggest a new simple approach for creating luminescence hybrid silica films. • Ln-containing hybrid silica films demonstrate yellow, green and red emissions. • Tb(III)-containing hybrid film have a high lifetime. • We report effects of linearly polarized emission in hybrid film.

  9. Evaluation of copper, aluminum bronze, and copper-nickel container material for the Yucca mountain project

    International Nuclear Information System (INIS)

    Kass, J.

    1990-01-01

    Copper, 70 percent aluminum bronze, and 70/30 copper-nickel were evaluated as potential waste-packaging materials as part of the Yucca Mountain Project. The proposed waste repository site is under a desert mountain in southern Nevada. The expected temperatures at the container surface are higher than at other sites, about 250C at the beginning of the containment period; they could fall below the boiling point of water during this period, but will be exposed to very little water, probably less than 5 l/a. Initial gamma flux will be 10 4 rad/h, and no significant hydrostatic or lithostatic pressure is expected. Packages will contain PWR or BWR fuel, or processed-glass waste. Three copper alloys are being considered for containers: oxygen-free copper (CDA 102); 7 percent aluminum bronze (CDA 613); and 70/30 copper-nickel (CDA 715). Phase separation due to prolonged thermal exposure could be a problem for the two alloys, causing embrittlement. The reduction of internal oxides present in pure copper by hydrogen could cause mechanical degradation. Corrosion and oxidation rates measured for the three materials in well water with and without gamma irradiation at flux rates about ten times higher than those expected were all quite small. The corrosion/oxidation rates for CDA715 show a marked increase under irradiation, but are still acceptable. In the presence of ammonia and other nitrogen-bearing species stress corrosion cracking (SCC) is a concern. Welded U-bend specimens of all three materials have been tested for up to 10000 h in highly irradiated environments, showing no SCC. There was some alloy segregation in the Al bronze specimens. The investigators believe that corrosion and mechanical properties will not present problems for these materials at this site. Further work is needed in the areas of weld inspection, welding techniques, embrittlement of weld metal, the effects of dropping the containers during emplacement, and stress corrosion cracking. Other materials

  10. Gas-handling system for studies of tritium-containing materials

    International Nuclear Information System (INIS)

    Carstens, D.H.W.

    1975-01-01

    A gas handling system for preparation and study of tritium containing compounds and materials is described. The system at any one time can handle amounts of DT gas up to about 3 moles and has provisions for purification, storage, and measurement of the gas. Experimental conditions covering the ranges 20 to 800 0 C and 0.1 Pa to 137 MPa (10 -2 torr to 20,000 psi) can be maintained. (auth)

  11. Tunable multichannel filter in photonic crystal heterostructure containing permeability-negative materials

    International Nuclear Information System (INIS)

    Hu Xiaoyong; Liu Zheng; Gong Qihuang

    2008-01-01

    A tunable multichannel filter is demonstrated theoretically based on a one-dimensional photonic crystal heterostructure containing permeability-negative material. The filtering properties of the photonic crystal filter, including the channel number and frequency, can be tuned by adjusting the structure parameters or by a pump laser. The angular response of the photonic crystal filter and the influences of the losses on the filtering properties are also analyzed

  12. Tunable multichannel filter in photonic crystal heterostructure containing permeability-negative materials

    Energy Technology Data Exchange (ETDEWEB)

    Hu Xiaoyong [State Key Laboratory for Mesoscopic Physics, Department of Physics, Peking University, Beijing 100871 (China)], E-mail: xiaoyonghu@pku.edu.cn; Liu Zheng [State Key Laboratory for Mesoscopic Physics, Department of Physics, Peking University, Beijing 100871 (China); Gong Qihuang [State Key Laboratory for Mesoscopic Physics, Department of Physics, Peking University, Beijing 100871 (China)], E-mail: qhgong@pku.edu.cn

    2008-01-14

    A tunable multichannel filter is demonstrated theoretically based on a one-dimensional photonic crystal heterostructure containing permeability-negative material. The filtering properties of the photonic crystal filter, including the channel number and frequency, can be tuned by adjusting the structure parameters or by a pump laser. The angular response of the photonic crystal filter and the influences of the losses on the filtering properties are also analyzed.

  13. Spectral studies on the interaction of acetylacetone with aluminum-containing MCM-41 mesoporous materials

    International Nuclear Information System (INIS)

    Zanjanchi, M.A.; Vaziri, M.

    2008-01-01

    Diffuse reflectance spectroscopy (DRS) was used to study the interaction of acetylacetone (acac) with the mesoporous aluminum-containing MCM-41 materials. A room temperature synthesis method was used for preparation of purely siliceous MCM-41 and for aluminum-containing MCM-41 materials. Samples with Si/Al ratios of 50, 20, 10 and 5 were synthesized. The synthesized mesoporous materials possess highly ordered structure and high surface area as evidenced from X-ray diffraction and nitrogen physisorption measurements, respectively. The treatment of the as-synthesized aluminum-containing MCM-41 samples with acac shows a distinct band at ∼290 nm. This band is assigned to six coordinated aluminum atoms in the structure which is produced by diffusion of acac molecules through surfactant micelles and their interaction with aluminum atoms. The 290-nm band disappears upon several successive washing of the sample with ethanol. The treatment of the calcined aluminum-containing MCM-41 sample with acac produces the same 290-nm band where its intensity increases with the aluminum content of the sample. The intensity of this band is reduced upon successive ethanol washing, but remains nearly constant after three times washing. This irremovable aluminum species can be assigned to framework aluminum. The measured acidity for our aluminum-containing MCM-41 samples correlates linearly with the intensity of 290-nm band for the ethanol treated samples. This supports the idea that the Bronsted acidity in aluminum-modified MCM-41 samples is a function of the amount of tetrahedral framework aluminum in the structure

  14. Direct containment heating experiments in Zion Nuclear Power Plant geometry using prototypic materials

    International Nuclear Information System (INIS)

    Binder, J.L.; McUmber, L.M.; Spencer, B.W.

    1993-01-01

    Direct Containment Heating (DCH) experiments have been completed which utilize prototypic core materials. The experiments reported on here are a continuation of the Integral Effects Testing (IET) DCH program. The experiments incorporated a 1/40 scale model of the Zion Nuclear Power Plant containment structures. The model included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven. Iron-alumina thermite with chromium was used as a core melt stimulant in the earlier IET experiments. These earlier IET experiments at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL) provided useful data on the effect of scale on DCH phenomena; however, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. Three tests have been completed, DCH-U1A, U1B and U2. DCH-U1A and U1B employed an inerted containment atmosphere and are counterpart to the IET-1RR test with iron/alumina thermite. DCH-U2 employed nominally the same atmosphere composition of its counterpart iron/alumina test, IET-6. All tests, with prototypic material, have produced lower peak containment pressure rises; 45, 111 and 185 kPa in U1A, U1B and U2, compared to 150 and 250 kPa IET-1RR and 6. Hydrogen production, due to metal-steam reactions, was 33% larger in U1B and U2 compared to IET-1RR and IET-6. The pressurization efficiency was consistently lower for the corium tests compared to the IET tests

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

    Energy Technology Data Exchange (ETDEWEB)

    Groot, Joost de

    2012-06-28

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

  16. Improving Performance of Cold-Chain Insulated Container with Phase Change Material: An Experimental Investigation

    Directory of Open Access Journals (Sweden)

    Li Huang

    2017-12-01

    Full Text Available The cold-chain transportation is an important means to ensure the drug and food safety. An cold-chain insulated container incorporating with Phase Change Material (PCM has been developed for a temperature-controlled transportation in the range of 2~8 °C. The container configuration and different preconditioning methods have been determined to realize a 72-h transportation under extremely high, extremely low, and alternating temperature conditions. The experimental results showed that the temperature-controlled time was extended from 1 h to more than 80 h and the internal temperature maintained at 4~5 °C by using a PCM with a melting/freezing point of 5 °C, while the container presented a subcooling effect in a range of −1~2 °C when using water as PCM. The experimental values of the temperature-controlled time agreed well with the theoretical values.

  17. Martensitic/ferritic steels as container materials for liquid mercury target of ESS

    International Nuclear Information System (INIS)

    Dai, Y.

    1996-01-01

    In the previous report, the suitability of steels as the ESS liquid mercury target container material was discussed on the basis of the existing database on conventional austenitic and martensitic/ferritic steels, especially on their representatives, solution annealed 316 stainless steel (SA 316) and Sandvik HT-9 martensitic steel (HT-9). Compared to solution annealed austenitic stainless steels, martensitic/ferritic steels have superior properties in terms of strength, thermal conductivity, thermal expansion, mercury corrosion resistance, void swelling and irradiation creep resistance. The main limitation for conventional martensitic/ferritic steels (CMFS) is embrittlement after low temperature (≤380 degrees C) irradiation. The ductile-brittle transition temperature (DBTT) can increase as much as 250 to 300 degrees C and the upper-shelf energy (USE), at the same time, reduce more than 50%. This makes the application temperature range of CMFS is likely between 300 degrees C to 500 degrees C. For the present target design concept, the temperature at the container will be likely controlled in a temperature range between 180 degrees C to 330 degrees C. Hence, CMFS seem to be difficult to apply. However, solution annealed austenitic stainless steels are also difficult to apply as the maximum stress level at the container will be higher than the design stress. The solution to the problem is very likely to use advanced low-activation martensitic/ferritic steels (LAMS) developed by the fusion materials community though the present database on the materials is still very limited

  18. Stress corrosion cracking of steam generator tubing materials in lead containing solution

    International Nuclear Information System (INIS)

    Kim, H.P.; Hwang, S.S.; Kim, J.S.; Hong, J.H.

    2007-01-01

    Stress corrosion cracking (SCC) in lead (Pb) containing environments has been one of key issues in the nuclear power industry since Pb had been identified as a cause of the SCC of steam generator (SG) tubing materials in some power plants. To mitigate or prevent degradation of SG tubing materials, a mechanistic understanding of SCC in Pb containing environment is needed, along with an understanding of the source and transport behaviors of Pb species in the secondary circuit. In this work, SCC behaviors of Alloy 600 in Pb containing environments were studied. Influences of microstructures of Alloy 600 and the inhibitive additives were investigated using the C-ring and the slow strain rate tests in caustic solution and demineralized water at 315 o C. Microstructures of Alloy 600 were varied by heat treatment at different temperatures. The additives examined were nickel boride (NiB) and cerium boride (CeB 6 ). The surface films were analyzed using Auger Electron Spectroscopy (AES) and Energy Dispersive X-ray Spectroscopy (EDS). The SCC mode varied with microstructure. Effectiveness of the additives in Pb containing environments is discussed. (author)

  19. Experimental studies of dynamic impact response with scale models of lead shielded radioactive material shipping containers

    International Nuclear Information System (INIS)

    Robinson, R.A.; Hadden, J.A.; Basham, S.J.

    1978-01-01

    Preliminary experimental studies of dynamic impact response of scale models of lead-shielded radioactive material shipping containers are presented. The objective of these studies is to provide DOE/ECT with a data base to allow the prediction of a rational margin of confidence in overviewing and assessing the adequacy of the safety and environmental control provided by these shipping containers. Replica scale modeling techniques were employed to predict full scale response with 1/8, 1/4, and 1/2 scale models of shipping containers that are used in the shipment of spent nuclear fuel and high level wastes. Free fall impact experiments are described for scale models of plain cylindrical stainless steel shells, stainless steel shells filled with lead, and replica scale models of radioactive material shipping containers. Dynamic induced strain and acceleration measurements were obtained at several critical locations on the models. The models were dropped from various heights, attitudes to the impact surface, with and without impact limiters and at uniform temperatures between -40 and 175 0 C. In addition, thermal expansion and thermal gradient induced strains were measured at -40 and 175 0 C. The frequency content of the strain signals and the effect of different drop pad compositions and stiffness were examined. Appropriate scale modeling laws were developed and scaling techniques were substantiated for predicting full scale response by comparison of dynamic strain data for 1/8, 1/4, and 1/2 scale models with stainless steel shells and lead shielding

  20. Tension in the tie-down chains of a shipping container for hazardous material

    International Nuclear Information System (INIS)

    Lo, T.Y.

    1995-01-01

    Chains are frequently used to tie the shipping containers of hazardous material to a truck bed. The tie-down system is nonlinear when the container is subjected to a triaxial force during transit. It is nonlinear because chains cannot carry compressive force, and the base of the container may partially lift off from the truck bed. A method was developed to calculate the amount of tension in the chains. This methodology includes three assumptions: (1) No friction exists between the container and the truck bed; (2) The container and the truck bed are rigid; and (3) All chains are properly tightened (i.e., no slacks) during preparation for shipment. The methodology employs an iterative process of a linear tie-down system. This linear system is derived from the nonlinear system with two additional assumptions: (a) All chains can carry compression as well as tension; and (b) There is a point contact between the container and the truck bed. This linear system has a closed-form solution. After the first solution of the linear system is obtained, the unreasonable, or physically impossible, rotational degree of freedom of the container and the chains with compressive force are eliminated in the follow-up iterative calculations using the same linear system. Unreasonable rotation is detected when the results indicate a rotation of the container into the truck bed. Zero rotational value is assigned to this degree of freedom in the follow-up iterations. For chains under compression, the chain length is replaced with a fictitiously large value because a long chain carries essentially no load. This process usually needs only two or three iterations and can easily be carried out in a spread sheet using such programs as Microsoft Excel or LOTUS 123

  1. Device for Detection of Explosives, Nuclear and Other Hazardous Materials in Luggage and Cargo Containers

    Science.gov (United States)

    Kuznetsov, Andrey; Evsenin, Alexey; Gorshkov, Igor; Osetrov, Oleg; Vakhtin, Dmitry

    2009-12-01

    Device for detection of explosives, radioactive and heavily shielded nuclear materials in luggage and cargo containers based on Nanosecond Neutron Analysis/Associated Particles Technique (NNA/APT) is under construction. Detection module consists of a small neutron generator with built-in position-sensitive detector of associated alpha-particles, and several scintillator-based gamma-ray detectors. Explosives and other hazardous chemicals are detected by analyzing secondary high-energy gamma-rays from reactions of fast neutrons with materials inside a container. The same gamma-ray detectors are used to detect unshielded radioactive and nuclear materials. An array of several neutron detectors is used to detect fast neutrons from induced fission of nuclear materials. Coincidence and timing analysis allows one to discriminate between fission neutrons and scattered probing neutrons. Mathematical modeling by MCNP5 and MCNP-PoliMi codes was used to estimate the sensitivity of the device and its optimal configuration. Comparison of the features of three gamma detector types—based on BGO, NaI and LaBr3 crystals is presented.

  2. Device for Detection of Explosives, Nuclear and Other Hazardous Materials in Luggage and Cargo Containers

    International Nuclear Information System (INIS)

    Kuznetsov, Andrey; Evsenin, Alexey; Osetrov, Oleg; Vakhtin, Dmitry; Gorshkov, Igor

    2009-01-01

    Device for detection of explosives, radioactive and heavily shielded nuclear materials in luggage and cargo containers based on Nanosecond Neutron Analysis/Associated Particles Technique (NNA/APT) is under construction. Detection module consists of a small neutron generator with built-in position-sensitive detector of associated alpha-particles, and several scintillator-based gamma-ray detectors. Explosives and other hazardous chemicals are detected by analyzing secondary high-energy gamma-rays from reactions of fast neutrons with materials inside a container. The same gamma-ray detectors are used to detect unshielded radioactive and nuclear materials. An array of several neutron detectors is used to detect fast neutrons from induced fission of nuclear materials. Coincidence and timing analysis allows one to discriminate between fission neutrons and scattered probing neutrons. Mathematical modeling by MCNP5 and MCNP-PoliMi codes was used to estimate the sensitivity of the device and its optimal configuration. Comparison of the features of three gamma detector types--based on BGO, NaI and LaBr 3 crystals is presented.

  3. A method to calculate flux distribution in reactor systems containing materials with grain structure

    International Nuclear Information System (INIS)

    Stepanek, J.

    1980-01-01

    A method is proposed to compute the neutron flux spatial distribution in slab, spherical or cylindrical systems containing zones with close grain structure of material. Several different types of equally distributed particles embedded in the matrix material are allowed in one or more zones. The multi-energy group structure of the flux is considered. The collision probability method is used to compute the fluxes in the grains and in an ''effective'' part of the matrix material. Then the overall structure of the flux distribution in the zones with homogenized materials is determined using the DPN ''surface flux'' method. Both computations are connected using the balance equation during the outer iterations. The proposed method is written in the code SURCU-DH. Two testcases are computed and discussed. One testcase is the computation of the eigenvalue in simplified slab geometry of an LWR container of one zone with boral grains equally distributed in an aluminium matrix. The second is the computation of the eigenvalue in spherical geometry of the HTR pebble-bed cell with spherical particles embedded in a graphite matrix. The results are compared to those obtained by repeated use of the WIMS Code. (author)

  4. Hierarchy concepts: classification and preparation strategies for zeolite containing materials with hierarchical porosity.

    Science.gov (United States)

    Schwieger, Wilhelm; Machoke, Albert Gonche; Weissenberger, Tobias; Inayat, Amer; Selvam, Thangaraj; Klumpp, Michael; Inayat, Alexandra

    2016-06-13

    'Hierarchy' is a property which can be attributed to a manifold of different immaterial systems, such as ideas, items and organisations or material ones like biological systems within living organisms or artificial, man-made constructions. The property 'hierarchy' is mainly characterised by a certain ordering of individual elements relative to each other, often in combination with a certain degree of branching. Especially mass-flow related systems in the natural environment feature special hierarchically branched patterns. This review is a survey into the world of hierarchical systems with special focus on hierarchically porous zeolite materials. A classification of hierarchical porosity is proposed based on the flow distribution pattern within the respective pore systems. In addition, this review might serve as a toolbox providing several synthetic and post-synthetic strategies to prepare zeolitic or zeolite containing material with tailored hierarchical porosity. Very often, such strategies with their underlying principles were developed for improving the performance of the final materials in different technical applications like adsorptive or catalytic processes. In the present review, besides on the hierarchically porous all-zeolite material, special focus is laid on the preparation of zeolitic composite materials with hierarchical porosity capable to face the demands of industrial application.

  5. An examination of source material requirements contained in 10 CFR Part 40

    International Nuclear Information System (INIS)

    Nussbaumer, D.; Smith, D.A.; Wiblin, C.

    1992-10-01

    This report identifies issues for consideration for rule-making to update the requirements for source material in 10 CFR Part 40 and examines options for resolving these issues. The contemplated rulemaking is intended to update 10 CFR Part 40 to reflect current radiation protection principles and regulatory practices. It is expected that such an update would make requirements for the control of source material more comparable to those pertaining to byproduct material contained in 10 CFR Part 30. The newer biological data and dose calculation methodology reflected in revised 10 CFR Part 20 will be used in analyses of potential regulatory amendments. This report presents historical background information and discussion on the various issues identified and makes preliminary recommendations concerning needed regulatory changes and approaches to rulemaking

  6. Research and development activities at INE concerning corrosion of final repository container materials

    International Nuclear Information System (INIS)

    Kienzler, Bernhard

    2017-01-01

    The present work provides a historical overview of the research and development activities carried out at the (Nuclear) Research Center Karlsruhe (today KIT) since the beginning of the 1980s on the corrosion of materials which might be suitable for construction of containers for highly radioactive wastes. The report relates almost exclusively to the work performed by Dr. Emmanuel Smailos, who elaborated the corrosion of various materials at the Institute for Nuclear Waste Disposal (INE). The requirements for the containers and materials, which were subject to changes in time, are presented. The changes were strongly influenced by the changed perception of the use of nuclear energy. The selection of the materials under investigations, the boundary conditions for the corrosion experiments and the analytical methods are described. Results of the corrosion of the materials such as finegrained steel, Hastelloy C4, nodular cast iron, titanium-palladium and copper or copper-nickel alloys in typical salt solutions are summarized. The findings of special investigations, e.g. corrosion under irradiation or the influence of sulfide on the corrosion rates are shown. For construction of disposal canisters, experiments were conducted to determine the contact corrosion, the influence of the hydrogen embrittlement of Ti-Pd and fine-grained steels on the corrosion behavior as well as the corrosion behavior of welding and the influence of different welding processes with the resulting heat-affected zones on the corrosion behavior. The work was contributed to several European research programs and was well recognized in the USA. Investigations on the corrosion of steels in non-saline solutions and corrosion under interim storage conditions as well as under the expected conditions of the Konrad repository for low-level radioactive wastes are also described. In addition, the experiments on ceramic materials are presented and the results of the corrosion of Al 2 O 3 and ZrO 2 ceramics

  7. Los Alamos National Laboratory new generation standard nuclear material storage container - the SAVY4000 design

    International Nuclear Information System (INIS)

    Stone, Timothy Amos

    2010-01-01

    Incidents involving release of nuclear materials stored in containers of convenience such as food pack cans, slip lid taped cans, paint cans, etc. has resulted in defense board concerns over the lack of prescriptive performance requirements for interim storage of nuclear materials. Los Alamos National Laboratory (LANL) has shared in these incidents and in response proactively moved into developing a performance based standard involving storage of nuclear material (RD003). This RD003 requirements document has sense been updated to reflect requirements as identified with recently issued DOE M 441.1-1 'Nuclear Material Packaging Manual'. The new packaging manual was issued at the encouragement of the Defense Nuclear Facilities Safety Board with a clear directive for protecting the worker from exposure due to loss of containment of stored materials. The Manual specifies a detailed and all inclusive approach to achieve a high level of protection; from package design and performance requirements, design life determinations of limited life components, authorized contents evaluations, and surveillance/maintenance to ensure in use package integrity over time. Materials in scope involve those stored outside an approved engineered-contamination barrier that would result in a worker exposure of in excess of 5 rem Committed Effective Does Equivalent (CEDE). Key aspects of meeting the challenge as developed around the SAVY-3000 vented storage container design will be discussed. Design performance and acceptance criteria against the manual, bounding conditions as established that the user must ensure are met to authorize contents in the package (based upon the activity of heat-source plutonium (90% Pu-238) oxide, which bounds the requirements for weapons-grade plutonium oxide), interface as a safety class system within the facility under the LANL plutonium facility DSA, design life determinations for limited life components, and a sense of design specific surveillance program

  8. 41 CFR 101-42.1101 - Federal supply classification (FSC) groups and classes which contain hazardous materials.

    Science.gov (United States)

    2010-07-01

    ..., building paper, and thermal insulation materials Asbestos cloth which has loose fibers or particles that... shipment. 5965 Headsets, handsets, microphones, and speakers Items containing magnetic material. 5970... hazardous chemicals, solvents. 6625 Electrical and electronic properties measuring and testing instruments...

  9. Simulation and Experimental Validation of Electromagnetic Signatures for Monitoring of Nuclear Material Storage Containers

    International Nuclear Information System (INIS)

    Aker, Pamela M.; Bunch, Kyle J.; Jones, Anthony M.

    2013-01-01

    Previous research at the Pacific Northwest National Laboratory (PNNL) has demonstrated that the low frequency electromagnetic (EM) response of a sealed metallic container interrogated with an encircling coil is a strong function of its contents and can be used to form a distinct signature which can confirm the presence of specific components without revealing hidden geometry or classified design information. Finite element simulations have recently been performed to further investigate this response for a variety of configurations composed of an encircling coil and a typical nuclear material storage container. Excellent agreement was obtained between simulated and measured impedance signatures of electrically conducting spheres placed inside an AT-400R nuclear container. Simulations were used to determine the effects of excitation frequency and the geometry of the encircling coil, nuclear container, and internal contents. The results show that it is possible to use electromagnetic models to evaluate the application of the EM signature technique to proposed versions of nuclear weapons containers which can accommodate restrictions imposed by international arms control and treaty verification legislation

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

    Science.gov (United States)

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

    2018-06-01

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

  11. Process for surface treatment of zirconium-containing cladding materials for fuel element or other components for nuclear reactors

    International Nuclear Information System (INIS)

    Videm, K.G.; Lunde, L.R.; Kooyman, H.H.

    1975-01-01

    A process for the surface treatment of zirconium-base cladding materials for fuel elements or other components for nuclear reactors is described. The treatment includes pickling the cladding material in a fluoride-containing bath, and then applying a protective coating through oxidation to the pickled cladding material. The fluoride-containing contaminants which remain on the surface of the cladding material during pickling are removed or rendered harmless by anodic oxidation

  12. Radiation Shielding Materials Containing Hydrogen, Boron, and Nitrogen: Systematic Computational and Experimental Study. Phase I

    Science.gov (United States)

    Thibeault, Sheila A.; Fay, Catharine C.; Lowther, Sharon E.; Earle, Kevin D.; Sauti, Godfrey; Kang, Jin Ho; Park, Cheol; McMullen, Amelia M.

    2012-01-01

    The key objectives of this study are to investigate, both computationally and experimentally, which forms, compositions, and layerings of hydrogen, boron, and nitrogen containing materials will offer the greatest shielding in the most structurally robust combination against galactic cosmic radiation (GCR), secondary neutrons, and solar energetic particles (SEP). The objectives and expected significance of this research are to develop a space radiation shielding materials system that has high efficacy for shielding radiation and that also has high strength for load bearing primary structures. Such a materials system does not yet exist. The boron nitride nanotube (BNNT) can theoretically be processed into structural BNNT and used for load bearing structures. Furthermore, the BNNT can be incorporated into high hydrogen polymers and the combination used as matrix reinforcement for structural composites. BNNT's molecular structure is attractive for hydrogen storage and hydrogenation. There are two methods or techniques for introducing hydrogen into BNNT: (1) hydrogen storage in BNNT, and (2) hydrogenation of BNNT (hydrogenated BNNT). In the hydrogen storage method, nanotubes are favored to store hydrogen over particles and sheets because they have much larger surface areas and higher hydrogen binding energy. The carbon nanotube (CNT) and BNNT have been studied as potentially outstanding hydrogen storage materials since 1997. Our study of hydrogen storage in BNNT - as a function of temperature, pressure, and hydrogen gas concentration - will be performed with a hydrogen storage chamber equipped with a hydrogen generator. The second method of introducing hydrogen into BNNT is hydrogenation of BNNT, where hydrogen is covalently bonded onto boron, nitrogen, or both. Hydrogenation of BN and BNNT has been studied theoretically. Hyper-hydrogenated BNNT has been theoretically predicted with hydrogen coverage up to 100% of the individual atoms. This is a higher hydrogen content

  13. Fluoride release and recharge abilities of contemporary fluoride-containing restorative materials and dental adhesives.

    Science.gov (United States)

    Dionysopoulos, Dimitrios; Koliniotou-Koumpia, Eugenia; Helvatzoglou-Antoniades, Maria; Kotsanos, Nikolaos

    2013-01-01

    The aim of this study was to evaluate the fluoride release of five fluoride-releasing restorative materials and three dental adhesives, before and after NaF solution treatment. Five restorative materials (Fuji IX GP, GC Corp.; Ketac N100, 3M ESPE; Dyract Extra, Dentsply; Beautifil II, Shofu Inc.; Wave, SDI) and three dental adhesives (Stae, SDI; Fluorobond II - Shofu Inc.; Prime & Bond NT, Dentsply) were investigated before and after NaF solution treatment. A fluoride ion-selective electrode was to measure fluoride concentrations. During the 86-day period before NaF solution treatment, Fuji IX GP released the highest amount of fluoride among the restorative materials while Prime & Bond NT was the highest among the dental adhesives. After NaF solution treatment, Fuji IX GP again ranked the highest in fluoride release among the restorative materials while Fluorobond II ranked the highest among dental adhesives. It was concluded that the compositions and setting mechanisms of fluoride-containing dental materials influenced their fluoride release and recharge abilities.

  14. Standard practice for sampling special nuclear materials in multi-container lots

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1987-01-01

    1.1 This practice provides an aid in designing a sampling and analysis plan for the purpose of minimizing random error in the measurement of the amount of nuclear material in a lot consisting of several containers. The problem addressed is the selection of the number of containers to be sampled, the number of samples to be taken from each sampled container, and the number of aliquot analyses to be performed on each sample. 1.2 This practice provides examples for application as well as the necessary development for understanding the statistics involved. The uniqueness of most situations does not allow presentation of step-by-step procedures for designing sampling plans. It is recommended that a statistician experienced in materials sampling be consulted when developing such plans. 1.3 The values stated in SI units are to be regarded as the standard. 1.4 This standard does not purport to address all of the safety problems, if any, associated with its use. It is the responsibility of the user of this standar...

  15. Processing method and processing device for liquid waste containing surface active agent and radioactive material

    International Nuclear Information System (INIS)

    Nishi, Takashi; Matsuda, Masami; Baba, Tsutomu; Yoshikawa, Ryozo; Yukita, Atsushi.

    1998-01-01

    Washing liquid wastes containing surface active agents and radioactive materials are sent to a deaerating vessel. Ozone is blown into the deaerating vessel. The washing liquid wastes dissolved with ozone are introduced to a UV ray irradiation vessel. UV rays are irradiated to the washing liquid wastes, and hydroxy radicals generated by photodecomposition of dissolved ozone oxidatively decompose surface active agents contained in the washing liquid wastes. The washing liquid wastes discharged from the UV ray irradiation vessel are sent to an activated carbon mixing vessel and mixed with powdery activated carbon. The surface active agents not decomposed in the UV ray irradiation vessel are adsorbed to the activated carbon. Then, the activated carbon and washing liquid wastes are separated by an activated carbon separating/drying device. Radioactive materials (iron oxide and the like) contained in the washing liquid wastes are mostly granular, and they are separated and removed from the washing liquid wastes in the activated carbon separating/drying device. (I.N.)

  16. Preparation and characterization of macrocapsules containing microencapsulated PCMs (phase change materials) for thermal energy storage

    International Nuclear Information System (INIS)

    Han, Pengju; Lu, Lixin; Qiu, Xiaolin; Tang, Yali; Wang, Jun

    2015-01-01

    This paper was aimed to prepare, characterize and determine the comprehensive evaluation of promising composite macrocapsules containing microencapsulated PCMs (phase change materials) with calcium alginate gels as the matrix material. Macrocapsules containing microcapsules were fabricated by piercing-solidifying incuber method. Two kinds of microcapsules with n-tetradecane as core material, UF (urea-formaldehyde) and PMMA (poly(methyl methacrylate)) respectively as shell materials were prepared initially. For application concerns, thermal durability and mechanical property of macrocapsules were investigated by TGA (thermal gravimetric analysis) and Texture Analyser for the first time, respectively. The results showed excellent thermal stability and the compressive resistance of macrocapsules was sufficient for common application. The morphology and chemical structure of the prepared microcapsules and macrocapsules were characterized by SEM (scanning electron microscopy) and FT-IR (fourier transform infrared) spectroscopy method. Phase change behaviors and thermal durability of microcapsules and macrocapsules were investigated by DSC (differential scanning calorimetry). In order to improve latent heat of composite microcapsules, the core-shell weight ratio of tetradecane/UF shell microcapsules was chosen as 5.5:1 which obtained the phase change enthalpy of 194.1 J g −1 determined by DSC. In conclusion, these properties make it a feasible composite in applications of textile, building and cold-chain transportation. - Highlights: • We improved the phase change enthalpy with a higher core-shell ratio. • Urea-formaldehyde was firstly used as a shell material in the composite. • Mechanical and thermal durability property of the macrocapsules was firstly investigated in our work.

  17. Feasibility assessment of copper-base waste package container materials in a tuff repository

    International Nuclear Information System (INIS)

    Acton, C.F.; McCright, R.D.

    1986-01-01

    This report discussed progress made during the second year of a two-year study on the feasibility of using copper or a copper-base alloy as a container material for a waste package in a potential repository in tuff rock at the Yucca Mountain site in Nevada. Corrosion testing in potentially corrosive irradiated environments received emphasis during the feasibility study. Results of experiments to evaluate the effect of a radiation field on the uniform corrosion rate of the copper-base materials in repository-relevant aqueous environments are given as well as results of an electrochemical study of the copper-base materials in normal and concentrated J-13 water. Results of tests on the irradiation of J-13 water and on the subsequent formation of hydrogen peroxide are given. A theoretical study was initiated to predict the long-term corrosion behavior of copper in the repository. Tests were conducted to determine whether copper would adversely affect release rates of radionuclides to the environment because of degradation of the Zircaloy cladding. A manufacturing survey to determine the feasibility of producing copper containers utilizing existing equipment and processes was completed. The cost and availability of copper was also evaluated and predicted to the year 2000. Results of this feasibility assessment are summarized

  18. Process of preparing ethanol by continuous fermentation of polysaccharide-containing materials

    Energy Technology Data Exchange (ETDEWEB)

    Ehnstroem, L.K.J.

    1981-04-16

    The invention concerns a process of preparing ethanol by continuous fermentation of polysaccharide - containing raw materials. Fermentation, hereby, occurs in one or several fermentors while dividing one stream of the fermentation liquid into a yeast-concentrate stream and a yeast-free stream and, if neccessary, a sludge stream. The yeast-concentrate stream is re-fed into the fermentor and at least part of the yeast-free stream is directed into a simple evaporator corresponding to one or several distilling stages where it is separated partially in an ethanol-enriched initial vapour stream supplying a facility to produce the desired ethanol quality, and partially in a liquid initial bottom stream re-fed at least in part into the fermentor. The characteristic feature of this new process is that a raw-material stream is fed into a closed circuit containing the fermentor and the evaporator, and that, in the evaporator, the raw-material stream is hydrolysed to a fermentable state. This hydrolysis is carried out most favourably by enzymes - preferably a gluco-amylase - at a temperature ranging from 35/sup 0/C to 75/sup 0/C.

  19. Pyrolysis behavior of different type of materials contained in the rejects of packaging waste sorting plants.

    Science.gov (United States)

    Adrados, A; De Marco, I; Lopez-Urionabarrenechea, A; Caballero, B M; Laresgoiti, M F

    2013-01-01

    In this paper rejected streams coming from a waste packaging material recovery facility have been characterized and separated into families of products of similar nature in order to determine the influence of different types of ingredients in the products obtained in the pyrolysis process. The pyrolysis experiments have been carried out in a non-stirred batch 3.5 dm(3) reactor, swept with 1 L min(-1) N(2), at 500°C for 30 min. Pyrolysis liquids are composed of an organic phase and an aqueous phase. The aqueous phase is greater as higher is the cellulosic material content in the sample. The organic phase contains valuable chemicals as styrene, ethylbenzene and toluene, and has high heating value (HHV) (33-40 MJ kg(-1)). Therefore they could be used as alternative fuels for heat and power generation and as a source of valuable chemicals. Pyrolysis gases are mainly composed of hydrocarbons but contain high amounts of CO and CO(2); their HHV is in the range of 18-46 MJ kg(-1). The amount of COCO(2) increases, and consequently HHV decreases as higher is the cellulosic content of the waste. Pyrolysis solids are mainly composed of inorganics and char formed in the process. The cellulosic materials lower the quality of the pyrolysis liquids and gases, and increase the production of char. Copyright © 2012 Elsevier Ltd. All rights reserved.

  20. Morphological changes during enhanced carbonation of asbestos containing material and its comparison to magnesium silicate minerals

    International Nuclear Information System (INIS)

    Gadikota, Greeshma; Natali, Claudio; Boschi, Chiara; Park, Ah-Hyung Alissa

    2014-01-01

    The disintegration of asbestos containing materials (ACM) over time can result in the mobilization of toxic chrysotile ((Mg, Fe) 3 Si 2 O 5 (OH) 4 )) fibers. Therefore, carbonation of these materials can be used to alter the fibrous morphology of asbestos and help mitigate anthropogenic CO 2 emissions, depending on the amount of available alkaline metal in the materials. A series of high pressure carbonation experiments were performed in a batch reactor at P CO2 of 139 atm using solvents containing different ligands (i.e., oxalate and acetate). The results of ACM carbonation were compared to those of magnesium silicate minerals which have been proposed to permanently store CO 2 via mineral carbonation. The study revealed that oxalate even at a low concentration of 0.1 M was effective in enhancing the extent of ACM carbonation and higher reaction temperatures also resulted in increased ACM carbonation. Formation of phases such as dolomite ((Ca, Mg)(CO 3 ) 2 ), whewellite (CaC 2 O 4 ·H 2 O) and glushinskite (MgC 2 O 4 ·2H 2 O) and a reduction in the chrysotile content was noted. Significant changes in the particle size and surface morphologies of ACM and magnesium silicate minerals toward non-fibrous structures were observed after their carbonation

  1. Morphological changes during enhanced carbonation of asbestos containing material and its comparison to magnesium silicate minerals

    Energy Technology Data Exchange (ETDEWEB)

    Gadikota, Greeshma [Department of Chemical Engineering, Columbia University, 500 West 120th Street, New York, NY 10027 (United States); Natali, Claudio; Boschi, Chiara [Institute of Geosciences and Earth Resources – National Research Council, Pisa (Italy); Park, Ah-Hyung Alissa, E-mail: ap2622@columbia.edu [Department of Earth and Environmental Engineering, Columbia University, 500 West 120th Street, New York, NY 10027 (United States); Department of Chemical Engineering, Columbia University, 500 West 120th Street, New York, NY 10027 (United States); Lenfest Center for Sustainable Energy, Columbia University, 500 West 120th Street, New York, NY 10027 (United States)

    2014-01-15

    The disintegration of asbestos containing materials (ACM) over time can result in the mobilization of toxic chrysotile ((Mg, Fe){sub 3}Si{sub 2}O{sub 5}(OH){sub 4})) fibers. Therefore, carbonation of these materials can be used to alter the fibrous morphology of asbestos and help mitigate anthropogenic CO{sub 2} emissions, depending on the amount of available alkaline metal in the materials. A series of high pressure carbonation experiments were performed in a batch reactor at P{sub CO2} of 139 atm using solvents containing different ligands (i.e., oxalate and acetate). The results of ACM carbonation were compared to those of magnesium silicate minerals which have been proposed to permanently store CO{sub 2} via mineral carbonation. The study revealed that oxalate even at a low concentration of 0.1 M was effective in enhancing the extent of ACM carbonation and higher reaction temperatures also resulted in increased ACM carbonation. Formation of phases such as dolomite ((Ca, Mg)(CO{sub 3}){sub 2}), whewellite (CaC{sub 2}O{sub 4}·H{sub 2}O) and glushinskite (MgC{sub 2}O{sub 4}·2H{sub 2}O) and a reduction in the chrysotile content was noted. Significant changes in the particle size and surface morphologies of ACM and magnesium silicate minerals toward non-fibrous structures were observed after their carbonation.

  2. Luminescent hybrid materials functionalized with lanthanide ethylenodiaminotetraacetate complexes containing β-diketonate as antenna ligands

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, Franklin P.; Costa, Israel F.; Espínola, José Geraldo P.; Faustino, Wagner M.; Moura, Jandeilson L. [Departamento de Química-Universidade Federal da Paraíba, 58051-970 João Pessoa, PB (Brazil); Brito, Hermi F.; Paolini, Tiago B. [Departamento de Química Fundamental-Instituto de Química da Universidade de São Paulo, 05508-900 São Paulo, SP (Brazil); Felinto, Maria Cláudia F.C. [Instituto de Pesquisas energéticas e Nucleares-IPEN, 05508-900 São Paulo, SP (Brazil); Teotonio, Ercules E.S., E-mail: teotonioees@quimica.ufpb.br [Departamento de Química-Universidade Federal da Paraíba, 58051-970 João Pessoa, PB (Brazil)

    2016-02-15

    Three organic–inorganic hybrid materials based on silica gel functionalized with (3-aminopropyl)trimethoxysilane (APTS), [3-(2-aminoetilamino)-propil]-trimetoxissilano (DAPTS) and 3-[2-(2-aminoetilamino)etilamino] propiltrimetoxysilane (TAPTS) and subsequently modified with EDTA derivative were prepared by nonhomogeneous route and were then characterized. The resulting materials named SilXN-EDTA (X=1 for APTS, 2 for DAPTS and 3 for TAPTS) were used to obtain new lanthanide Ln{sup 3+}-β-diketonate (Ln{sup 3+}=Eu{sup 3+}, Gd{sup 3+} and Tb{sup 3+}) complexes covalently linked to the functionalized silica gel surfaces (named SilXN-EDTALn-dik, dik=tta, dbm, bzac and acac). The photophysical properties of the new luminescent materials were investigated and compared with those with similar system presenting water molecules coordinated to the lanthanide ions, SilXN-EDTALn-H{sub 2}O. The SilXN-EDTAEu-dik and SilXN-EDTATb-dik systems displayed characteristic red and green luminescence when excited by UV radiation. Furthermore, the quantitative results showed that the emission quantum efficiency (η), experimental intensity parameters Ω{sub 2} and Ω{sub 4}, and Einstein's emission coefficient (A{sub 0J}) of the SilXN-EDTAEu-dik materials were largely dependent on the ligands. Based on the luminescence data, the most efficient intramolecular energy transfer processes were found to the SilXN-EDTAEu-dik (dik: tta and dbm) and SilXN-EDTATb-acac materials, which exhibited more pure emission colors. These materials are promising red and green phosphors, respectively. - Highlights: • New highly luminescent hybrid materials containing lanthanide-EDTA complexes. • The effect of three silylanting agent on the adsorption and luminescent properties has been studied. • The luminescence sensitizing by different β-diketonate ligands have been investigated.

  3. THE NEED FOR A NEW JOINING TECHNOLOGY FOR THE CLOSURE WELDING OF RADIOACTIVE MATERIALS CONTAINERS

    International Nuclear Information System (INIS)

    CANNELL GR; HILL BE; GRANT GJ

    2008-01-01

    One of the activities associated with cleanup throughout the Department of Energy (DOE) complex is packaging radioactive materials into storage containers. Much of this work will be performed in high-radiation environments requiring fully remote operations, for which existing, proven systems do not currently exist. These conditions demand a process that is capable of producing acceptable (defect-free) welds on a consistent basis; the need to perform weld repair, under fully-remote operations, can be extremely costly and time consuming. Current closure welding technology (fusion welding) is not well suited for this application and will present risk to cleanup cost and schedule. To address this risk, Fluor and the Pacific Northwest National Laboratory (PNNL), are proposing that a new and emerging joining technology, Friction Stir Welding (FSW), be considered for this work. FSW technology has been demonstrated in other industries (aerospace and marine) to produce near flaw-free welds on a consistent basis. FSW is judged capable of providing the needed performance for fully-remote closure welding of containers for radioactive materials for the following reasons: FSW is a solid-state process; material is not melted. As such, FSW does not produce the type of defects associated with fusion welding, e.g., solidification-induced porosity, cracking, distortion due to weld shrinkage, and residual stress. In addition, because FSW is a low-heat input process, material properties (mechanical, corrosion and environmental) are preserved and not degraded as can occur with 'high-heat' fusion welding processes. When compared to fusion processes, FSW produces extremely high weld quality. FSW is performed using machine-tool technology. The equipment is simple and robust and well-suited for high radiation, fully-remote operations compared to the relatively complex equipment associated with the fusion-welding processes. Additionally, for standard wall thicknesses of radioactive materials

  4. Carbon black vs. black carbon and other airborne materials containing elemental carbon: Physical and chemical distinctions

    International Nuclear Information System (INIS)

    Long, Christopher M.; Nascarella, Marc A.; Valberg, Peter A.

    2013-01-01

    Airborne particles containing elemental carbon (EC) are currently at the forefront of scientific and regulatory scrutiny, including black carbon, carbon black, and engineered carbon-based nanomaterials, e.g., carbon nanotubes, fullerenes, and graphene. Scientists and regulators sometimes group these EC-containing particles together, for example, interchangeably using the terms carbon black and black carbon despite one being a manufactured product with well-controlled properties and the other being an undesired, incomplete-combustion byproduct with diverse properties. In this critical review, we synthesize information on the contrasting properties of EC-containing particles in order to highlight significant differences that can affect hazard potential. We demonstrate why carbon black should not be considered a model particle representative of either combustion soots or engineered carbon-based nanomaterials. Overall, scientific studies need to distinguish these highly different EC-containing particles with care and precision so as to forestall unwarranted extrapolation of properties, hazard potential, and study conclusions from one material to another. -- Highlights: •Major classes of elemental carbon-containing particles have distinct properties. •Despite similar names, carbon black should not be confused with black carbon. •Carbon black is distinguished by a high EC content and well-controlled properties. •Black carbon particles are characterized by their heterogenous properties. •Carbon black is not a model particle representative of engineered nanomaterials. -- This review demonstrates the significant physical and chemical distinctions between elemental carbon-containing particles e.g., carbon black, black carbon, and engineered nanomaterials

  5. Development of internal dose assessment procedure for workers in industries using raw materials containing naturally occurring radioactive materials

    International Nuclear Information System (INIS)

    Choi, Cheol Kyu; KIm, Yong Geon; Ji, Seung Woo; Kim, Kwang Pyo; Koo, Bon Cheol; Chang, Byung Uck

    2016-01-01

    It is necessary to assess radiation dose to workers due to inhalation of airborne particulates containing naturally occurring radioactive materials (NORM) to ensure radiological safety required by the Natural Radiation Safety Management Act. The objective of this study is to develop an internal dose assessment procedure for workers at industries using raw materials containing natural radionuclides. The dose assessment procedure was developed based on harmonization, accuracy, and proportionality. The procedure includes determination of dose assessment necessity, preliminary dose estimation, airborne particulate sampling and characterization, and detailed assessment of radiation dose. The developed dose assessment procedure is as follows. Radioactivity concentration criteria to determine dose assessment necessity are 10 Bq·g-1 for 40K and 1 Bq·g-1 for the other natural radionuclides. The preliminary dose estimation is performed using annual limit on intake (ALI). The estimated doses are classified into 3 groups (<0.1 mSv, 0.1-0.3 mSv, and >0.3 mSv). Air sampling methods are determined based on the dose estimates. Detailed dose assessment is performed using air sampling and particulate characterization. The final dose results are classified into 4 different levels (<0.1 mSv, 0.1-0.3 mSv, 0.3-1 mSv, and >1 mSv). Proper radiation protection measures are suggested according to the dose level. The developed dose assessment procedure was applied for NORM industries in Korea, including coal combustion, phosphate processing, and monazite handing facilities. The developed procedure provides consistent dose assessment results and contributes to the establishment of optimization of radiological protection in NORM industries

  6. Development of internal dose assessment procedure for workers in industries using raw materials containing naturally occurring radioactive materials

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Cheol Kyu; KIm, Yong Geon; Ji, Seung Woo; Kim, Kwang Pyo [College of Engineering, Kyung Hee University, Yongin (Korea, Republic of); Koo, Bon Cheol; Chang, Byung Uck [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-09-15

    It is necessary to assess radiation dose to workers due to inhalation of airborne particulates containing naturally occurring radioactive materials (NORM) to ensure radiological safety required by the Natural Radiation Safety Management Act. The objective of this study is to develop an internal dose assessment procedure for workers at industries using raw materials containing natural radionuclides. The dose assessment procedure was developed based on harmonization, accuracy, and proportionality. The procedure includes determination of dose assessment necessity, preliminary dose estimation, airborne particulate sampling and characterization, and detailed assessment of radiation dose. The developed dose assessment procedure is as follows. Radioactivity concentration criteria to determine dose assessment necessity are 10 Bq·g-1 for 40K and 1 Bq·g-1 for the other natural radionuclides. The preliminary dose estimation is performed using annual limit on intake (ALI). The estimated doses are classified into 3 groups (<0.1 mSv, 0.1-0.3 mSv, and >0.3 mSv). Air sampling methods are determined based on the dose estimates. Detailed dose assessment is performed using air sampling and particulate characterization. The final dose results are classified into 4 different levels (<0.1 mSv, 0.1-0.3 mSv, 0.3-1 mSv, and >1 mSv). Proper radiation protection measures are suggested according to the dose level. The developed dose assessment procedure was applied for NORM industries in Korea, including coal combustion, phosphate processing, and monazite handing facilities. The developed procedure provides consistent dose assessment results and contributes to the establishment of optimization of radiological protection in NORM industries.

  7. Use of various types of carbon-containing raw materials to produce thermal energy

    Directory of Open Access Journals (Sweden)

    В. Б. Кусков

    2016-08-01

    Full Text Available Many types of carbon-containing organic compounds and all possible carbon-containing products or wastes in low demand can be used to produce thermal energy. A technology has been developed for producing highly flammable briquettes on the basis of bituminous coal. These briquettes have a special incendiary layer. It is easily ignites from low energy heat sources (e.g. matches, and then flame spreads to the rest of briquette. Use of coal slacks and paper wastes as carbon-containing components playing the role of binders provides an opportunity to get a fuel briquette easy in terms of production and plain in composition while at the same time dispose of coal and paper wastes. Such briquettes may also have a special incendiary layer. Technology for fuel briquettes production from wood and slate wastes employed no binding agents, as wood products acted as binders. Thus technologies have been developed to produce fuel briquettes from various carbon-containing materials in low demand. The briquettes are intended for household boilers, fireplaces, different ovens in order to cook food, heat residential and utility premises, cabins, etc.

  8. Optimization of fly ash as sand replacement materials (SRM) in cement composites containing coconut fiber

    Science.gov (United States)

    Nadzri, N. I. M.; Jamaludin, S. B.; Mazlee, M. N.; Jamal, Z. A. Z.

    2016-07-01

    The need of utilizing industrial and agricultural wastes is very important to maintain sustainability. These wastes are often incorporated with cement composites to improve performances in term of physical and mechanical properties. This study presents the results of the investigation of the response of cement composites containing coconut fiber as reinforcement and fly ash use as substitution of sand at different hardening days. Hardening periods of time (7, 14 and 28 days) were selected to study the properties of cement composites. Optimization result showed that 20 wt. % of fly ash (FA) is a suitable material for sand replacement (SRM). Meanwhile 14 days of hardening period gave highest compressive strength (70.12 MPa) from the cement composite containing 9 wt. % of coconut fiber and fly ash. This strength was comparable with the cement without coconut fiber (74.19 MPa) after 28 days of curing.

  9. Numerical Simulation on Dense Packing of Granular Materials by Container Oscillation

    Directory of Open Access Journals (Sweden)

    Jun Liu

    2013-01-01

    Full Text Available The packing of granular materials is a basic and important problem in geomechanics. An approach, which generates dense packing of spheres confined in cylindrical and cuboidal containers in three steps, is introduced in this work. A loose packing structure is first generated by means of a reference lattice method. Then a dense packing structure is obtained in a container by simulating dropping of particles under gravitational forces. Furthermore, a scheme that makes the bottom boundary fluctuate up and down was applied to obtain more denser packing. The discrete element method (DEM was employed to simulate the interactions between particle-particle and particle-boundary during the particles' motions. Finally, two cases were presented to indicate the validity of the method proposed in this work.

  10. Method to prepare essentially organic waste liquids containing radioactive or toxic materials

    International Nuclear Information System (INIS)

    Baehr, W.; Drobnik, S.H.; Hild, W.; Kroebel, R.; Meyer, A.; Naumann, G.

    1976-01-01

    Waste solutions occuring in nuclear technology containing radioactive or toxic materials can be solidified by mixing with a polymerisable mixture with subsequent polymerization. An improvement of this method, especially for liquids in which the radioactive components are present as organic compounds is achieved by adding a mixture of at least one monomeric vinyl compound, at least one polyvinyl compound and appropriate catalysts and by polymerizing at temperatures between 15 and 150 0 C. Should the waste liquid contain mineral acid, this is first neutralized by the addition of CaO or MgO. In processing oils or soaps, the addition of swelling agent for polystyrol resins is advantageous. 16 examples illustrate the invention. (UWI) [de

  11. Contained x-ray diffraction goniometer for examination of radioactive materials

    International Nuclear Information System (INIS)

    Smith, P.K.; Osgood, B.C.; Blaser, D.E.; Howell, R.E.; Stuhler, H.; Stauver, J.

    1987-11-01

    Radioactive materials are being characterized for chemical form and certain physical properties with an x-ray diffraction goniometer customized for containment in a shielded alpha glovebox. A Siemens D500 goniometer was customized by Siemens to locate the associated electronics and x-ray generator outside the glovebox to minimize corrosion and facilitate maintenance. A graphite monochromator is used with a shielded scintillation detector to separate diffracted x-radiation from nuclear radiation. The diffraction system is computer automated for data acquisition and reduction. The facility is designed to handle primarily alpha- and beta-emitting samples with moderate neutron and gamma radiation. Samples containing plutonium, enriched uranium, and other transuranic elements are analyzed in support of site nuclear operations and development programs on nuclear waste, chemical separations, reactor fuels, and product forms

  12. Characteristics of Airborne Particulates Containing Naturally Occurring Radioactive Materials in Monazite Industry

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Geon; Choi, Cheol Kyu; Park, Il; Kim, Min Jun; Go, A Ra; Ji, Seung Woo; Kim, Kwang Pyo [Kyunghee University, Yongin (Korea, Republic of); Koo, Bon Cheol [KINS, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of this study was to characterize physicochemical properties of airborne particulates at a monazite pulverization industry. The properties included particulate size distribution, concentration, shape, density, and radioactivity concentration. Monazite is one of the minerals containing naturally occurring radioactive material (NORM). Therefore, external and internal exposure can be occurred to the workers in monazite industry. The major exposure pathway of the workers is internal exposure due to inhalation of airborne particulates. According to International Commission on Radiological Protection (ICRP), radiation dose due to inhaled particulates containing NORM depends on particulate properties. Therefore, ICRP recommended the internal dose assessment using measured physicochemical properties of the airborne particulates. In the absence of specific information, ICRP provided default reference values. In this study, we characterized physicochemical properties of airborne particulates at a monazite pulverization industry. The databases of particulate information can be used for accurate internal dose assessment of worker.

  13. The ''nuclear car wash'': a scanner to detect illicit special nuclear material in cargo containers

    International Nuclear Information System (INIS)

    Slaughter, D. R.; Accatino, M. R.; Bernstein, A.; Dougan, A. D.; Hall, J. M.; Loshak, A.; Manatt, D. R.; Pohl, B. A.; Prussin, S. G.; Walling, R. S.; Weirup, D. L.

    2004-01-01

    There is an urgent need to improve the reliability of screening cargo containers for illicit nuclear material that may be hidden there for terrorist purposes. A screening system is described for detection of fissionable material hidden in maritime cargo containers. The system makes use of a low intensity neutron beam for producing fission; and the detection of the abundant high-energy γ rays emitted in the β-decay of short-lived fission products and β-delayed neutrons. The abundance of the delayed γ rays is almost an order of magnitude larger than that of the delayed neutrons normally used to detect fission and they are emitted on about the same time scale as the delayed neutrons, i.e., ∼1 min. The energy and temporal distributions of the delayed γ rays provide a unique signature of fission. Because of their high energy, these delayed γ rays penetrate loW--Z cargoes much more readily than the delayed neutrons. Coupled with their higher abundance, the signal from the delayed γ rays escaping from the container is predicted to be as much as six decades more intense than the delayed neutron signal, depending upon the type and thickness of the intervening cargo. The γ rays are detected in a large array of scintillators located along the sides of the container as it is moved through them. Measurements have confirmed the signal strength in somewhat idealized experiments and have also identified one interference when 14.5 MeV neutrons from the D, T reaction are used for the interrogation. The interference can be removed easily by the appropriate choice of the neutron source

  14. Search for hidden high-Z materials inside containers with the Muon Portal Project

    International Nuclear Information System (INIS)

    Rocca, P La; Bandieramonte, M; Blancato, A A; Bonanno, D; Indelicato, V; Presti, D Lo; Petta, C; Antonuccio, V; Becciani, U; Belluso, M; Billotta, S; Bonanno, G; Costa, A; Garozzo, S; Massimino, P; Belluomo, F; Fallica, G; Leonora, E; Longhitano, F; Longo, S

    2014-01-01

    The Muon Portal is a recently born project that plans to build a large area muon detector for a noninvasive inspection of shipping containers in the ports, searching for the presence of potential fissile (U, Pu) threats. The technique employed by the project is the well-known muon tomography, based on cosmic muon scattering from high-Z materials. The design and operational parameters of the muon portal under construction will be described in this paper, together with preliminary simulation and test results

  15. Search for hidden high-Z materials inside containers with the Muon Portal Project

    Science.gov (United States)

    La Rocca, P.; Antonuccio, V.; Bandieramonte, M.; Becciani, U.; Belluomo, F.; Belluso, M.; Billotta, S.; Blancato, A. A.; Bonanno, D.; Bonanno, G.; Costa, A.; Fallica, G.; Garozzo, S.; Indelicato, V.; Leonora, E.; Longhitano, F.; Longo, S.; Lo Presti, D.; Massimino, P.; Petta, C.; Pistagna, C.; Pugliatti, C.; Puglisi, M.; Randazzo, N.; Riggi, F.; Riggi, S.; Romeo, G.; Russo, G. V.; Santagati, G.; Valvo, G.; Vitello, F.; Zaia, A.; Zappalà, G.

    2014-01-01

    The Muon Portal is a recently born project that plans to build a large area muon detector for a noninvasive inspection of shipping containers in the ports, searching for the presence of potential fissile (U, Pu) threats. The technique employed by the project is the well-known muon tomography, based on cosmic muon scattering from high-Z materials. The design and operational parameters of the muon portal under construction will be described in this paper, together with preliminary simulation and test results.

  16. Luminescent materials based on Tb, Eu-containing layered double hydroxides

    International Nuclear Information System (INIS)

    Zhuravleva, N.G.; Eliseev, A.A.; Lukashin, A.V.; Kinast, U.; Tret'yakov, Yu.D.

    2004-01-01

    Luminescent materials on the basis of magnesium-aluminium layered double hydroxides with intercalated anionic complexes of terbium and europium picolinates were synthesized. Relying on data of spectroscopy, elementary and X-ray phase analyses, the change in the rare earth complex structure and metal/ligand ratio, depending on the hydroxide layer charge, determined by Mg/Al ratio in the double hydroxide, were ascertained. The values of quantum yields of luminescence for terbium-containing samples amounted to 30-50% [ru

  17. Adsorption and revaporisation studies on iodine oxide aerosols deposited on containment surface materials in LWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-12-15

    During a hypothetical severe nuclear accident, the radiation field will be very high in the nuclear reactor containment building. As a result gaseous radiolysis products will be formed. Elemental iodine can react in the gaseous phase with ozone to form solid iodine oxide aerosol particles (iodine oxide). Within the AIAS (Adsorption of Iodine oxide Aerosols on Surfaces) project the interactions of iodine oxide (IOx) aerosols with common containment surface materials were investigated. Common surface materials in Swedish and Finnish LWRs are Teknopox Aqua V A paint films and metal surfaces such as Cu, Zn, Al and SS, as well as Pt and Pd surfaces from hydrogen recombiners. Non-radioactive and {sup 131}I labelled iodine oxide aerosols were produced with the EXSI CONT facility from elemental iodine and ozone at VTT Technical Research Centre of Finland. The iodine oxide deposits were analysed with microscopic and spectroscopic measurement techniques to identify the kind of iodine oxide formed and if a chemical conversion on the different surface materials occurs. The revaporisation behaviour of the deposited iodine oxide aerosol particles from the different surface materials was studied under the influence of heat, humidity and gamma irradiation at Chalmers University of Technology, Sweden. Studies on the effects of humidity were performed using the FOMICAG facility, while heat and irradiation experiments were performed in a thermostated heating block and with a gammacell 22 having a dose rate of 14 kGy/h. The revaporisation losses were measured using a HPGe detector. The revaporisated {sup 131}I species from the surfaces were chemically tested for elemental iodine formation. The parameter dominating the degradation of the produced iodine oxide aerosols was humidity. Cu and Zn surfaces were found to react with iodine from the iodine oxide aerosols to form iodides, while no metal iodides were detected for Al and SS samples. Most of the iodine oxide aerosols are assumed to

  18. Adsorption and revaporisation studies on iodine oxide aerosols deposited on containment surface materials in LWR

    International Nuclear Information System (INIS)

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

    2012-12-01

    During a hypothetical severe nuclear accident, the radiation field will be very high in the nuclear reactor containment building. As a result gaseous radiolysis products will be formed. Elemental iodine can react in the gaseous phase with ozone to form solid iodine oxide aerosol particles (iodine oxide). Within the AIAS (Adsorption of Iodine oxide Aerosols on Surfaces) project the interactions of iodine oxide (IOx) aerosols with common containment surface materials were investigated. Common surface materials in Swedish and Finnish LWRs are Teknopox Aqua V A paint films and metal surfaces such as Cu, Zn, Al and SS, as well as Pt and Pd surfaces from hydrogen recombiners. Non-radioactive and 131 I labelled iodine oxide aerosols were produced with the EXSI CONT facility from elemental iodine and ozone at VTT Technical Research Centre of Finland. The iodine oxide deposits were analysed with microscopic and spectroscopic measurement techniques to identify the kind of iodine oxide formed and if a chemical conversion on the different surface materials occurs. The revaporisation behaviour of the deposited iodine oxide aerosol particles from the different surface materials was studied under the influence of heat, humidity and gamma irradiation at Chalmers University of Technology, Sweden. Studies on the effects of humidity were performed using the FOMICAG facility, while heat and irradiation experiments were performed in a thermostated heating block and with a gammacell 22 having a dose rate of 14 kGy/h. The revaporisation losses were measured using a HPGe detector. The revaporisated 131 I species from the surfaces were chemically tested for elemental iodine formation. The parameter dominating the degradation of the produced iodine oxide aerosols was humidity. Cu and Zn surfaces were found to react with iodine from the iodine oxide aerosols to form iodides, while no metal iodides were detected for Al and SS samples. Most of the iodine oxide aerosols are assumed to be

  19. Species identification of processed animal proteins (PAPs) in animal feed containing feed materials from animal origin.

    Science.gov (United States)

    Axmann, Sonja; Adler, Andreas; Brandstettner, Agnes Josephine; Spadinger, Gabriela; Weiss, Roland; Strnad, Irmengard

    2015-01-01

    Since June 2013 the total feed ban of processed animal proteins (PAPs) was partially lifted. Now it is possible to mix fish feed with PAPs from non-ruminants (pig and poultry). To guarantee that fish feed, which contains non-ruminant PAPs, is free of ruminant PAPs, it has to be analysed with a ruminant PCR assay to comply with the total ban of feeding PAPs from ruminants. However, PCR analysis cannot distinguish between ruminant DNA, which originates from proteins such as muscle and bones, and ruminant DNA, which comes from feed materials of animal origin such as milk products or fat. Thus, there is the risk of obtaining positive ruminant PCR signals based on these materials. The paper describes the development of the combination of two analysis methods, micro-dissection and PCR, to eliminate the problem of 'false-positive' PCR signals. With micro-dissection, single particles can be isolated and subsequently analysed with PCR.

  20. Membrane and Films Based on Novel Crown-Containing Dyes as Promising Chemosensoring Materials

    Directory of Open Access Journals (Sweden)

    Sergei Yu. Zaitsev

    2010-12-01

    Full Text Available This paper discusses several works on supramolecular systems such as monolayer and multilayer, polymer films of various crown-containing dyes, surface-active monomers and polymers. Design, production and investigation of the membrane nanostructures based on crown ethers is a rapidly developing field at the “junction” of materials sciences and nanotechnology. These nanostructures can serve as convenient models for studying the self-organization and molecular recognition processes at interfaces that are typical for biomembranes. Based on the results obtained for such structures by absorption and fluorescence spectroscopy, atomic force and Brewster-angle microscopy, surface pressure and surface potential isotherm measurements, the possibility of developing micro- and nanomaterials possessing a set of specified properties (including chemosensor, photochromic and photorefractive materials is demonstrated.

  1. Nupack, the new Asme code for radioactive material transportation packaging containments

    International Nuclear Information System (INIS)

    Turula, P.

    1998-01-01

    The American Society of Mechanical Engineers (ASME) has added a new division to the nuclear construction section of its Boiler and Pressure Vessel Code (B and PVC). This Division, commonly referred to as 'Nupack', has been written to provide a consistent set of technical requirements for containment vessels of transportation packagings for high-level radioactive materials. This paper provides an introduction to Nupack, discusses some of its technical provisions, and describes how it can be used the design and construction of packaging components. Nupack's general provisions and design requirements are emphasized, while treatment of materials, fabrication and inspection is left for another paper. Participation in the Nupack development work described in this paper was supported by the U.S. Department of Energy. (authors)

  2. Full scale impact testing for environmental and safety control of energy material shipping container systems

    International Nuclear Information System (INIS)

    Seagren, R.D.

    1978-01-01

    Heavily-shielded energy material shipping systems, similar in size and weight to those presently employed to transport irradiated reactor fuel elements, are being destructively tested under dynamic conditions. In these tests, the outer and inner steel shells interact in a complex manner with the massive biological shielding in the system. Results obtained from these tests provide needed information for new design concepts. Containment failure (and the resulting release of radioactive material to the environment which might occur in an extremely severe accident) is most likely through the seals and other ancillary features of the shipping systems. Analyses and experiments provide engineering data on the behavior of these shipping systems under severe accident conditions and information for predicting potential survivability and environmental control with a rational margin of safety

  3. 49 CFR 837.3 - Published reports, material contained in the public accident investigation dockets, and accident...

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 7 2010-10-01 2010-10-01 false Published reports, material contained in the... OF RECORDS IN LEGAL PROCEEDINGS § 837.3 Published reports, material contained in the public accident... submitted, in writing, to the Public Inquiries Branch. Demands for specific published reports and studies...

  4. 10 CFR 32.74 - Manufacture and distribution of sources or devices containing byproduct material for medical use.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Manufacture and distribution of sources or devices... SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Generally Licensed Items § 32.74 Manufacture and distribution of sources or devices containing byproduct material for...

  5. Vaccination of dogs with six different candidate leishmaniasis vaccines composed of a chimerical recombinant protein containing ribosomal and histone protein epitopes in combination with different adjuvants.

    Science.gov (United States)

    Poot, J; Janssen, L H M; van Kasteren-Westerneng, T J; van der Heijden-Liefkens, K H A; Schijns, V E J C; Heckeroth, A

    2009-07-16

    Chimerical protein "Q", composed of antigenic ribosomal and histone sequences, in combination with live BCG is a promising canine leishmaniasis vaccine candidate; one of the few vaccine candidates that have been tested successfully in dogs. Unfortunately, live BCG is not an appropriate adjuvant for commercial application due to safety problems in dogs. In order to find a safe adjuvant with similar efficacy to live BCG, muramyl dipeptide, aluminium hydroxide, Matrix C and killed Propionibacterium acnes in combination with either E. coli- or baculovirus-produced recombinant JPCM5_Q protein were tested. Groups of five or seven dogs were vaccinated with six different adjuvant-antigen combinations and challenged with a high dose intravenous injection of Leishmania infantum JPC strain promastigotes. All candidate vaccines proved to be safe, and both humoral and cellular responses to the recombinant proteins were detected at the end of the prime-boost vaccination scheme. However, clinical and parasitological data obtained during the 10 month follow-up period indicated that protection was not induced by either of the six candidate vaccines. Although no direct evidence was obtained, our data suggest that live BCG may have a significant protective effect against challenge with L. infantum in dogs.

  6. Pyrolysis behavior of different type of materials contained in the rejects of packaging waste sorting plants

    Energy Technology Data Exchange (ETDEWEB)

    Adrados, A., E-mail: aitziber.adrados@ehu.es [Chemical and Environmental Engineering Department, School of Engineering of Bilbao, Alameda. Urquijo s/n, 48013 Bilbao (Spain); De Marco, I.; Lopez-Urionabarrenechea, A.; Caballero, B.M.; Laresgoiti, M.F. [Chemical and Environmental Engineering Department, School of Engineering of Bilbao, Alameda. Urquijo s/n, 48013 Bilbao (Spain)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Study of the influence of materials in the pyrolysis of real plastic waste samples. Black-Right-Pointing-Pointer Inorganic compounds remain unaltered. Black-Right-Pointing-Pointer Cellulosic components give rise to an increase in char formation. Black-Right-Pointing-Pointer Cellulosic components promote the production of aqueous phase. Black-Right-Pointing-Pointer Cellulosic components increase CO and CO{sub 2} contents in the gases. - Abstract: In this paper rejected streams coming from a waste packaging material recovery facility have been characterized and separated into families of products of similar nature in order to determine the influence of different types of ingredients in the products obtained in the pyrolysis process. The pyrolysis experiments have been carried out in a non-stirred batch 3.5 dm{sup 3} reactor, swept with 1 L min{sup -1} N{sub 2}, at 500 Degree-Sign C for 30 min. Pyrolysis liquids are composed of an organic phase and an aqueous phase. The aqueous phase is greater as higher is the cellulosic material content in the sample. The organic phase contains valuable chemicals as styrene, ethylbenzene and toluene, and has high heating value (HHV) (33-40 MJ kg{sup -1}). Therefore they could be used as alternative fuels for heat and power generation and as a source of valuable chemicals. Pyrolysis gases are mainly composed of hydrocarbons but contain high amounts of CO and CO{sub 2}; their HHV is in the range of 18-46 MJ kg{sup -1}. The amount of CO-CO{sub 2} increases, and consequently HHV decreases as higher is the cellulosic content of the waste. Pyrolysis solids are mainly composed of inorganics and char formed in the process. The cellulosic materials lower the quality of the pyrolysis liquids and gases, and increase the production of char.

  7. Cesium release from ceramic waste form materials in simulated canister corrosion product containing solutions

    Energy Technology Data Exchange (ETDEWEB)

    Vittorio, Luca; Drabarek, Elizabeth; Chronis, Harriet; Griffith, Christopher S

    2004-07-01

    It has previously been demonstrated that immobilization of Cs{sup +} and/or Sr{sup 2+} sorbed on hexagonal tungsten oxide bronze (HTB) adsorbent materials can be achieved by heating the materials in air at temperatures in the range 500 - 1300 deg C. Highly crystalline powdered HTB materials formed by heating at 800 deg C show leach characteristics comparable to Cs-containing hot-pressed hollandites in the pH range from 0 to 12. As a very harsh leaching test, and also to model in a basic manner, leaching in the presence of canister corrosion products in oxidising environments, leaching of the bronzoid phases has been undertaken in Fe(NO{sub 3}){sub 3} solutions of increasing concentration. This is done in comparison with Cs -hollandite materials in order to compare the leaching characteristics of these two materials under such conditions. Both the Cs-loaded bronze and hollandite materials leach severely in Fe(NO{sub 3}){sub 3} losing virtually all of the immobilized Cs in a period of four days at 150 deg C. Total release of Cs and conversion of hollandite to titanium and iron titanium oxides begins to be observed at relatively low concentrations and is virtually complete after four days reaction in 0.5 mol/L Fe(NO{sub 3}){sub 3}. In the case of the bronze, all of the Cs is also extracted but the HTB structure is preserved. The reaction presumably involves an ion-exchange mechanism and iron oxide with a spinel structure is also observed at high Fe concentrations. (authors)

  8. Cesium release from ceramic waste form materials in simulated canister corrosion product containing solutions

    International Nuclear Information System (INIS)

    Vittorio, Luca; Drabarek, Elizabeth; Chronis, Harriet; Griffith, Christopher S.

    2004-01-01

    It has previously been demonstrated that immobilization of Cs + and/or Sr 2+ sorbed on hexagonal tungsten oxide bronze (HTB) adsorbent materials can be achieved by heating the materials in air at temperatures in the range 500 - 1300 deg C. Highly crystalline powdered HTB materials formed by heating at 800 deg C show leach characteristics comparable to Cs-containing hot-pressed hollandites in the pH range from 0 to 12. As a very harsh leaching test, and also to model in a basic manner, leaching in the presence of canister corrosion products in oxidising environments, leaching of the bronzoid phases has been undertaken in Fe(NO 3 ) 3 solutions of increasing concentration. This is done in comparison with Cs -hollandite materials in order to compare the leaching characteristics of these two materials under such conditions. Both the Cs-loaded bronze and hollandite materials leach severely in Fe(NO 3 ) 3 losing virtually all of the immobilized Cs in a period of four days at 150 deg C. Total release of Cs and conversion of hollandite to titanium and iron titanium oxides begins to be observed at relatively low concentrations and is virtually complete after four days reaction in 0.5 mol/L Fe(NO 3 ) 3 . In the case of the bronze, all of the Cs is also extracted but the HTB structure is preserved. The reaction presumably involves an ion-exchange mechanism and iron oxide with a spinel structure is also observed at high Fe concentrations. (authors)

  9. Direct conversion of plutonium-containing materials to borosilicate glass for storage or disposal

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Beahm, E.C.

    1995-01-01

    A new process, the Glass Material Oxidation and Dissolution System (GMODS), has been invented for the direct conversion of plutonium metal, scrap, and residue into borosilicate glass. The glass should be acceptable for either the long-term storage or disposition of plutonium. Conversion of plutonium from complex chemical mixtures and variable geometries into homogeneous glass (1) simplifies safeguards and security; (2) creates a stable chemical form that meets health, safety, and environmental concerns; (3) provides an easy storage form; (4) may lower storage costs; and (5) allows for future disposition options. In the GMODS process, mixtures of metals, ceramics, organics, and amorphous solids containing plutonium are fed directly into a glass melter where they are directly converted to glass. Conventional glass melters can accept materials only in oxide form; thus, it is its ability to accept materials in multiple chemical forms that makes GMODS a unique glass making process. Initial proof-of-principle experiments have converted cerium (plutonium surrogate), uranium, stainless steel, aluminum, and other materials to glass. Significant technical uncertainties remain because of the early nature of process development

  10. YIELD FORMING EFFECT OF APPLICATION OF COMPOSTS CONTAINING POLYMER MATERIALS ENRICHED IN BIOCOMPONENTS

    Directory of Open Access Journals (Sweden)

    Florian Gambuś

    2014-01-01

    Full Text Available In a pot experiment the impact of composts containing polymeric materials modified with biocomponents on the diversity of crops of oats and mustard was examined. The composts used in the study were produced in the laboratory from wheat and rape straw, and pea seed cleaning waste with 8-percent addition of chopped biopolymer materials (films which were prepared in the Central Mining Institute (GIG in Katowice. Three polymers differing in content of starch and density were selected for the composting. The pot experiment was conducted on three substrates: light and medium soil and on the sediment obtained after flotation of zinc and lead ores, coming from the landfill ZGH “Boleslaw” S.A. in Bukowno. The need for using such materials and substrates results from the conditions of processing some morphological fractions of municipal waste and from improving methods of reclamation. Yield enhancing effect of composts depends on the substrate on which the compost was used, cultivated plants and crop succession. Application of composts prepared with 8% of polymeric materials based on polyethylene, modified with starch as biocomponent, resulted in significantly lower yields in sandy (light soil in case of oats and, in some cases, in medium soil. Subsequent plant yield did not differ significantly between the objects fertilized with compost.

  11. Electro-active polymers containing pendent 2,7-diarylfluorene fragments as materials for OLEDs

    Science.gov (United States)

    Krucaite, G.; Tavgeniene, D.; Peciulyte, L.; Buika, G.; Liu, L.; Zhang, B.; Xie, Z.; Grigalevicius, S.

    2016-05-01

    Poly[2-phenyl-7-(4-vinylphenyl)-9,9-diethylfluorene)], poly[2-(1-naphtyl)-7-(4-vinylphenyl)-9,9-diethylfluorene)] and poly[2-(4-biphenyl)-7-(4-vinylphenyl)-9,9-diethylfluorene)] were synthesized and characterized by NMR spectroscopy, elemental analysis and gel permeation chromatography. The derivatives represent materials of high thermal stability with initial thermal destruction temperatures from 390°C to 400 °C. The glass transition temperatures of the amorphous materials were 182 °C, 151 °C and 159 °C respectively. Hole-transporting properties of the polymeric materials were tested in the structures of organic light emitting diodes with Alq3 as the green emitter and electron transporting material. The device containing hole-transporting layers of polymer with 2-(4-biphenyl)-7-(4-vinylphenyl)-9,9-diethylfluorene moieties exhibited the best overall performance with turn on voltage of 3.6 V, a maximum photometric efficiency of 3.1 cd/A and maximum brightness of about 5300 cd/m2.

  12. Arsenic removal using steel manufacturing byproducts as permeable reactive materials in mine tailing containment systems.

    Science.gov (United States)

    Ahn, Joo Sung; Chon, Chul-Min; Moon, Hi-Soo; Kim, Kyoung-Woong

    2003-05-01

    Steel manufacturing byproducts were tested as a means of treating mine tailing leachate with a high As concentration. Byproduct materials can be placed in situ as permeable reactive barriers to control the subsurface release of leachate from tailing containment systems. The tested materials had various compositions of elemental Fe, Fe oxides, Ca-Fe oxides and Ca hydroxides typical of different steel manufacturing processes. Among these materials, evaporation cooler dust (ECD), oxygen gas sludge (OGS), basic oxygen furnace slag (BOFS) and to a lesser degree, electrostatic precipitator dust (EPD) effectively removed both As(V) and As(III) during batch experiments. ECD, OGS and BOFS reduced As concentrations to <0.5mg/l from 25mg/l As(V) or As(III) solution in 72 h, exhibiting higher removal capacities than zero-valent iron. High Ca concentrations and alkaline conditions (pH ca. 12) provided by the dissolution of Ca hydroxides may promote the formation of stable, sparingly soluble Ca-As compounds. When initial pH conditions were adjusted to 4, As reduction was enhanced, probably by adsorption onto iron oxides. The elution rate of retained As from OGS and ECD decreased with treatment time, and increasing the residence time in a permeable barrier strategy would be beneficial for the immobilization of As. When applied to real tailing leachate, ECD was found to be the most efficient barrier material to increase pH and to remove As and dissolved metals.

  13. Ceramics for Molten Materials Containment, Transfer and Handling on the Lunar Surface

    Science.gov (United States)

    Standish, Evan; Stefanescu, Doru M.; Curreri, Peter A.

    2009-01-01

    As part of a project on Molten Materials Transfer and Handling on the Lunar Surface, molten materials containment samples of various ceramics were tested to determine their performance in contact with a melt of lunar regolith simulant. The test temperature was 1600 C with contact times ranging from 0 to 12 hours. Regolith simulant was pressed into cylinders with the approximate dimensions of 1.25 dia x 1.25cm height and then melted on ceramic substrates. The regolith-ceramic interface was examined after processing to determine the melt/ceramic interaction. It was found that the molten regolith wetted all oxide ceramics tested extremely well which resulted in chemical reaction between the materials in each case. Alumina substrates were identified which withstood contact at the operating temperature of a molten regolith electrolysis cell (1600 C) for eight hours with little interaction or deformation. This represents an improvement over alumina grades currently in use and will provide a lifetime adequate for electrolysis experiments lasting 24 hours or more. Two types of non-oxide ceramics were also tested. It was found that they interacted to a limited degree with the melt resulting in little corrosion. These ceramics, Sic and BN, were not wetted as well as the oxides by the melt, and so remain possible materials for molten regolith handling. Tests wing longer holding periods and larger volumes of regolith are necessary to determine the ultimate performance of the tested ceramics.

  14. Preparation and use of polymeric materials containing hydrophobic anions and plasticizers for separation of cesium and strontium

    International Nuclear Information System (INIS)

    Abney, K.D.; Kinkead, S.A.; Mason, C.F.V.; Rais, J.

    1997-01-01

    Preparation and use is described for polymeric materials containing hydrophobic anions and plasticizers for extraction of cesium and strontium. The use of polymeric materials containing plasticizers which are solvents for hydrophobic anions such as derivatives of cobalt dicarbollide or tetraphenylborate which are capable of extracting cesium and strontium ions from aqueous solutions in contact with the polymeric materials, is described. The polymeric material may also include a synergistic agent for a given ion like polyethylene glycol or a crown ether, for removal of radioactive isotopes of cesium and strontium from solutions of diverse composition and, in particular, for solutions containing large excess of sodium nitrate

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Science.gov (United States)

    Wingard, C. D.

    2018-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-01

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

  18. Mechanical and Electrical Properties of Sulfur-Containing Polymeric Materials Prepared via Inverse Vulcanization

    Directory of Open Access Journals (Sweden)

    Sergej Diez

    2017-02-01

    Full Text Available Recently, new methods have been developed for the utilization of elemental sulfur as a feedstock for novel polymeric materials. One promising method is the inverse vulcanization, which is used to prepare polymeric structures derived from sulfur and divinyl comonomers. However, the mechanical and electrical properties of the products are virtually unexplored. Hence, in the present study, we synthesized a 200 g scale of amorphous, hydrophobic as well as translucent, hyperbranched polymeric sulfur networks that provide a high thermal resistance (>220 °C. The polymeric material properties of these sulfur copolymers can be controlled significantly by varying the monomers as well as the feed content. The investigated comonomers are divinylbenzene (DVB and 1,3-diisopropenylbenzene (DIB. Plastomers with low elastic content and high shape retention containing 12.5%–30% DVB as well as low viscose waxy plastomers with a high flow behavior containing a high DVB content of 30%–35% were obtained. Copolymers with 15%–30% DIB act, on the one hand, as thermoplastics and, on the other hand, as vitreous thermosets with a DIB of 30%–35%. Results of the thermogravimetric analysis (TGA, the dynamic scanning calorimetry (DSC and mechanical characterization, such as stress–strain experiments and dynamic mechanical thermal analysis, are discussed with the outcome that they support the assumption of a polymeric cross-linked network structure in the form of hyper-branched polymers.

  19. Polymers Containing Diphenylvinyl-Substituted Indole Rings as Charge-Transporting Materials for OLEDs

    Science.gov (United States)

    Grigalevicius, S.; Zostautiene, R.; Sipaviciute, D.; Stulpinaite, B.; Volyniuk, D.; Grazulevicius, J. V.; Liu, L.; Xie, Z.; Zhang, B.

    2016-02-01

    Monomers and polymers containing electronically isolated diphenylvinyl-substituted indole rings were synthesized and characterized by nuclear magnetic resonance (NMR) and mass spectroscopies as well as by gel permeation chromatography. The polymers represent amorphous materials with glass transition temperatures of 91-109°C and thermal decomposition starting above 307°C. Electron photoemission spectra of thin films of the synthesized polymers revealed ionization potentials of 5.54-5.58 eV. The synthesized polymers were tested as hole-transporting materials in simple electroluminescent organic light-emitting diode (OLED) devices with tris(quinolin-8-olato)aluminium (Alq3) as an emitter as well as an electron-transporting layer. A green OLED device containing a hole-transporting layer of poly[1-(2,3-epithiopropyl)-2-methyl-3-(2,2-diphenylvinyl)índole] exhibited the best overall performance with a driving voltage of 4.0 V, maximum photometric efficiency of 2.8 cd/A and maximum brightness of about 4200 cd/m2.

  20. Modelling effective dielectric properties of materials containing diverse types of biological cells

    International Nuclear Information System (INIS)

    Huclova, Sonja; Froehlich, Juerg; Erni, Daniel

    2010-01-01

    An efficient and versatile numerical method for the generation of different realistically shaped biological cells is developed. This framework is used to calculate the dielectric spectra of materials containing specific types of biological cells. For the generation of the numerical models of the cells a flexible parametrization method based on the so-called superformula is applied including the option of obtaining non-axisymmetric shapes such as box-shaped cells and even shapes corresponding to echinocytes. The dielectric spectra of effective media containing various cell morphologies are calculated focusing on the dependence of the spectral features on the cell shape. The numerical method is validated by comparing a model of spherical inclusions at a low volume fraction with the analytical solution obtained by the Maxwell-Garnett mixing formula, resulting in good agreement. Our simulation data for different cell shapes suggest that around 1MHz the effective dielectric properties of different cell shapes at different volume fractions significantly deviate from the spherical case. The most pronounced change exhibits ε eff between 0.1 and 1 MHz with a deviation of up to 35% for a box-shaped cell and 15% for an echinocyte compared with the sphere at a volume fraction of 0.4. This hampers the unique interpretation of changes in cellular features measured by dielectric spectroscopy when simplified material models are used.

  1. The carbonaceous sorbent based on the secondary silica-containing material from oil extraction industry

    Science.gov (United States)

    Starostina, I. V.; Stolyarov, D. V.; Anichina, Ya N.; Porozhnyuk, E. V.

    2018-01-01

    The object of research in this work is the silica-containing waste of oil extraction industry - the waste kieselghur (diatomite) sludge from precoat filtering units, used for the purification of vegetable oils from organic impurities. As a result of the thermal modification of the sludge, which contains up to 70% of organic impurities, a finely-dispersed low-porous carbonaceous mineral sorption material is formed. The modification of the sludge particles surface causes the substantial alteration of its physical, chemical, adsorption and structural properties - the organic matter is charred, the particle size is reduced, and on the surface of diatomite particles a carbon layer is formed, which deposits in macropores and partially occludes them. The amount of mesopores is increased, along with the specific surface of the obtained product. The optimal temperature of sludge modification is 500°C. The synthesized carbonaceous material can be used as an adsorbing agent for the purification of wastewater from heavy metal ions. The sorption capacity of Cu2+ ions amounted to 14.2 mg·g-1 and for Ni2+ ions - 17.0 mg·g-1. The obtained values exceed the sorption capacity values of the initial kieselghur, used as a filtering charge, for the researched metal ions.

  2. Annual energy analysis of concrete containing phase change materials for building envelopes

    International Nuclear Information System (INIS)

    Thiele, Alexander M.; Jamet, Astrid; Sant, Gaurav; Pilon, Laurent

    2015-01-01

    Highlights: • Adding PCM to concrete walls can significantly reduce the cooling needs of buildings. • Climate, season, and wall orientation strongly affect energy and cost savings. • The PCM melting temperature should be near the desired indoor temperature. • Benefits are maximum for outdoor temperature oscillating around set indoor temperature. • Adding PCM had little effect on heating energy needs and associated cost savings. - Abstract: This paper examines the annual energy and cost savings potential of adding microencapsulated phase change material to the exterior concrete walls of an average-sized single family home in California climate zones 3 (San Francisco, CA) and 9 (Los Angeles, CA). The annual energy and cost savings were larger for South- and West-facing walls than for other walls. They were also the largest when the phase change temperature was near the desired indoor temperature. The addition of microencapsulated phase change material to the building walls reduced the cooling load in summer substantially more than the heating load in winter. This was attributed to the cold winter temperatures resulting in nearly unidirectional heat flux on many days. The annual cooling load reduction in an average-sized single family home in San Francisco and in Los Angeles ranged from 85% to 100% and from 53% to 82%, respectively, for phase change material volume fraction ranging from 0.1 to 0.3. The corresponding annual electricity cost savings ranged from $36 to $42 in San Francisco and from $94 to $143 in Los Angeles. From an energy standpoint, the best climate for using building materials containing uniformly distributed microencapsulated phase change material would have outdoor temperature oscillations centered around the desired indoor temperature for the entire year

  3. Development of a re-brazeable containment system for special nuclear material storage and transport

    International Nuclear Information System (INIS)

    Pierce, J.D.; Stephens, J.J.; Walker, C.A.; Hosking, F.M.; Curlee, R.M.

    1995-01-01

    This report describes a novel means of closing and sealing small type B radioactive material transport packages for surface or air transport as governed by 10CFR71 or NUREG-0360 has been developed at Sandia National Laboratories (SNL). This method is a controlled brazing process that may be used to attach and seal a closure lid to a containment vessel and then remove it at a later time. The process may be performed multiple times without the need for special preparations of the braze joint. A number of advantages for utilization of this technique have been determined. A brazed seal has integrity at high temperatures for better protection in accident or abnormal environments. A properly designed joint has essentially the same strength as the parent metal. A closure that is brazed, therefore, will no longer be the anticipated point of failure for a broad range of accident environments. This technique will allow the containment vessel design to be optimized with a lighter, more uniform wall thickness throughout. Finally, with a well defined process for sealing, mechanical inspection, leak testing, and then reopening at a later time, automation of the process is relatively straightforward and the overall system should be as easy to use as one that utilizes elastomeric seals for containment

  4. Method for contamination control and barrier apparatus with filter for containing waste materials that include dangerous particulate matter

    Science.gov (United States)

    Pinson, Paul A.

    1998-01-01

    A container for hazardous waste materials that includes air or other gas carrying dangerous particulate matter has incorporated in barrier material, preferably in the form of a flexible sheet, one or more filters for the dangerous particulate matter sealably attached to such barrier material. The filter is preferably a HEPA type filter and is preferably chemically bonded to the barrier materials. The filter or filters are preferably flexibly bonded to the barrier material marginally and peripherally of the filter or marginally and peripherally of air or other gas outlet openings in the barrier material, which may be a plastic bag. The filter may be provided with a backing panel of barrier material having an opening or openings for the passage of air or other gas into the filter or filters. Such backing panel is bonded marginally and peripherally thereof to the barrier material or to both it and the filter or filters. A coupling or couplings for deflating and inflating the container may be incorporated. Confining a hazardous waste material in such a container, rapidly deflating the container and disposing of the container, constitutes one aspect of the method of the invention. The chemical bonding procedure for producing the container constitutes another aspect of the method of the invention.

  5. Method for contamination control and barrier apparatus with filter for containing waste materials that include dangerous particulate matter

    International Nuclear Information System (INIS)

    Pinson, P.A.

    1998-01-01

    A container for hazardous waste materials that includes air or other gas carrying dangerous particulate matter has incorporated barrier material, preferably in the form of a flexible sheet, and one or more filters for the dangerous particulate matter sealably attached to such barrier material. The filter is preferably a HEPA type filter and is preferably chemically bonded to the barrier materials. The filter or filters are preferably flexibly bonded to the barrier material marginally and peripherally of the filter or marginally and peripherally of air or other gas outlet openings in the barrier material, which may be a plastic bag. The filter may be provided with a backing panel of barrier material having an opening or openings for the passage of air or other gas into the filter or filters. Such backing panel is bonded marginally and peripherally thereof to the barrier material or to both it and the filter or filters. A coupling or couplings for deflating and inflating the container may be incorporated. Confining a hazardous waste material in such a container, rapidly deflating the container and disposing of the container, constitutes one aspect of the method of the invention. The chemical bonding procedure for producing the container constitutes another aspect of the method of the invention. 3 figs

  6. Low temperature rheological properties of asphalt mixtures containing different recycled asphalt materials

    Directory of Open Access Journals (Sweden)

    Ki Hoon Moon

    2017-01-01

    Full Text Available Reclaimed Asphalt Pavement (RAP and Recycled Asphalt Shingles (RAS are valuable materials commonly reused in asphalt mixtures due to their economic and environmental benefits. However, the aged binder contained in these materials may negatively affect the low temperature performance of asphalt mixtures. In this paper, the effect of RAP and RAS on low temperature properties of asphalt mixtures is investigated through Bending Beam Rheometer (BBR tests and rheological modeling. First, a set of fourteen asphalt mixtures containing RAP and RAS is prepared and creep stiffness and m-value are experimentally measured. Then, thermal stress is calculated and graphically and statistically compared. The Huet model and the Shift-Homothety-Shift in time-Shift (SHStS transformation, developed at the École Nationale des Travaux Publics de l'État (ENTPE, are used to back calculate the asphalt binder creep stiffness from mixture experimental data. Finally, the model predictions are compared to the creep stiffness of the asphalt binders extracted from each mixture, and the results are analyzed and discussed. It is found that an addition of RAP and RAS beyond 15% and 3%, respectively, significantly change the low temperature properties of asphalt mixture. Differences between back-calculated results and experimental data suggest that blending between new and old binder occurs only partially. Based on the recent finding on diffusion studies, this effect may be associated to mixing and blending processes, to the effective contact between virgin and recycled materials and to the variation of the total virgin-recycled thickness of the binder film which may significantly influence the diffusion process. Keywords: Reclaimed Asphalt Pavement (RAP, Recycled Asphalt Shingles (RAS, Thermal stress, Statistical comparison, Back-calculation, Binder blending

  7. Estimates of fire environments in ship holds containing radioactive material packages

    International Nuclear Information System (INIS)

    Koski, J.A.; Cole, J.K.; Hohnstreiter, G.F.; Wix, S.D.

    1995-01-01

    Fire environments that occur on cargo ships differ significantly from the fire environments found in land transport. Cargo ships typically carry a large amount of flammable fuel for propulsion and shipboard power, and may transport large quantities of flammable cargo. As a result, sea mode transport accident records contain instances of long lasting and intense fires. Since Irradiated Nuclear Fuel (INF) casks are not carried on tankers with large flammable cargoes, most of these dramatic, long burning fires are not relevant threats, and transport studies must concentrate on those fires that are most likely to occur. By regulation, INF casks must be separated from flammable cargoes by a fire-resistant, liquid-tight partition. This makes a fire in an adjacent ship hold the most likely fire threat. The large size of a cargo ship relative to any spent nuclear fuel casks on board, however, may permit a severe, long lasting fire to occur with little or no thermal impact on the casks. Although some flammable materials such as shipping boxes or container floors may exist in the same hold with the cask, the amount of fuel available may not provide a significant threat to the massive transport casks used for radioactive materials. This shipboard fire situation differs significantly from the regulatory conditions specified in 10 CFR 71 for a fully engulfing pool fire. To learn more about the differences, a series of simple thermal analyses has been completed to estimate cask behavior in likely marine and land thermal accident situations. While the calculations are based on several conservative assumptions, and are only preliminary, they illustrate that casks are likely to heat much more slowly in shipboard hold fires than in an open pool fire. The calculations also reinforce the basic regulatory concept that for radioactive materials, the shipping cask, not the ship, is the primary protection barrier to consider

  8. Container material and design considerations for storage of low-level radioactive waste

    International Nuclear Information System (INIS)

    Temus, C.J.

    1987-01-01

    With the threat of increased burial site restrictions and increased surcharges; the ease with which waste is sent to the burial site has been reduced. For many generators of waste the only alternative after maximizing volume reduction efforts is to store the waste. Even after working through the difficult decision of deciding what type of storage facility to have, the decision of what type of container to store the waste in has to still be made. This paper explores the many parameters that affect not only the material selection but also the design. The proper selection of materials affect the ability of the container to survive the storage period. The material selection also directly affects the design and utilization of the storage facility. The impacts to the facility include the functional aspects as well as its operational cost and liability as related to such things as fire insurance and active environmental control systems. The advantages and disadvantages of many of the common systems such as carbon steel, various coatings, polyethylene, stainless steel, composites and concrete will be discussed and evaluated. Recognizing that the waste is to be disposed of in the future differentiates it from waste that is shipped directly to the disposal site. The stored waste has to have the capability to be handled not only once like the disposal site waste but potentially several times before ultimate disposal. This handling may be by several different systems both at the storage facility and the burial site. Some of these systems due to ALARA considerations are usually remote requiring various interfaces, while not interfering with handling, transportation or disposal operations

  9. Natural waste materials containing chitin as adsorbents for textile dyestuffs: batch and continuous studies.

    Science.gov (United States)

    Figueiredo, S A; Loureiro, J M; Boaventura, R A

    2005-10-01

    In this work three natural waste materials containing chitin were used as adsorbents for textile dyestuffs, namely the Anodonta (Anodonta cygnea) shell, the Sepia (Sepia officinalis) and the Squid (Loligo vulgaris) pens. The selected dyestuffs were the Cibacron green T3G-E (CI reactive green 12), and the Solophenyl green BLE 155% (CI direct green 26), both from CIBA, commonly used in cellulosic fibres dyeing, the most used fibres in the textile industry. Batch equilibrium studies showed that the materials' adsorption capacities increase after a simple and inexpensive chemical treatment, which increases their porosity and chitin relative content. Kinetic studies suggested the existence of a high internal resistance in both systems. Fixed bed column experiments performed showed an improvement in adsorbents' behaviour after chemical treatment. However, in the column experiments, the biodegradation was the main mechanism of dyestuff removal, allowing the materials' bioregeneration. The adsorption was strongly reduced by the pore clogging effect of the biomass. The deproteinised Squid pen (grain size 0.500-1.41 mm) is the adsorbent with highest adsorption capacity (0.27 and 0.037 g/g, respectively, for the reactive and direct dyestuffs, at 20 degrees C), followed by the demineralised Sepia pen and Anodonta shell, behaving like pure chitin in all experiments, but showing inferior performances than the granular activated carbon tested in the column experiments.

  10. Compatibility of Space Nuclear Power Plant Materials in an Inert He/Xe Working Gas Containing Reactive Impurities

    International Nuclear Information System (INIS)

    MM Hall

    2006-01-01

    A major materials selection and qualification issue identified in the Space Materials Plan is the potential for creating materials compatibility problems by combining dissimilar reactor core, Brayton Unit and other power conversion plant materials in a recirculating, inert He/Xe gas loop containing reactive impurity gases. Reported here are results of equilibrium thermochemical analyses that address the compatibility of space nuclear power plant (SNPP) materials in high temperature impure He gas environments. These studies provide early information regarding the constraints that exist for SNPP materials selection and provide guidance for establishing test objectives and environments for SNPP materials qualification testing

  11. The interaction between contacting barrier materials for containment of radioactive wastes

    International Nuclear Information System (INIS)

    Chang, Hao-Chun; Wang, Chun-Yao; Huang, Wei-Hsing

    2012-01-01

    Document available in extended abstract form only. The disposal of low-level radioactive wastes requires multi-barrier facilities to contain the wastes from contamination. Typically, the engineered barrier is composed of a concrete vault backfilled with sand/bentonite mixture. The backfill material is a mixture of bentonite and sand/gravel produced from crushing the rocks excavated at the site. With a great swelling potential, bentonite is expected to serve the sealing function, while the crushed sand/gravel improves the workability of the mixture. Due to the nature of radioactive wastes, the disposal site is designed for a service life of 300 years or more, which is much longer than typical engineering or earth works. With such a long service life, the site is subject to groundwater intrusion and geochemical evolution. The near-field environment evolution can be a complex problem in a disposal site. In the vicinity of the concrete vault in a disposal site, the high-alkali concrete environment can cause changes in the pore solution and alter the nature of backfill materials. Therefore, the interaction between the concrete and the backfill material needs to be assessed, such that the barriers serve the expected functions for a long time. Materials and Methods A locally available Zhishin clay and a bentonite originated from Black Hill, Wyoming, USA were used as raw clay materials in this study. Zhishin clay and Black Hill (BH) bentonite are mixed with Taitung area hard shale to produce the backfill material. An experimental program was conducted analysing the soil properties of these 2 bentonites. And an accelerated migration test was devised to understand the loss of calcium leaching of concrete on characteristics of backfill material. The 2 barrier materials (concrete and backfill) were placed in contact and then an electric gradient applied to accelerate the move of cations between the 2 barriers. Fig. 1 shows a schematic diagram of the accelerated migration test

  12. Investigation on fabrication of SiC/SiC composite as a candidate material for fuel sub-assembly

    International Nuclear Information System (INIS)

    Lee, Jae-Kwang; Naganuma, Masayuki; Park, Joon-Soo; Kohyama, Akira

    2005-01-01

    The possibility of SiC/SiC (Silicon carbide fiber reinforced Silicon carbide) composites application for fuel sub-assembly of Fast Breeder Reactor was investigated. To select a raw material of SiC/SiC composites, a few kinds of SiC nano powder was estimated by SEM observation and XRD analysis. Furthermore, SiC monolithic was sintered from them and estimated by flexural test. SiC nano-powder which showed good sinterability, it was used for fabrication of SiC/SiC composites by Hot Pressing method. From the sintering condition of 1800, 1820degC temperature and 15, 20 MPa pressure, SiC/SiC composite was fabricated and then estimated by tensile test. SiC/SiC composite, which made by 1820degC and 20 MPa condition, showed the highest mechanical strength by the monotonic tensile test. SiC/SiC composite, which made by 1800degC and 15 MPa condition, showed a stable fracture behavior at the monotonic and cyclic tensile test. And then, the hoop stress of ideal model of SiC/SiC composites was discussed. It was confirmed that applicability of SiC/SiC composites by Hot Pressing method for fuel sub-assembly structural material. To make it real attractive one, to maintain the reliability and safety as a high temperature structural material, the design and process study on SiC/Sic composites material will be continued. (author)

  13. Experimental demonstration of radiation effects on the performance of a stirling-alternator convertor and candidate materials evaluation

    Science.gov (United States)

    Mireles, Omar R.

    Free-piston Stirling power convertors are under consideration by NASA for service in the Advanced Stirling Radioisotope Generator (ASRG) and Fission Surface Power (FSP) systems to enable aggressive exploration missions by providing a reliable and constant power supply. The ASRG must withstand environmental radiation conditions, while the FSP system must tolerate a mixed neutron and gamma-ray environment resulting from self-irradiation. Stirling-alternators utilize rare earth magnets and a variety of organic materials whose radiation limits dominate service life estimates and shielding requirements. The project objective was to demonstrate the performance of the alternator, identify materials that exhibit excessive radiation sensitivity, identify radiation tolerant substitutes, establish empirical dose limits, and demonstrate the feasibility of cost effective nuclear and radiation tests by selection of the appropriate personnel and test facilities as a function of hardware maturity. The Stirling Alternator Radiation Test Article (SARTA) was constructed from linear alternator components of a Stirling convertor and underwent significant pre-exposure characterization. The SARTA was operated at the Sandia National Laboratories Gamma Irradiation Facility to a dose of over 40 Mrad. Operating performance was within nominal variation, although modestly decreasing trends occurred in later runs as well as the detection of an electrical fault after the final exposure. Post-irradiation disassembly and internal inspection revealed minimal degradation of the majority of the organic components. Radiation testing of organic material coupons was conducted since the majority of the literature was inconsistent. These inconsistencies can be attributed to testing at environmental conditions vastly different than those Stirling-alternator organics will experience during operation. Samples were irradiated at the Texas A&M TRIGA reactor to above expected FSP neutron fluence. A thorough

  14. Possibilities of using new technology materials in constructing the radioactive waste containers The paper will consider using the latest technologies in material science for building

    International Nuclear Information System (INIS)

    Itu, Razvan Bogdan

    2008-01-01

    The paper will consider using the latest technologies in materials science for building the radioactive waste containers. A new amorphous steel has been discovered by the scientists from the University of Virginia, a material three times stronger then conventional steel and non-magnetic. Scientists shown that this steel, DARVA - Glass 101, has superior anticorrosive proprieties. The paper will also consider using Para-Aramides in protecting the radioactive waste containers. Chemical and physical properties of these materials shown a great tensile strength and the inter-chain bonds make these materials extremely strong. (author)

  15. High Efficient Enrichment and Activated Dissolution of Refractory Low Grade Rh-containing Material

    Institute of Scientific and Technical Information of China (English)

    WU Xiaofeng; DONG Haigang; TONG Weifeng; ZHAO Jiachun; ZENG Rui

    2012-01-01

    Aiming to the low-grade rhodium-containing waste materials,a new process was proposed to enrich and activate rhodium by smelting using iron oxide as a trapping agent and activator.A rhodium concentrate was obtained by the separation of base metals and precious metals.The concentrate was reacted with dilute aqua regia to obtain rhodium solution.The factors influencing the enrichment and activation effects were discussed in this paper.The results showed that the dissolution rate is greater than 99% under the optimum conditions.In this process,the activation of rhodium was finished in the enrichment process.The iron oxide is both a trapping agent and activator,which simplifies the process and reduce the cost.

  16. Lignases and aldo-keto reductases for conversion of lignin-containing materials to fermentable products

    Science.gov (United States)

    Scharf, Michael; Sethi, Amit

    2016-09-13

    Termites have specialized digestive systems that overcome the lignin barrier in wood to release fermentable simple sugars. Using the termite Reticulitermes flavipes and its gut symbionts, high-throughput titanium pyrosequencing and proteomics approaches experimentally compared the effects of lignin-containing diets on host-symbiont digestome composition. Proteomic investigations and functional digestive studies with recombinant lignocellulases conducted in parallel provided strong evidence of congruence at the transcription and translational levels and provide enzymatic strategies for overcoming recalcitrant lignin barriers in biofuel feedstocks. Briefly described, therefore, the disclosure provides a system for generating a fermentable product from a lignified plant material, the system comprising a cooperating series of at least two catalytically active polypeptides, where said catalytically active polypeptides are selected from the group consisting of: cellulase Cell-1, .beta.-glu cellulase, an aldo-keto-reductase, a catalase, a laccase, and an endo-xylanase.

  17. Experimental study of a laboratory concrete material representative of containment buildings: desorption isotherms and permeability determination

    International Nuclear Information System (INIS)

    Semete, P.; Fevrier, B.; Delorme, J.; Sanahuja, J.; Desgree, P.; Le Pape, Y.

    2015-01-01

    The isotherm sorption curve is a first order parameter for the calculations of concrete drying and/or creep using Finite Element Analysis. An experimental campaign was undertaken by EDF MMC in order to characterize the first desorption isotherm at room temperature of a laboratory material representative of concrete containment buildings. Long term drying tests were carried out on cement paste and on three samples geometries on concrete (with radial and axial one-dimensional drying on thin disks and multi-dimensional drying on Representative Elementary Volumes). The measurements results (porosity, densities and mass loss curves) are provided and the isotherms obtained for the four different configurations are compared. Several analyses of the results are proposed including the assessment of a criterion for the determination of the moisture content final balance (estimation of the asymptotic mass loss) and the back-analysis of equivalent permeability. (authors)

  18. Improvements in or relating to the production of metal-containing material in particulate form

    International Nuclear Information System (INIS)

    Woodhead, J.L.; Scott, K.T.B.; Ball, P.W.

    1977-01-01

    The process described refers mainly to production of the material in the form of very small spheres. It comprises forming a metal compound-containing gel precipitate by mixing a solution or sol of the metal compound with a soluble organic polymer and contacting the mixture with a precipitating reagent to precipitate the metal as an insoluble compound bound with the polymer. The precipitate is then subjected in the liquid phase to a breaking down and dispersing process to produce an intermediate product suitable for spray drying, and the intermediate product is spray dried to form the particulate product. The breaking down and dispersing process may be performed by means of a colloid mill or vibratory stirrer. Examples of application of the process are described. (U.K.)

  19. Fiber reinforced concrete as a material for nuclear reactor containment buildings

    International Nuclear Information System (INIS)

    Mallikarjuna; Banthia, N.; Mindess, S.

    1991-01-01

    The fiber reinforced concrete as a constructional material for nuclear reactor containment buildings calls for an examination of its individual characteristics and potentialities due to its inherent superiority over normal plain and reinforced concrete. In the present investigation, first, to study the static behavior of straight, hooked-end and crimped fibers, recently developed nonlinear three-dimensional interface (contact) element has been used in conjunction with the eight nodded hexahedron and two nodded bar elements for concrete and steel fiber respectively. Then impact tests were carried out on fiber reinforced concrete beams with an instrumented drop weight impact machine. Two different concrete mixes were tested: normal strength and high strength concrete specimens. Fibers in the concrete mix found to significantly increase the ductility and the impact resistance of the composite. Deformed fibers increase peak pull-out load and pull-out distance, and perform better in the steel fiber reinforced concrete (SFRC) structures. (author)

  20. Attenuation and efficacy of human parainfluenza virus type 1 (HPIV1 vaccine candidates containing stabilized mutations in the P/C and L genes

    Directory of Open Access Journals (Sweden)

    Skiadopoulos Mario H

    2007-07-01

    Full Text Available Abstract Background Two recombinant, live attenuated human parainfluenza virus type 1 (rHPIV1 mutant viruses have been developed, using a reverse genetics system, for evaluation as potential intranasal vaccine candidates. These rHPIV1 vaccine candidates have two non-temperature sensitive (non-ts attenuating (att mutations primarily in the P/C gene, namely CR84GHNT553A (two point mutations used together as a set and CΔ170 (a short deletion mutation, and two ts att mutations in the L gene, namely LY942A (a point mutation, and LΔ1710–11 (a short deletion, the last of which has not been previously described. The latter three mutations were specifically designed for increased genetic and phenotypic stability. These mutations were evaluated on the HPIV1 backbone, both individually and in combination, for attenuation, immunogenicity, and protective efficacy in African green monkeys (AGMs. Results The rHPIV1 mutant bearing the novel LΔ1710–11 mutation was highly ts and attenuated in AGMs and was immunogenic and efficacious against HPIV1 wt challenge. The rHPIV1-CR84G/Δ170HNT553ALY942A and rHPIV1-CR84G/Δ170HNT553ALΔ1710–11 vaccine candidates were highly ts, with shut-off temperatures of 38°C and 35°C, respectively, and were highly attenuated in AGMs. Immunization with rHPIV1-CR84G/Δ170HNT553ALY942A protected against HPIV1 wt challenge in both the upper and lower respiratory tracts. In contrast, rHPIV1-CR84G/Δ170HNT553ALΔ1710–11 was not protective in AGMs due to over-attenuation, but it is expected to replicate more efficiently and be more immunogenic in the natural human host. Conclusion The rHPIV1-CR84G/Δ170HNT553ALY942A and rHPIV1-CR84G/Δ170HNT553ALΔ1710–11 vaccine candidates are clearly highly attenuated in AGMs and clinical trials are planned to address safety and immunogenicity in humans.