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Sample records for vessel material capsule

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

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-12-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  8. Analysis of mechanical property data obtained from nuclear pressure vessel surveillance capsules

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1977-01-01

    A typical pressure vessel surveillance capsule examination program provides mechanical property data from tensile, Charpy V-notch impact, and, in some cases, fracture mechanics specimens. This data must be analyzed in conjunction with the unirradiated baseline mechanical property data to determine the effect of irradiation on the mechanical properties. In the case of Charpy impact specimens, for example, irradiation typically causes an increase in the transition temperature, and a decrease in the upper shelf energy level. The results of the Charpy impact and other mechanical specimen tests must be evaluated to determine if property changes are occurring in the manner expected when the reactor was put into service. The large amount of data obtained from surveillance capsule examinations in recent years enables one to make fairly good predictions. After the changes in the mechanical properties of specimens from a particular surveillance capsule have been experimentally determined and evaluated, they must be related to the reactor pressure vessel. This requires a knowledge of the neutron fluence of the surveillance capsule, and the ratio of the surveillance capsule fluence to the pressure vessel wall fluence. This ratio is frequently specified by the reactor manufacturer, or can be calculated from a knowledge of the geometry and materials of the reactor components inside the pressure vessel. A knowledge of the exact neutron fluence of the capsule specimens and the capsule to vessel wall neutron fluence ratio is of great importance, since inaccuracies in these numbers cause just as serious a problem as inaccuracies in the mechanical property determinations. A further area causing analysis difficulties is problems encountered in recent capsule programs relating to capsule design, construction, operation, and dismantling. (author)

  9. Capsule Development and Utilization for Material Irradiation Tests

    International Nuclear Information System (INIS)

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

    2003-05-01

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

  10. Hot cell examination on the surveillance capsule of SA 533 cl. 1 reactor pressure vessel (1st test report)

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yong Sun; Jung, Y. H.; Yoo, B. O.; Baik, S. J.; Oh, W. H.; Soong, W. S.; Hong, K. P

    2000-08-01

    The post-irradiated examinations such as impact test, tensile test, composition analysis and etc. were conducted to monitor and to evaluate the radiation-induced changes, so called radiation embrittlement, in the mechanical properties of ferritic materials. Those data should be applied to confirm safety as well as reliability of reactor pressure vessel. The scopes and contents of hot cell examination on the surveillance capsule are as follows; - Capsule transportation, cutting, dismantling and classification - Shim block and Dosimeter cutting and dismantling - Impact test - Tensile test - Composition analysis by EPMA - SEM observation on the fractured surface - Hardness test - Radwaste treatment.

  11. Development of glia and blood vessels in the internal capsule of rats.

    Science.gov (United States)

    Earle, K L; Mitrofanis, J

    1998-02-01

    We have explored two aspects of internal capsule development that have not been described previously, namely, the development of glia and of blood vessels. To these ends, we used antibodies to glial fibrillary acidic protein (GFAP) and to vimentin (to identify astrocytes and to radial glia) and Griffonia simplicifolia (lectin; to identify microglia and blood vessels). Further, we made intracardiac injections of Evans Blue to examine the permeability of this dye in the vessels of the internal capsule during neonatal development. Our results show that large numbers of radial glia, astrocytes and microglia are not labelled with these markers in the white matter of the internal capsule until about birth; very few are labelled earlier, during the critical stages of corticofugal and corticopetal axonal ingrowth (E15-E20). The large glial labelling in the internal capsule at birth is accompanied by a dense vascular innervation of the capsule; as with the glia, very few labelled patent vessels are seen earlier. After intracardiac injections of Evans Blue, we find that the blood vessels of the internal capsule are not particularly permeable to Evans Blue. At each age examined (P0, P5, P15), blood vessels are outlined very clearly and there is no diffuse haze of fluorescence within the extracellular space, which is indicative of a leaky vessel. There are three striking differences between the glial environment of the internal capsule and that of the adjacent thalamus. First, the internal capsule is never rich with radial glial fibres (vimentin- and GFAP-immunoreactive) during development (except at P0), whereas the thalamus has many radial fibres from very early development (E15-E17). Second, astrocytes (vimentin- and GFAP-immunoreactive) first become apparent in the internal capsule (E20-P0) well before they do in the thalamus (P15). Third, the internal capsule houses a large transient population of amoeboid microglia (P0-P22), whereas the thalamus does not; only ramified

  12. Capsule development and utilization for material irradiation tests

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  13. Capsule development and utilization for material irradiation tests

    International Nuclear Information System (INIS)

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

    2000-05-01

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

  14. Capsule development and utilization for material irradiation tests

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  15. 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.)

  16. Analysis of the necessity for inserting new surveillance capsule into the Kori Unit 1 RPV to monitor material fracture toughness

    International Nuclear Information System (INIS)

    Song, Taek Ho

    2007-01-01

    In association with monitoring of reactor pressure vessel (RPV) fracture toughness, surveillance capsule test specimens have been used to monitor the material property of nuclear reactor vessel. As far as Kori Unit 1 is concerned, 6 capsules were put into the vessel before commercial operation of the plant. Up to now, all the six capsules have been withdrawn to test and monitor the fracture toughness of RPV material. The last capsule has been withdrawn on June this year, and the Kori unit 1 has been shut downed since July 2007 and will be shut downed until December this year for about 6 months, preparing the life extension of the plant to operate the plant 10 more years. With the situation that all the surveillance capsules have been withdrawn, public ask the following question, 'To extend the life of Kori Unit 1 more than 10 years, is it necessary to insert new surveillance capsules into the Kori Unit 1 to monitor RPV material fracture toughness?' In connection with this issue, planning project have been carried out since spring this year. In this paper, it is described that inserting new surveillance capsule into the Kori Unit 1 RPV has some meaning in some public acceptance point of view and is not necessary in material engineering point of view

  17. Analytical Model for the Probability Characteristics of a Crack Penetrating Capsules in Capsule-Based Self-Healing Cementitious Materials

    Directory of Open Access Journals (Sweden)

    Zhong LV

    2017-08-01

    Full Text Available Autonomous crack healing using pre-embedded capsules containing healing agent is becoming a promising approach to restore the strength of damaged structures. In addition to the material properties, the size and volume fraction of capsules influence crack healing in the matrix. Understanding the crack and capsule interaction is critical in the development and design of structures made of capsule-based self-healing materials. Continuing our previous study, in this contribution a more practical rupturing mode of capsules characterizing the rupturing manner of capsules fractured by cracks in cementitious materials is presented, i.e., penetrating mode. With the underlying assumption that a crack penetrating capsules undoubtedly leads to crack healing, geometrical probability theory is employed to develop the quantitative relationship between crack size and capsule size, capsule concentration in capsule-based self-healing virtual cementitious material. Moreover, an analytical expression of probability of a crack penetrating with randomly dispersed capsules is developed in two-dimensional material matrix setup. The influences of the induced rupturing modes of capsules embedded on the self-healing efficiency are analyzed. Much attention is paid to compare the penetrating probability and the hitting probability, in order to assist the designer to make a choice of the optimal rupturing modes of capsules embedded. The accuracy of results of the theoretical model is also compared with Monte-Carlo numerical analysis of crack interacting with capsules. It shows that the developed probability characteristics of a crack interaction with capsules for different rupturing modes is helpful to provide guidelines for designer working with capsule-based self-healing cementitious materials.DOI: http://dx.doi.org/10.5755/j01.ms.23.3.16888

  18. Radiation research of materials using irradiation capsules

    International Nuclear Information System (INIS)

    Chamrad, B.

    1976-01-01

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

  19. Data package for the Turkey Point material interaction test capsules

    International Nuclear Information System (INIS)

    Krogness, J.C.; Davis, R.B.

    1979-01-01

    Objective of the Materials Interaction Test (MIT) is to obtain interaction information on candidate package storage materials and geologies under prototypic temperatures in gamma and low level neutron fields. Compatibility, structural properties, and chemical transformations will be studied. The multiple test samples are contained within test capsules connected end-to-end to form a test train. Only passive instrumentation has been used to monitor temperatures and record neutron fluence. The test train contains seven capsules: three to test compatibility, two for structural tests, and two for chemical transformation studies. The materials tested are potential candidates for the spent fuel package canister and repository geologies

  20. The microstructure of capsule containing self-healing materials: A micro-computed tomography study

    Energy Technology Data Exchange (ETDEWEB)

    Van Stappen, Jeroen, E-mail: Jeroen.Vanstappen@ugent.be [UGCT/PProGRess, Dept. of Geology, Ghent University, Krijgslaan 281 S8, B-9000 Ghent (Belgium); SIM vzw, Technologiepark 935, B-9052 Zwijnaarde (Belgium); Bultreys, Tom [UGCT/PProGRess, Dept. of Geology, Ghent University, Krijgslaan 281 S8, B-9000 Ghent (Belgium); Gilabert, Francisco A. [Mechanics of Materials and Structures, Dept. of Materials Science and Engineering, Ghent University, Technologiepark Zwijnaarde 903, B-9052 Zwijnaarde (Belgium); SIM vzw, Technologiepark 935, B-9052 Zwijnaarde (Belgium); Hillewaere, Xander K.D. [Polymer Chemistry Research Group, Dept. of Organic and Macromolecular Chemistry, Ghent University, Krijgslaan 281 S4-bis, B-9000 Ghent (Belgium); SIM vzw, Technologiepark 935, B-9052 Zwijnaarde (Belgium); Gómez, David Garoz [Mechanics of Materials and Structures, Dept. of Materials Science and Engineering, Ghent University, Technologiepark Zwijnaarde 903, B-9052 Zwijnaarde (Belgium); SIM vzw, Technologiepark 935, B-9052 Zwijnaarde (Belgium); Van Tittelboom, Kim [Magnel Laboratory for Concrete Research, Dept. of Structural Engineering, Ghent University, Technologiepark Zwijnaarde 904, B-9052 Ghent (Belgium); Dhaene, Jelle [UGCT/Radiation Physics, Dept. of Physics and Astronomy, Ghent University, Proeftuinstraat 86, B-9000 Ghent (Belgium); De Belie, Nele [Magnel Laboratory for Concrete Research, Dept. of Structural Engineering, Ghent University, Technologiepark Zwijnaarde 904, B-9052 Ghent (Belgium); Van Paepegem, Wim [Mechanics of Materials and Structures, Dept. of Materials Science and Engineering, Ghent University, Technologiepark Zwijnaarde 903, B-9052 Zwijnaarde (Belgium); Du Prez, Filip E. [Polymer Chemistry Research Group, Dept. of Organic and Macromolecular Chemistry, Ghent University, Krijgslaan 281 S4-bis, B-9000 Ghent (Belgium); Cnudde, Veerle [UGCT/PProGRess, Dept. of Geology, Ghent University, Krijgslaan 281 S8, B-9000 Ghent (Belgium)

    2016-09-15

    Autonomic self-healing materials are materials with built-in (micro-) capsules or vessels, which upon fracturing release healing agents in order to recover the material's physical and mechanical properties. In order to better understand and engineer these materials, a thorough characterization of the material's microstructural behavior is essential and often overlooked. In this context, micro-computed tomography (μCT) can be used to investigate the three dimensional distribution and (de)bonding of (micro-) capsules in their native state in a polymer system with self-healing properties. Furthermore, in-situ μCT experiments in a self-healing polymer and a self-healing concrete system can elucidate the breakage and leakage behavior of (micro-) capsules at the micrometer scale. While challenges related to image resolution and contrast complicate the characterization in specific cases, non-destructive 3D imaging with μCT is shown to contribute to the understanding of the link between the microstructure and the self-healing behavior of these complex materials. - Highlights: • μCT imaging allows for the analysis of microcapsule distribution patterns in self-healing materials. • μCT allows for qualitative and quantitative measurements of healing agent release from carriers in self-healing materials. • Experimental set-ups can be optimized by changing chemical compounds in the system to ensure maximum quality imaging.

  1. Capsule Development and Utilization for Material Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-15

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

  2. Capsule Development and Utilization for Material Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-15

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

  3. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

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

    1980-10-01

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

  4. Flexible Composite-Material Pressure Vessel

    Science.gov (United States)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  6. Components production and assemble of the irradiation capsule of the Surveillance Program of Materials of the nuclear power plant of Laguna Verde

    International Nuclear Information System (INIS)

    Medrano, A.

    2009-01-01

    To predict the effects of the neutrons radiation and the thermal environment about the mechanical properties of the reactor vessel materials of the nuclear power plant of Laguna Verde, a surveillance program is implemented according to the outlines settled by Astm E185-02 -Standard practice for design of surveillance programs for light-water moderated nuclear power reactor vessels-. This program includes the installation of three irradiation capsules of similar materials to those of the reactor vessels, these samples are test tubes for mechanical practices of impact and tension. In the National Institute of Nuclear Research and due to the infrastructure as well as of the actual human resources of the Pilot Plant of Nuclear Fuel Assembles Production it was possible to realize the materials rebuilding extracted in 2005 of Unit 2 of nuclear power plant of Laguna Verde as well as the production, assemble and reassignment of the irradiation capsule made in 2006. At the present time the surveillance materials extracted in 2008 of Unit 1 of the nuclear power plant of Laguna Verde are reconstituting and the components are manufactured for the assembles of the irradiation capsule that will be reinstalled in the reactor vessel in 2010. The purpose of the present work is to describe the necessary components as well as its disposition during the assembles of the irradiation capsule for the surveillance program of the reactors vessel of the nuclear power plant of Laguna Verde. (Author)

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

    International Nuclear Information System (INIS)

    Conlin, J.A.

    1985-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

  9. Work plan for testing silicone impression material and fixture on pool cell capsule

    International Nuclear Information System (INIS)

    Lundeen, J.E.

    1994-01-01

    The purpose of this work plan is to provide a safe procedure to test a cesium capsule impression fixture at Waste Encapsulation and Storage Facility (WESF). The impression will be taken with silicone dental impression material pressed down upon the capsule using the impression fixture. This test will evaluate the performance of the fixture and impression material under high radiation and temperature conditions on a capsule in a WESF pool cell

  10. Capsule shell material impacts the in vitro disintegration and dissolution behaviour of a green tea extract

    OpenAIRE

    Glube, Natalie; Moos, Lea von; Duchateau, Guus

    2013-01-01

    Purpose In vitro disintegration and dissolution are routine methods used to assess the performance and quality of oral dosage forms. The purpose of the current work was to determine the potential for interaction between capsule shell material and a green tea extract and the impact it can have on the release. Methods A green tea extract was formulated into simple powder-in-capsule formulations of which the capsule shell material was either of gelatin or HPMC origin. The disintegration times we...

  11. 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.

  12. Equation of state study of Laser Megajoule capsules ablator materials

    International Nuclear Information System (INIS)

    Colin-Lalu, Pierre

    2016-01-01

    This PhD thesis enters the field of inertial confinement fusion studies. In particular, it focuses on the equation of state tables of ablator materials synthesized on LMJ capsules. This work is indeed aims at improving the theoretical models introduced into the equation of state tables. We focused in the Mbar-eV pressure-temperature range because it can be access on kJ-scale laser facilities.In order to achieve this, we used the QEOS model, which is simple to use, configurable, and easily modifiable.First, quantum molecular dynamics (QMD) simulations were performed to generate cold compression curve as well as shock compression curves along the principal Hugoniot. Simulations were compared to QEOS model and showed that atomic bond dissociation has an effect on the compressibility. Results from these simulations are then used to parametrize the Grueneisen parameter in order to generate a tabulated equation of state that includes dissociation. It allowed us to show its influence on shock timing in a hydrodynamic simulation.Second, thermodynamic states along the Hugoniot were measured during three experimental campaigns upon the LULI2000 and GEKKO XII laser facilities. Experimental data confirm QMD simulations.This study was performed on two ablator materials which are an undoped polymer CHO, and a silicon-doped polymer CHOSi. Results showed universal shock compression properties. (author) [fr

  13. Materials for high temperature reactor vessels

    International Nuclear Information System (INIS)

    Buenaventura Pouyfaucon, A.

    2004-01-01

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

  14. Criteria for cesium capsules to be shipped as special form radioactive material

    International Nuclear Information System (INIS)

    Lundeen, J.E.

    1994-01-01

    The purpose of this report is to compile all the documentation which defines the criteria for Waste Encapsulation and Storage Facility (WESF) cesium capsules at the IOTECH facility and Applied Radiant Energy Corporation (ARECO) to be shipped as special form radioactive material in the Beneficial Uses Shipping System (BUSS) Cask. The capsules were originally approved as special form in 1975, but in 1988 the integrity of the capsules came into question. WHC developed the Pre-shipment Acceptance Test Criteria for capsules to meet in order to be shipped as special form material. The Department of Energy approved the criteria and directed WHC to ship the capsules at IOTECH and ARECO meeting this criteria to WHC as special form material

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  16. Potential of sago starch/carrageenan mixture as gelatin alternative for hard capsule material

    Science.gov (United States)

    Poeloengasih, Crescentiana Dewi; Pranoto, Yudi; Anggraheni, Frida Dwi; Marseno, Djagal Wiseso

    2017-03-01

    In order to replace gelatin in capsule shell production, blends of sago starch and carrageenan were developed. Films and capsules were prepared with 10% (w/v) of sago starch, 25% (w/w starch) of glycerol and various carrageenan concentration (1, 2, 3% w/w starch) in two different kappa/iota-carrageenan ratio (1:3 and 3:1). The resulted films and capsules were characterized by mechanical property, water vapor and oxygen permeability. In addition, moisture absorption and solubility of capsule in acid solution were investigated. The results reveal that addition of carrageenan makes the films stronger and less permeable. Higher kappa-carrageenan content improved tensile strength and barrier properties of the films, whereas higher iota-carrageenan content produced films with higher elongation, moisture absorption and capsule solubility in acid solution. Capsule with 2% (w/w starch) of carrageenan at kappa-/iota-ratio 3:1 had the lowest moisture absorption, whereas capsule with 3% (w/w starch) of carrageenan at kappa/iota ratio 1:3 had the highest solubility. It is illustrated that sago starch/carrageenan blends can be used as hard capsule material.

  17. Development of a sealing process of capsules for surveillance test tubes of the vessel in nuclear power plants

    International Nuclear Information System (INIS)

    Romero C, J.; Fernandez T, F.; Perez R, N.; Rocamontes A, M.; Garcia R, R.

    2007-01-01

    The surveillance capsule is composed by the support, three capsules for impact test tubes, five capsules for tension test tubes and one porta dosemeters. The capsules for test tubes are of two types: rectangular capsule for Charpy test tubes and cylindrical capsule for tension test tubes. This work describes the development of the welding system to seal the capsules for test tubes that should contain helium of ultra high purity to a pressure of 1 atmosphere. (Author)

  18. 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.)

  19. A performance test of a capsule for a material irradiation in the OR holes of HANARO

    International Nuclear Information System (INIS)

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

    2008-01-01

    A test for a pressure drop and a vibration was performed to develop a material capsule for an irradiation at the OR hole in HANARO. It was analyzed before the test that a diameter of a material capsule for the OR holes should be more than 49mm by an evaluation of a flow rate and pressure drop in theory. According to this estimation, 3 kinds of mock-up capsules with a diameter of 52, 54, 56 mm were made and applied to a pressure drop test. As a result of the pressure drop test, the requirement for a pressure and a flow rate in HANARO was confirmed to be satisfied for the 3 kinds of diameters. The capsules with diameters of 54, 56mm were applied to a vibration test by taking into consideration a receptive capacity of the specimens. The capsule with a diameter of 56mm satisfied the requirement for an allowable limit of the vibration acceleration applied in HANARO. The heat transfer coefficient and the temperature on the surface of a capsule were estimated. As the temperature on the surface of the capsule was calculated to be 43.7 .deg. C, the ONB condition in HANARO was satisfied

  20. Capsule shell material impacts the in vitro disintegration and dissolution behaviour of a green tea extract☆

    Science.gov (United States)

    Glube, Natalie; Moos, Lea von; Duchateau, Guus

    2013-01-01

    Purpose In vitro disintegration and dissolution are routine methods used to assess the performance and quality of oral dosage forms. The purpose of the current work was to determine the potential for interaction between capsule shell material and a green tea extract and the impact it can have on the release. Methods A green tea extract was formulated into simple powder-in-capsule formulations of which the capsule shell material was either of gelatin or HPMC origin. The disintegration times were determined together with the dissolution profiles in compendial and biorelevant media. Results All formulations disintegrated within 30 min, meeting the USP criteria for botanical formulations. An immediate release dissolution profile was achieved for gelatin capsules in all media but not for the specified HPMC formulations. Dissolution release was especially impaired for HPMCgell at pH 1.2 and for both HPMC formulations in FeSSIF media suggesting the potential for food interactions. Conclusions The delayed release from studied HPMC capsule materials is likely attributed to an interaction between the catechins, the major constituents of the green tea extract, and the capsule shell material. An assessment of in vitro dissolution is recommended prior to the release of a dietary supplement or clinical trial investigational product to ensure efficacy. PMID:25755998

  1. Capsule shell material impacts the in vitro disintegration and dissolution behaviour of a green tea extract.

    Science.gov (United States)

    Glube, Natalie; Moos, Lea von; Duchateau, Guus

    2013-01-01

    In vitro disintegration and dissolution are routine methods used to assess the performance and quality of oral dosage forms. The purpose of the current work was to determine the potential for interaction between capsule shell material and a green tea extract and the impact it can have on the release. A green tea extract was formulated into simple powder-in-capsule formulations of which the capsule shell material was either of gelatin or HPMC origin. The disintegration times were determined together with the dissolution profiles in compendial and biorelevant media. All formulations disintegrated within 30 min, meeting the USP criteria for botanical formulations. An immediate release dissolution profile was achieved for gelatin capsules in all media but not for the specified HPMC formulations. Dissolution release was especially impaired for HPMCgell at pH 1.2 and for both HPMC formulations in FeSSIF media suggesting the potential for food interactions. The delayed release from studied HPMC capsule materials is likely attributed to an interaction between the catechins, the major constituents of the green tea extract, and the capsule shell material. An assessment of in vitro dissolution is recommended prior to the release of a dietary supplement or clinical trial investigational product to ensure efficacy.

  2. 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.)

  3. Vessel calibration for accurate material accountancy at RRP

    International Nuclear Information System (INIS)

    Yanagisawa, Yuu; Ono, Sawako; Iwamoto, Tomonori

    2004-01-01

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

  4. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  7. A probabilistic method for determining the volume fraction of pre-embedded capsules in self-healing materials

    International Nuclear Information System (INIS)

    Lv, Zhong; Chen, Huisu

    2014-01-01

    Autonomous healing of cracks using pre-embedded capsules containing healing agent is becoming a promising approach to restore the strength of damaged structures. In addition to the material properties, the size and volume fraction of capsules influence crack healing in the matrix. Understanding the crack and capsule interaction is critical in the development and design of structures made of self-healing materials. Assuming that the pre-embedded capsules are randomly dispersed we theoretically model flat ellipsoidal crack interaction with capsules and determine the probability of a crack intersecting the pre-embedded capsules i.e. the self-healing probability. We also develop a probabilistic model of a crack simultaneously meeting with capsules and catalyst carriers in two-component self-healing system matrix. Using a risk-based healing approach, we determine the volume fraction and size of the pre-embedded capsules that are required to achieve a certain self-healing probability. To understand the effect of the shape of the capsules on self-healing we theoretically modeled crack interaction with spherical and cylindrical capsules. We compared the results of our theoretical model with Monte-Carlo simulations of crack interaction with capsules. The formulae presented in this paper will provide guidelines for engineers working with self-healing structures in material selection and sustenance. (paper)

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

    Science.gov (United States)

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

    2012-09-25

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

  9. Reactor Structural Materials: Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chaouadi, R.

    2000-01-01

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

  10. Data package for the Turkey Point material interaction test capsules

    International Nuclear Information System (INIS)

    Krogness, J.C.; Davis, R.B.

    1980-02-01

    Objective of the test is to obtain interaction information on candidate package storage materials and geologies under prototypic temperatures in gamma and low-level neutron fields. This document provides a fabrication record of the experiment

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

    Directory of Open Access Journals (Sweden)

    S.A. Santa

    2017-06-01

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

  12. Crack growth rates in vessel head penetration materials

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  13. Alternative materials for fitting the capsule of radioactive waste storage

    International Nuclear Information System (INIS)

    Cruz, B. de la; Rivas, P.; Hernandez, A.; Marin, C.; Villar, M. V.; Iglesia de la, A.

    1999-01-01

    Phosphate minerals of the apatite group and zeolites (clinoptilolites, mordenite and synthetic are thermally stable, at least up to a six-month period at 200 degree centigree. Both materials have a low distribution coefficient (Rd), which increases significantly when impurities like iodides, chlorides and most of the sulphates are eliminated. The adsorption capacity of the impurity-free phosphates is low, since 70 to 90% of the adsorbed I- in the sorption tests is retained. This implies that most of the adsorbed radionuclides would be retained. The impurity-rich phosphates have low thermal conductivity, good compaction capacity and they do not produce much dust. The porosity and bulk density of phosphates and zeolites are constant as the temperature and heating period increase. The superficial area tends to increase as temperature increases from 100 degree centigree to 150 degree centigree, through no important variations are observed when the temperature reaches 200 degree centigree or the heating period is extended from 3 to 12 months. (Author)

  14. Multilayer Pressure Vessel Materials Testing and Analysis Phase 2

    Science.gov (United States)

    Popelar, Carl F.; Cardinal, Joseph W.

    2014-01-01

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

  15. Investigation of special capsule technologies for material in-pile irradiation test and development plan in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, M. S.; Son, J. M.; Kim, D. S.; Park, S. J.; Cho, Y. G.; Seo, C. K.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    In-pile test for several materials such as Zr alloy, stainless steel, Cr-Ni steel etc. which are used as structural material of the advanced reactor and KNGR(Korea Next Generation Reactor) like SMART, is necessary to produce the design data for developing new reactor materials. Advanced countries like USA, Europe and Japan etc. are not only performing the simple irradiation test for materials, but developing many kinds of special capsule to perform in-pile test having special purpose. For the special test items of fuel rod, fission products, total heat generation, swelling, deformation, sweep gas, temperature ramping and BOCA etc. are being actively concerned. There are capsules measuring creep, fatigue, crack growth, and controlling fluence etc. for special irradiation test of materials. In addition, the advanced countries are developing several instrument technologies suitable for the special capsules. In HANARO, non-instrumented, instrumented material capsules and non-instrumented fuel capsule have been developed and they have been utilized in the irradiation test for users, and creep capsule loading single specimen was made and is planned to test in the reactor soon. For some forthcoming years, special capsules not only measuring creep deformation with multi-specimens, fatigue, controlling fluence but crack propagation and gas sweep considering the requirements of users will be developed in HANARO.

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

    International Nuclear Information System (INIS)

    Koziol, J.J.

    1977-01-01

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

  17. Evaulation of B4C as an ablator material for NIF capsules. Revision 1

    International Nuclear Information System (INIS)

    Burnham, A.K.; Alford, C.S.; Makowiecki, D.M.; Dittrich, T.R.; Wallace, R.J.; Honea, E.C.; King, C.M.; Steinman, D.

    1997-01-01

    Boron carbide (B 4 C) is examined as a potential fuel container and ablator for implosion capsules on the National Ignition Facility (NIF). A capsule of pure B 4 C encasing a layer of solid DT implodes stably and ignites with anticipated NIF x-ray drives, producing 18 MJ of energy. Thin films of B 4 C were found to be resistant to oxidation and modestly transmitting in the infrared (IR), possibly enabling IR fuel characterization and enhancement for thin permeation barriers but not for full-thickness capsules. Polystyrene mandrels 0.5 mm in diameter were successfully coated with 0.15-2.0 micrometers of B 4 C. Thickness estimated from optical density agreed well with those measured by scanning electron microscopy (SEM). The B 4 C microstructure was columnar but finer than for Be made at the same conditions. B 4 C is a very strong material, with a fiber tensile strength capable of holding NIF fill pressures at room temperature, but it is also very brittle, and microscopic flaws or grain structure may limit the noncryogenic fill pressure. Argon (Ar) permeation rates were measured for a few capsules that had been further coated with 5 micrometers of plasma polymer. The B 4 C coatings tended to crack under tensile load. Some shells filled more slowly than they leaked, suggesting that the cracks open and close under opposite pressure loading. As observed earlier for Ti coatings, 0.15-micrometer layers of B 4 C had better gas retention properties than 2-micrometer layers, possibly because of fewer cracks. Permeation and fill strength issues for capsules with a full ablator thickness of B 4 C are unresolved. 21 refs., 6 figs

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

    Science.gov (United States)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  20. In situ mechanical-radiation effects test capsule for simulating fusion material environments

    International Nuclear Information System (INIS)

    Christensen, K.E.; Bennett, G.A.; Sommer, W.F.

    1981-01-01

    Conditions of radiation and simultaneous cyclic stress on materials are inherent in advanced energy source designs such as inertially and magnetically confined controlled thermonuclear reactors. A test capsule capable of applying a cyclic stress to test specimens while they are being irradiated in the 800-MeV proton beam at the Clinton P. Anderson Los Alamos Meson Physics Facility has been developed. The design and performance of this device are discussed in this report. This machine has facilities for seven pairs of differential samples; one sample of a pair receives an applied cyclic stress and its companion in an identical flux will be the unstressed control. Control of the sample temperature and in situ monitoring of sample elongation and load are provided in the design. Results of an earlier experiment will be discussed, along with those of preliminary bench tests of the redesigned capsule

  1. Evaluation of WWER-1000 vessel materials fracture toughness

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-08-01

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

  4. 49 CFR 176.194 - Stowage of Class 1 (explosive) materials on magazine vessels.

    Science.gov (United States)

    2010-10-01

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

  5. Accelerated irradiation test of gundremmingen reactor vessel trepan material

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1992-08-01

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

  6. Accelerated irradiation test of Gundremmingen reactor vessel trepan material

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-08-01

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

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

    International Nuclear Information System (INIS)

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

    2004-11-01

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

  8. Neutron doze distribution in capsules for surveillance of radiation embrittlement of pressure vessel in Krsko nuclear power plant; Porazdelitev nevtronske doze v sondah za kontrolo povecanja krhkosri tlacne posode JE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Najzer, M; Remec, I; Kodeli, I [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1984-07-01

    Calculation of neutron fluence and spectrum distribution in the capsule with samples for radiation embrittlement of PWR pressure vessel surveillance program of Krsko nuclear power plant is presented. Two dimensional computer code DOT 3 has been used and neutron cross sections were taken from DLC-2D library. result is that fluence magnitude in the capsules changes for up to 70%, so when evaluating results of mechanical tests of samples it is necessary to take into account actual position of samples within the capsule. (author)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  11. Clearance potential of ITER vacuum vessel activated materials

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  12. Computer system for International Reactor Pressure Vessel Materials Database support

    International Nuclear Information System (INIS)

    Arutyunjan, R.; Kabalevsky, S.; Kiselev, V.; Serov, A.

    1997-01-01

    This report presents description of the computer tools for support of International Reactor Pressure Vessel Materials Database developed at IAEA. Work was focused on raw, qualified, processed materials data, search, retrieval, analysis, presentation and export possibilities of data. Developed software has the following main functions: provides software tools for querying and search of any type of data in the database; provides the capability to update the existing information in the database; provides the capability to present and print selected data; provides the possibility of export on yearly basis the run-time IRPVMDB with raw, qualified and processed materials data to Database members; provides the capability to export any selected sets of raw, qualified, processed materials data

  13. New sacrificial material for ex-vessel core catcher

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  14. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  15. Modulation of active pharmaceutical material release from a novel 'tablet in capsule system' containing an effervescent blend.

    Science.gov (United States)

    Gohel, Mukesh C; Sumitra G, Manhapra

    2002-02-19

    The objective of the present study was to obtain programmed drug delivery from hard gelatin capsules containing a hydrophilic plug (HPMC or guar gum). The significance of factors such as type of plug (powder or tablet), plug thickness and the formulation of fill material on the release pattern of diltiazem HCl, a model drug, was investigated. The body portion of the hard gelatin capsules was cross-linked by the combined effect of formaldehyde and heat treatment. A linear relationship was observed between weight of HPMC K15M and log % drug released at 4 h from the capsules containing the plug in powder form. In order to accelerate the drug release after a lag time of 4 h, addition of an effervescent blend, NaHCO(3) and citric acid, in the capsules was found to be essential. The plugs of HPMC in tablet form, with or without a water soluble adjuvant (NaCl or lactose) were used for obtaining immediate drug release after the lag time. Sodium chloride did not cause significant influence on drug release whereas lactose favourably affected the drug release. The capsules containing HPMC K15M tablet plug (200 mg) and 35 mg effervescent blend in body portion of the capsule met the selection criteria of less than 10% drug release in 4 h and immediate drug release thereafter. It is further shown that the drug release was also dependant on the type of swellable hydrophilic agent (HPMC or guar gum) and molecular weight of HPMC (K15M or 20 cPs). The results reveal that programmed drug delivery can be obtained from hard gelatin capsules by systemic formulation approach.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  18. Structural materials for ITER in-vessel component design

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    Directory of Open Access Journals (Sweden)

    V. M. Revka

    2013-03-01

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

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

    Science.gov (United States)

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

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

  1. A study on the development of instrumented capsule for the material irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Hwan; Park, J M; Choo, K N; Maeng, W Y; Park, D K; Oh, J M; Park, S J; Jung, S H; Park, J S; Kim, T R; Park, J H; Yang, S Y; Jun, Y K; Yang, S H

    1997-08-01

    Extensive efforts have been made to establish design and manufacturing technology for the instrumented capsule and its related system, which should be compatible with the HANARO`s characteristics. (author). 86 refs., 45 tabs., 146 figs.

  2. 49 CFR 176.166 - Transport of Class 1 (explosive) materials on passenger vessels.

    Science.gov (United States)

    2010-10-01

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

  3. Hot cell examination on the surveillance capsule and HANARO capsule in IMEF

    International Nuclear Information System (INIS)

    Choo, Yong Sun; Oh, Wan Ho; Yoo, Byung Ok; Jung, Yang Hong; Ahn, Sang Bok; Baik, Seung Je; Song, Wung Sup; Hong, Kwon Pyo

    2000-01-01

    For the maintenance of integrity and safety of pressurizer of commercial power plant until its life span, it is required by US NRC 10CFR50 APP. G and H and ASTM E185-94 to periodically monitor irradiation embrittlement by neutron irradiation. In order to accomplished the requirement reactor operator has been carrying out the test by extracting the monitoring capsule embeded in reactor during the period of planned preventive maintenance. In relation to this irradiation samples are being used for prediction of reactor vessel life span and reactor vessel's adjusted reference temperature by irradiation of neutron flux enough to reach to end of life span. And also irradiation capsules with and without instrumentation are used for R and D on nuclear materials. Each capsule contains high radioactivity, therefore, post irradiation examination has to be handled by all means in the hot cell. The facility available for this purpose is Irradiated material examination facility (IMEF) to handle such works as capsule receiving, capsule cut and dismantling, sample classification, various examination, and finally development and improvement of examination equipment and instrumentation. (Hong, J. S.)

  4. Development of remote welding techniques for in-pile IASCC capsules and evaluation of material integrity on capsules for long irradiation period

    Science.gov (United States)

    Shibata, A.; Nakano, J.; Ohmi, M.; Kawamata, K.; Nakagawa, T.; Tsukada, T.

    2012-03-01

    To simulate irradiation assisted stress corrosion cracking (IASCC) behavior by in-pile experiments, it is necessary to irradiate specimens up to a neutron fluence that is higher than the IASCC threshold fluence. Pre-irradiated specimens must be relocated from pre-irradiation capsules to in-pile capsules. Hence, a remote welding machine has been developed. And the integrity of capsule housing for a long term irradiation was evaluated by tensile tests in air and slow strain rate tests in water. Two type specimens were prepared. Specimens were obtained from the outer tubes of capsule irradiated to 1.0-3.9 × 1026 n/m2 (E > 1 MeV). And specimens were irradiated in a leaky capsule to 0.03-1.0 × 1026 n/m2. Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.

  5. Development of remote welding techniques for in-pile IASCC capsules and evaluation of material integrity on capsules for long irradiation period

    International Nuclear Information System (INIS)

    Shibata, A.; Nakano, J.; Ohmi, M.; Kawamata, K.; Nakagawa, T.; Tsukada, T.

    2012-01-01

    To simulate irradiation assisted stress corrosion cracking (IASCC) behavior by in-pile experiments, it is necessary to irradiate specimens up to a neutron fluence that is higher than the IASCC threshold fluence. Pre-irradiated specimens must be relocated from pre-irradiation capsules to in-pile capsules. Hence, a remote welding machine has been developed. And the integrity of capsule housing for a long term irradiation was evaluated by tensile tests in air and slow strain rate tests in water. Two type specimens were prepared. Specimens were obtained from the outer tubes of capsule irradiated to 1.0–3.9 × 10 26 n/m 2 (E > 1 MeV). And specimens were irradiated in a leaky capsule to 0.03–1.0 × 10 26 n/m 2 . Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.

  6. Design and fabrication report on capsule (11M 19K for out of pile test) for irradiation testing of research reactor materials at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.G.; Yang, S.W.; Park, S.J.; Shim, K.T.; Choo, K.N.; Oh, J.M.; Lee, B.C.; Choi, M.H.; Kim, D.J.; Kim, J.M.; Kang, S.H.; Chun, Y.B.; Kim, T.K.; Jeong, Y.H.

    2012-05-15

    As a part of the research reactor development project with a plate type fuel, the irradiation tests of graphite (Gr), beryllium (Be), and zircaloy 4 materials using the capsule have been investigating to obtain the mechanical characteristics such as an irradiation growth, hardness, swelling and tensile strength at the temperature below 100 .deg. C and the 30 MW reactor power. Then, A capsule to be able to irradiate materials(graphite, Be, zircaloy 4) under 100 .deg. C at the HANARO was designed and fabricated. After performing out of pile testing in single channel test loop by using the capsule, the final design of the capsules to be irradiated in CT and IR2 test hole of HANARO was approved, and 2 sets of capsule were fabricated. These capsules will be loaded in CT and IR2 test hole of HANARO, and be started the irradiation from the end of June, 2012. After performing the irradiation testing of 2 sets of capsule, PIE (Post Irradiation Examination) on irradiated specimens (Gr, Be, and zircaloy 4) will be carry out in IMEF (Irradiated Material Examination Facility). So, the irradiation testing will be contributed to obtain the characteristic data induced neutron irradiation on Gr, Be, and zircaloy 4. And then, it is convinced that these data will be also contributed to obtain the license for JRTR (Jordan Research and Training Reactor) and new research reactor in Korea, and export research reactors.

  7. Polydopamine-coated capsules

    Science.gov (United States)

    White, Scott R.; Sottos, Nancy R.; Kang, Sen; Baginska, Marta B.

    2018-04-17

    One aspect of the invention is a polymer material comprising a capsule coated with PDA. In certain embodiments, the capsule encapsulates a functional agent. The encapsulated functional agent may be an indicating agent, healing agent, protecting agent, pharmaceutical drug, food additive, or a combination thereof.

  8. Statues of the study on the irradiated materials by a special capsule in HANARO

    International Nuclear Information System (INIS)

    Kang, Y.-H.; Kim, B.-G.; Cho, M.-S.; Choi, Y.

    2005-01-01

    A special capsule installed with multi-specimens for HANARO has been designed and its parts were fabricated based on the design criteria of sustaining it at the working conditions of 20 n/cm 2 of a fast neutron fluence with energies above 1 MeV and a maximum load of 200 MPa. The special capsule consists of four modules which work independently. Two of them are located in the upper part of the machine and the others are located in the lower part with a 90 degree rotation. Each module has three separate chambers, each of which contains a loading system, various measuring and controlling units. Each module was evaluated by determining the load-displacement curve of four zirconium specimens. Reliable load-displacement curves of the four specimens were obtained by a simultaneous loading at a controlled temperature. (authors)

  9. Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Bros, J.; Ballesteros, A.; Lopez, A.

    1993-01-01

    Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  11. Design, Fabrication, Test Report of the Material Capsule(08M-10K) with Double Thermal Media for High-temperature Irradiation

    International Nuclear Information System (INIS)

    Cho, Man Soon; Choo, K. N.; Kang, Y. H.; Sohn, J. M.; Shin, Y. T.; Park, S. J.; Kim, B. G.; Oh, S. Y.

    2010-01-01

    To overcome the restriction of the irradiation test at a high temperature of the existing material capsule with Al thermal media, a capsule suitable for the irradiation at the high temperature was developed and the performance test was undertaken. The 08M-10K capsule was designed and fabricated as that with double thermal media to verify the structural and external integrity in the high-temperature irradiation higher than 500 .deg. C. The thermal performance test was undertaken at the out-pile test facility, and the soundness of the double thermal media was confirmed with the naked eye after disassembling the capsule. Though the temperatures of the specimens reach 500±20 .deg. C as a result maintaining the capsule during 5 hours after setting the specimens temperatures in the target range, the high-temperature thermal media with double structure was confirmed to maintain the soundness. And the specimens and the thermal media were heated to 600 .deg. C for about 3 minutes, but the thermal media were maintained sound. Especially, the Al thermal media were heated for 5 hours in range of 500±20 .deg. C and for 3 minutes at the temperature of 600 .deg. C. However, the thermal media were confirmed to maintain the soundness. Whether a capsule has only Al thermal media or the high-temperature thermal media with double structure, any capsule can be used in the range of 500±20 .deg. C as the result of this experiment maintaining the specimens high-temperature

  12. Irradiation of reactor materials within projects VISA-2 and 3, 3. Construction and testing of the capsules and containers VISA - Phase I (Part I and II), Vol. I; Ozracivanje reaktorskih materijala po projektima VISA-2 i 3, 3. Osvajanje postupka izrade i ispitivanja kapsula i kenera VISA - I faza (I i II deo), I deo, Album I

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M; Putre, R [Institute of Nuclear Sciences Boris Kidric, Odeljenje za reaktorsku eksperimentalnu tehniku, Vinca, Beograd (Serbia and Montenegro)

    1964-02-15

    The task contains studies of materials for constructing capsules, leak tight capsules and containers made of aluminium and stainless steels. Special attention is devoted to study of welding the closures and claddings of the capsules as well as thermocouples.

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

    International Nuclear Information System (INIS)

    Petrequin, P.

    1995-01-01

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

  14. Physicochemical characterization and oxidative stability of fish oil-loaded electrosprayed capsules: Combined use of whey protein and carbohydrates as wall materials

    DEFF Research Database (Denmark)

    García Moreno, Pedro Jesús; Pelayo, Andres; Yu, Sen

    2018-01-01

    The encapsulation of fish oil in electrosprayed capsules using whey protein and carbohydrates (pullulan and dextran or glucose syrup) mixtures as glassy wall materials was studied. Capsules with fish oil emulsified by using only a rotor-stator emulsification exhibited higher oxidative stability...... than capsules where the oil was emulsified by high-pressure homogenization. Moreover, glucose syrup capsules (with a peroxide value, PV, of 19.7 ± 4.4 meq/kg oil and a content of 1-penten-3-ol of 751.0 ± 69.8 ng/g oil) were less oxidized than dextran capsules after 21 days of storage at 20 °C (PV of 24.......9 ± 0.4 meq/kg oil and 1-penten-3-ol of 1161.0 ± 222.0 ng/g oil). This finding may be attributed to differences in oxygen permeability between both types of capsules. These results indicated the potential of both combinations of whey protein, pullulan, and dextran or glucose syrup as shell materials...

  15. 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

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

    Science.gov (United States)

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

    2017-04-01

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

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

    International Nuclear Information System (INIS)

    Bazant, Z.P.

    1977-01-01

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

  18. Development of neutron fluence measurement and evaluation technology for the test materials in the capsule

    Energy Technology Data Exchange (ETDEWEB)

    Hong, U.; Choi, S. H.; Kang, H. D. [Kyungsan University, Kyungsan (Korea)

    2000-03-01

    The four kinds of the fluence monitor considered by self-shielding are design and fabricated for evaluation of neutron irradiation fluence. They are equipped with dosimeters consisting of Ni, Fe and Ti wires and so forth. The nuclear reaction rate is obtained by measurement on dosimeter using the spectroscopic analysis of induced {gamma}-ray. We established the nuetron fluence evaluating technology that is based on the measurement of the reaction rate considering reactor's irradiation history, burn-out, self-shielding in fluence monitor, and the influence of impurity in dosimeter. The distribution of high energy neutron flux on the vertical axis of the capsule shows fifth order polynomial equation and is good agree with theoretical value in the error range of 30% by MCNP/4A code. 22 refs., 50 figs., 27 tabs. (Author)

  19. The analysis of reactor vessel surveillance program data

    International Nuclear Information System (INIS)

    Norris, E.B.

    1979-01-01

    Commercial nuclear power reactor vessel surveillance programs are provided by the reactor supplier and are designed to meet the requirements of ASTM Method E 185. (3). Each surveillance capsule contains sets of Charpy V-notch (Csub(v)) specimens representing selected materials from the vessel beltline region and some reference steel, tension test specimens machined from selected beltline materials, temperature monitors, and neutron flux dosimeters. Surveillance capsules may also contain fracture mechanics specimens machined from selected vessel beltline materials. The major steps in the conduct of a surveillance program include (1) the testing of the surveillance specimens to determine the exposure conditions at the capsule location and the resulting embrittlement of the vessel steel, (2) the extrapolation of the capsule results to the pressure vessel wall, and (3) the determination of the heatup and cooldown limits for normal, upset, and test operation. This paper will present data obtained from commercial light water reactor surveillance programs to illustrate the methods of analysis currently in use at Southwest Research Institute and to demonstrate some of the limitations imposed by the data available. Details concerning the procedures for testing the surveillance capsule specimens will not be included because they are considered to be outside of the scope of this paper

  20. Processes to Open the Container and the Sample Catcher of the Hayabusa Returned Capsule in the Planetary Material Sample Curation Facility of JAXA

    Science.gov (United States)

    Fujimura, A.; Abe, M.; Yada, T.; Nakamura, T.; Noguchi, T.; Okazaki, R.; Ishibashi, Y.; Shirai, K.; Okada, T.; Yano, H.; hide

    2011-01-01

    Japanese spacecraft Hayabusa, which returned from near-Earth-asteroid Itokawa, successfully returned its reentry capsule to the Earth, the Woomera Prohibited Area in Australia in Jun 13th, 2010, as detailed in another paper [1]. The capsule introduced into the Planetary Material Sample Curation Facility in the Sagamihara campus of JAXA in the early morning of June 18th. Hereafter, we describe a series of processes for the returned capsule and the container to recover gas and materials in there. A transportation box of the recovered capsule was cleaned up on its outer surface beforehand and introduced into the class 10,000 clean room of the facility. Then, the capsule was extracted from the box and its plastic bag was opened and checked and photographed the outer surface of the capsule. The capsule was composed of the container, a backside ablator, a side ablator, an electronic box and a supporting frame. The container consists of an outer lid, an inner lid, a frame for latches, a container and a sample catcher, which is composed of room A and B and a rotational cylinder. After the first check, the capsule was packed in a plastic bag with N2 again, and transferred to the Chofu campus in JAXA, where the X-ray CT instrument is situated. The first X-ray CT analysis was performed on the whole returned capsule for confirming the conditions of latches and O-ring seal of the container. The analysis showed that the latches of the container should have worked normally, and that the double Orings of the container seemed to be sealed its sample catcher with no problem. After the first X-ray CT, the capsule was sent back to Sagamihara and introduced in the clean room to exclude the electronic box and the side ablator from the container by hand tools. Then the container with the backside ablator was set firmly to special jigs to fix the lid of container tightly to the container and set to a milling machine. The backside ablator was drilled by the machine to expose heads of bolts

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

    International Nuclear Information System (INIS)

    Steffen, H.P.

    1987-01-01

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

  2. Weld repair of helium degraded reactor vessel material

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Ning Dong; Yao Weida

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    Kalinin, G.; Barabash, V.

    2000-01-01

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

  7. The anterior lens capsule used as support material in RPE cell-transplantation

    DEFF Research Database (Denmark)

    Nicolini, J; Kiilgaard, Jens Folke; Wiencke, A K

    2000-01-01

    To investigate the use of an ocular basement membrane as support material for transplanted porcine RPE cells.......To investigate the use of an ocular basement membrane as support material for transplanted porcine RPE cells....

  8. Radio frequency identification (RFID) tagging for material tracking and future asset management : research project capsule.

    Science.gov (United States)

    2016-07-01

    Current Louisiana Department of Transportation and Development (DOTD) : practice allows multiple mixtures or materials to be submitted and subsequently : used during a highway construction project. Such materials include mechanically : stabilized ear...

  9. Fiscal 1999 leading research report. Research on advanced functional material technology based on micro- capsule technology; 1999 nendo micro capsule ka gijutsu ni yoru kokinoka zairyo gijutsu no chosa kenkyu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-03-01

    Study was made on production technology of the 5th- generation material 'eco-smart composite structure' based on micro-capsule technology. This material ought to have an intelligent function including self-detection and self- restoration of defects, environmental harmony, high strength and long life. At present, these demands are implemented by embedding a sensor, actuator and controller in structure, or optical fiber, shape memory alloy and computer chip in matrix, however, such complex material structure has adverse effect on environment. On the other hand, single structure material such as biodegradable polymer is unusable in severe environment such as the ocean and space. These problems can be resolved by combining composite material and micro-sphere technology, for example fiber-reinforced plastic (FRP) composite structure including micro-capsules with nano particles. According to targets, this material can have a self-restoration intelligent function, and other environment friendly functions such as easy decomposition, sound insulation, oscillation damping, electromagnetic shielding and hygroscopicity. (NEDO)

  10. Fiscal 1999 leading research report. Research on advanced functional material technology based on micro- capsule technology; 1999 nendo micro capsule ka gijutsu ni yoru kokinoka zairyo gijutsu no chosa kenkyu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-03-01

    Study was made on production technology of the 5th- generation material 'eco-smart composite structure' based on micro-capsule technology. This material ought to have an intelligent function including self-detection and self- restoration of defects, environmental harmony, high strength and long life. At present, these demands are implemented by embedding a sensor, actuator and controller in structure, or optical fiber, shape memory alloy and computer chip in matrix, however, such complex material structure has adverse effect on environment. On the other hand, single structure material such as biodegradable polymer is unusable in severe environment such as the ocean and space. These problems can be resolved by combining composite material and micro-sphere technology, for example fiber-reinforced plastic (FRP) composite structure including micro-capsules with nano particles. According to targets, this material can have a self-restoration intelligent function, and other environment friendly functions such as easy decomposition, sound insulation, oscillation damping, electromagnetic shielding and hygroscopicity. (NEDO)

  11. Development of a sealing process of capsules for surveillance test tubes of the vessel in nuclear power plants; Desarrollo de proceso de sellado de capsulas para probetas de vigilancia de la vasija en nucleoelectricas

    Energy Technology Data Exchange (ETDEWEB)

    Romero C, J.; Fernandez T, F.; Perez R, N.; Rocamontes A, M.; Garcia R, R. [ININ, Km 36.5 Carretera Mexico-Toluca, 52750 La Marquesa, Estado de Mexico (Mexico)

    2007-07-01

    The surveillance capsule is composed by the support, three capsules for impact test tubes, five capsules for tension test tubes and one porta dosemeters. The capsules for test tubes are of two types: rectangular capsule for Charpy test tubes and cylindrical capsule for tension test tubes. This work describes the development of the welding system to seal the capsules for test tubes that should contain helium of ultra high purity to a pressure of 1 atmosphere. (Author)

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    Science.gov (United States)

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

    2012-11-01

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

  14. Capsule Endoscopy

    Science.gov (United States)

    ... because experience with it is limited and traditional upper endoscopy is widely available. Why it's done Your doctor might recommend a capsule endoscopy procedure to: Find the cause of gastrointestinal bleeding. If you have unexplained bleeding in your digestive ...

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

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

    Science.gov (United States)

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

    1998-10-01

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

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

    International Nuclear Information System (INIS)

    Bass, Bennett Richard; Taylor, Nigel

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    International Nuclear Information System (INIS)

    Tas, A.; Teunissen, G.

    1986-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader

    2013-01-01

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

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

    Science.gov (United States)

    Kumar, B. Ramesh; Gangradey, R.

    2012-11-01

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

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

    International Nuclear Information System (INIS)

    Kumar, B Ramesh; Gangradey, R

    2012-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Amir T. Kalali

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

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

    Science.gov (United States)

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

    2000-12-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-12-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Science.gov (United States)

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

    2013-10-01

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

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

    International Nuclear Information System (INIS)

    Yoshida, Kazuo

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.

    1975-01-01

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

  12. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

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

    1974-12-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Sato, I.; Suzuki, K.

    1994-01-01

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

  18. Capsule development and utilization for material irradiation tests; study on the in-pile creep measuring method of zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong; Lee, Byung Kee; Lee, Jong Jea; Kim, Chang Sik; Kim, B. Hun; Cho, I. Sik [Sunmoon University, Asan (Korea)

    2002-02-01

    The final objective of this project is to obtain a design and fabrication technology of an in-pile creep test machine of zirconium alloys. First, design concepts of the in-pile creep test machines of various foreign countries were reviewed and a preliminary design of the equipment was carried. Second, the mock-up of the in-pile creep test machine was fabricated based on the preliminary design. The mock-up consisted of upper and lower grips, a yoke, a pressure chamber including a bellows, a push rod and LVDT. Each part was made of 304 L stainless steel. The average surface roughness of the parts was 1.0-14.7 {mu}m. The mock-up precisely determined an extension of a specimen by gas pressure. Finally, in-pile creep capsule was designed, fabricated and modified. High pure aluminum blocks were put in the capsule. Considering heat transfer coefficients of helium and nitrogen gases, the cooling efficiency is about 4 .deg. C at the condition of 300 .deg. C creep test. Yield strength, ultimate tensile strength and elongation at 300 .deg. C were 335 MPa, 591 MPa, 19.8%, respectively. which were lower than the values at room temperature, 353 MPa, 740 MPa, 12.5%. This study gave an important technology related to design, fabrication and performance tests of the in-pile creep test machine, which is applied to the fabrication of a special capsule and also used for the fundamental data for the fabrication of various in-pile creep capsules. 6 refs., 45 figs., 5 tabs. (Author)

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

    International Nuclear Information System (INIS)

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

    1993-09-01

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

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

    International Nuclear Information System (INIS)

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

    1980-02-01

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

  1. Characterization of Decommissioned PWR Vessel Internals Materials Samples: Material Certification, Fluence, and Temperature (Nonproprietary Version)

    International Nuclear Information System (INIS)

    Krug, M.; Shogan, R.; Fero, A.; Snyder, M.

    2004-01-01

    Pressurized water reactor (PWR) cores, operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs require detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel. This report contains basic material characterization information of the as-installed samples of reactor internals material which were harvested from a decommissioned PWR

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

    International Nuclear Information System (INIS)

    Sattari-Far, Iradj.

    1989-05-01

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

  3. Irradiation of reactor materials within projects VISA-2 and and 3. Construction and testing of the capsules and containers VISA - Phase II; Ozracivanje reaktorskih materijala po projektima VISA-2 i 3, 3. Osvajanje postupka izrade I ispitivanja kapsula i kenera VISA - II faza

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M; Smokovic, Z; Putre, R [Institute of Nuclear Sciences Boris Kidric, Odeljenje za reaktorsku eksperimentalnu tehniku, Vinca, Beograd (Serbia and Montenegro)

    1964-06-15

    The task contains studies of materials for constructing capsules, leak tight capsules and containers made of aluminium and stainless steels. This report contains study of welding, leak testing, corrosion problems vacuuming, quality control of welds. Detailed design specifications for fabrication of capsules and needed equipment are part of this report.

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  5. Experimental and Theoretical Investigations of Ferro Boron as In-vessel Shield Material in FBRs

    International Nuclear Information System (INIS)

    Keshavamurthy, R.S.; Raju, S.; Anthonysamy, S.; Murugan, S.; Sunil Kumar, D.; Rajan Babu, V.; Ravi Chandar, S.C.; Venkiteswaran, C.N.; Chetal, S.C.

    2013-01-01

    Neutron attenuation characteristics of Ferroboron has been investigated both theoretically and experimentally and Fe-B has been found to be favourable for use in FBRs. • Its use will result in significant savings in cost, without impairing shielding capabilities. • Extensive and in-depth out of pile characterization thermophysical properties and high temperature metallurgial compatibility tests with SS 304L clad were carried out. These as well as effects studied of interaction of sodium and Fe-B together on clad at high temparatures show excellent compatibility with clad. • Design of two types Ferro-Boron capsules was done for conducting an irradiation test in FBTR simulating 60 years of neutron fluence in FBRs, • Fabrication of the capsules was successfully carried out. • The capsule was loaded in FBTR and the capsule has been discharged after the intended 45 days of irradiation. • Post Irradiation Examination will be carried out to look for any possible effect due to irradiation

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

    International Nuclear Information System (INIS)

    Widia Lastana Istanto

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Xu Hong; Ma Li; Wang Junrong; Zhou Zhiwei

    2011-01-01

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

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

    Science.gov (United States)

    Czarnecki, Sezin; Duering, Rolf-Alexander

    2013-04-01

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

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

    Science.gov (United States)

    Balzar, Davor

    2010-10-01

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

  10. Diagnostic and therapeutic radio pharmaceutical capsules

    International Nuclear Information System (INIS)

    Haney, T.A.; Wedeking, P.W.; Morcos, N.A.

    1981-01-01

    An improved pharmaceutical radioactive capsule consisting of a non-toxic, water soluble material adapted to being ingested and rapidly disintegrating on contact with fluids of the gastro-intestinal tract is described. Each capsule is provided with filler material supporting a pharmaceutically useful radioactive compound absorbable from the gastro-intestinal tract. The capsule is preferably of gelatin, methyl cellulose or polyvinyl alcohol and the filler is a polyethylene glycol. The radioactive compound may be iodine e.g. sodium radioiodide I-131 or 123. The capsule may also contain a reducing agent e.g. sodium thiosulphate, sulphite, or bisulphite. (author)

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

    International Nuclear Information System (INIS)

    Farrell, K.

    1995-08-01

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

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

    International Nuclear Information System (INIS)

    Behrisch, R.

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  16. 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)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  19. Role of radiation embrittlement in reactor vessel integrity assessment

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  20. Triggered Release from Polymer Capsules

    Energy Technology Data Exchange (ETDEWEB)

    Esser-Kahn, Aaron P. [Univ. of Illinois, Urbana, IL (United States). Beckman Inst. for Advanced Science and Technology and Dept. of Chemistry; Odom, Susan A. [Univ. of Illinois, Urbana, IL (United States). Beckman Inst. for Advanced Science and Technology and Dept. of Chemistry; Sottos, Nancy R. [Univ. of Illinois, Urbana, IL (United States). Beckman Inst. for Advanced Science and Technology and Dept. of Materials Science and Engineering; White, Scott R. [Univ. of Illinois, Urbana, IL (United States). Beckman Inst. for Advanced Science and Technology and Dept. of Aerospace Engineering; Moore, Jeffrey S. [Univ. of Illinois, Urbana, IL (United States). Beckman Inst. for Advanced Science and Technology and Dept. of Chemistry

    2011-07-06

    Stimuli-responsive capsules are of interest in drug delivery, fragrance release, food preservation, and self-healing materials. Many methods are used to trigger the release of encapsulated contents. Here we highlight mechanisms for the controlled release of encapsulated cargo that utilize chemical reactions occurring in solid polymeric shell walls. Triggering mechanisms responsible for covalent bond cleavage that result in the release of capsule contents include chemical, biological, light, thermal, magnetic, and electrical stimuli. We present methods for encapsulation and release, triggering methods, and mechanisms and conclude with our opinions on interesting obstacles for chemically induced activation with relevance for controlled release.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-03-21

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

  2. Scaling effects in spiral capsule robots.

    Science.gov (United States)

    Liang, Liang; Hu, Rong; Chen, Bai; Tang, Yong; Xu, Yan

    2017-04-01

    Spiral capsule robots can be applied to human gastrointestinal tracts and blood vessels. Because of significant variations in the sizes of the inner diameters of the intestines as well as blood vessels, this research has been unable to meet the requirements for medical applications. By applying the fluid dynamic equations, using the computational fluid dynamics method, to a robot axial length ranging from 10 -5 to 10 -2  m, the operational performance indicators (axial driving force, load torque, and maximum fluid pressure on the pipe wall) of the spiral capsule robot and the fluid turbulent intensity around the robot spiral surfaces was numerically calculated in a straight rigid pipe filled with fluid. The reasonableness and validity of the calculation method adopted in this study were verified by the consistency of the calculated values by the computational fluid dynamics method and the experimental values from a relevant literature. The results show that the greater the fluid turbulent intensity, the greater the impact of the fluid turbulence on the driving performance of the spiral capsule robot and the higher the energy consumption of the robot. For the same level of size of the robot, the axial driving force, the load torque, and the maximum fluid pressure on the pipe wall of the outer spiral robot were larger than those of the inner spiral robot. For different requirements of the operating environment, we can choose a certain kind of spiral capsule robot. This study provides a theoretical foundation for spiral capsule robots.

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  4. Pressure Vessel Steel Research: Belgian Activities

    International Nuclear Information System (INIS)

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

    1994-03-01

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

  5. Pressure Vessel Steel Research: Belgian Activities

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-03-01

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

  6. Diagnostic and therapeutic capsules and method of producing

    International Nuclear Information System (INIS)

    Morcos, N.A.; Haney, T.A.; Wedeking, P.W.

    1981-01-01

    An article of manufacture comprising a pharmaceutical radioactive capsule formed essentially of a non-toxic, water soluble material adapted to being ingested and rapidly disintegrating on contract with fluids of the gastro-intestinal tract, and having a filler material supporting a pharmaceutically useful radioactive compound absorbable from the gastro-intestinal tract said filler material being supported by said capsule. And a method of filling a pharmaceutical radioactive capsule comprising providing filler material supporting a pharmaceutically useful radioactive compound and transporting said filler material carrying a pharmaceutically useful radioactive compound into the chamber of said capsule

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-15

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

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

    International Nuclear Information System (INIS)

    Johnson, R.E.

    1981-09-01

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

  12. Posterior capsule opacification.

    Science.gov (United States)

    Wormstone, I Michael; Wang, Lixin; Liu, Christopher S C

    2009-02-01

    Posterior Capsule Opacification (PCO) is the most common complication of cataract surgery. At present the only means of treating cataract is by surgical intervention, and this initially restores high visual quality. Unfortunately, PCO develops in a significant proportion of patients to such an extent that a secondary loss of vision occurs. A modern cataract operation generates a capsular bag, which comprises a proportion of the anterior and the entire posterior capsule. The bag remains in situ, partitions the aqueous and vitreous humours, and in the majority of cases, houses an intraocular lens. The production of a capsular bag following surgery permits a free passage of light along the visual axis through the transparent intraocular lens and thin acellular posterior capsule. However, on the remaining anterior capsule, lens epithelial cells stubbornly reside despite enduring the rigours of surgical trauma. This resilient group of cells then begin to re-colonise the denuded regions of the anterior capsule, encroach onto the intraocular lens surface, occupy regions of the outer anterior capsule and most importantly of all begin to colonise the previously cell-free posterior capsule. Cells continue to divide, begin to cover the posterior capsule and can ultimately encroach on the visual axis resulting in changes to the matrix and cell organization that can give rise to light scatter. This review will describe the biological mechanisms driving PCO progression and discuss the influence of IOL design, surgical techniques and putative drug therapies in regulating the rate and severity of PCO.

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

    International Nuclear Information System (INIS)

    1982-10-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    1994-01-01

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

  16. Surveillance tests for light-water cooled nuclear power reactor vessels in IMEF

    International Nuclear Information System (INIS)

    Choo, Yong-Sun; Ahn, Sang-Bok; Park, Dae-Gyu; Jung, Yang-Hong; Yoo, Byung-Ok; Oh, Wan-Ho; Baik, Seung-Je; Koo, Dae-Seo; Lee, Key-Soon

    1999-01-01

    The surveillance tests for light-water cooled nuclear power reactor vessels were established to monitor the radiation-induced changes in the mechanical properties of ferritic materials in the beltline according to US NRC 10 CFR 50 App. G, US NRC RG1.99-rev.2, ASTM E185-82 and E185-94 in Irradiated Materials Examination Facility(IMEF). The surveillance capsule was transported from NPPs pool sites to KAERI IMEF by using a shipping cask. The capsule was cut and dismantled by capsule cutting machine and milling machine in M2 hot cell. Charpy tests and tension tests were performed in M5a and M5b hot cells respectively. Especially the EPMA located at hot lab was used to analyze the Ni and Cu wt% composition of base metal and weld for predicting the adjusted reference temperature(ART). The established process and test results were summarized in this paper. (author)

  17. The indentation of pressurized elastic shells: from polymeric capsules to yeast cells

    KAUST Repository

    Vella, D.

    2011-08-10

    Pressurized elastic capsules arise at scales ranging from the 10 m diameter pressure vessels used to store propane at oil refineries to the microscopic polymeric capsules that may be used in drug delivery. Nature also makes extensive use of pressurized elastic capsules: plant cells, bacteria and fungi have stiff walls, which are subject to an internal turgor pressure. Here, we present theoretical, numerical and experimental investigations of the indentation of a linearly elastic shell subject to a constant internal pressure. We show that, unlike unpressurized shells, the relationship between force and displacement demonstrates two linear regimes. We determine analytical expressions for the effective stiffness in each of these regimes in terms of the material properties of the shell and the pressure difference. As a consequence, a single indentation experiment over a range of displacements may be used as a simple assay to determine both the internal pressure and elastic properties of capsules. Our results are relevant for determining the internal pressure in bacterial, fungal or plant cells. As an illustration of this, we apply our results to recent measurements of the stiffness of baker\\'s yeast and infer from these experiments that the internal osmotic pressure of yeast cells may be regulated in response to changes in the osmotic pressure of the external medium.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  19. Contribution of materials investigations and operating experience of reactor vessel internals to PWRs' safety, performance and reliability

    International Nuclear Information System (INIS)

    Lemaire, E.; Monteil, N.; Jardin, N.; Doll, M.

    2015-01-01

    The Reactor Pressure Vessel Internals (RVI) include all the components inside the pressure vessel, except the nuclear fuel, the rod cluster assemblies and the instrumentation. The RVI consist of bolted and welded structures that are divided into two sub-assemblies: the upper internals which are removed at every refueling outage and the lower internals which are systematically removed for inspection at every 10-year outage. The main functions of the RVI are to position the core, to support it, and to provide a coolant flow by channeling the fluid. Moreover, the lower internals contribute to a neutron protection of the reactor pressure vessel by absorbing most of the neutron flux from the core. Depending on their location and material composition, the RVI components can face different ageing phenomena, that are actual or potential (such as wear, fatigue, stress corrosion cracking, irradiation assisted stress corrosion cracking, hardening and loss of ductility due to neutron irradiation, irradiation creep and irradiation swelling). EDF has developed a strategy for managing ageing and demonstrating the capacity of the RVI to perform their design functions over 40 years of operation. This overall approach is periodically revisited to take into account the most recent knowledge obtained from the following main topics: Safety Analyses, Research-Development programs, In-Service Inspection (ISI) results, Maintenance programs and Metallurgical Examinations. Based on continuous improvements in those fields, the goal of this paper is to present the way that materials investigations and operating experience obtained on RVI are managed by EDF to improve RVI safety, performance and reliability. It is shown that a perspective of 60 years of operation of RVI components is supported by large Research-Development efforts combined with field experience. (authors)

  20. Multiple shell pressure vessel

    International Nuclear Information System (INIS)

    Wedellsborg, B.W.

    1988-01-01

    A method is described of fabricating a pressure vessel comprising the steps of: attaching a first inner pressure vessel having means defining inlet and outlet openings to a top flange, placing a second inner pressure vessel, having means defining inlet and outlet opening, concentric with and spaced about the first inner pressure vessel and attaching the second inner pressure vessel to the top flange, placing an outer pressure vessel, having inlet and outlet openings, concentric with and spaced apart about the second inner pressure vessel and attaching the outer pressure vessel to the top flange, attaching a generally cylindrical inner inlet conduit and a generally cylindrical inner outlet conduit respectively to the inlet and outlet openings in the first inner pressure vessel, attaching a generally cylindrical outer inlet conduit and a generally cylindrical outer outlet conduit respectively to the inlet and outlet opening in the second inner pressure vessel, heating the assembled pressure vessel to a temperature above the melting point of a material selected from the group, lead, tin, antimony, bismuth, potassium, sodium, boron and mixtures thereof, filling the space between the first inner pressure vessel and the second inner pressure vessel with material selected from the group, filling the space between the second inner pressure vessel and the outer pressure vessel with material selected from the group, and pressurizing the material filling the spaces between the pressure vessels to a predetermined pressure, the step comprising: pressurizing the spaces to a pressure whereby the wall of the first inner pressure vessel is maintained in compression during steady state operation of the pressure vessel

  1. Stochastic Capsule Endoscopy Image Enhancement

    Directory of Open Access Journals (Sweden)

    Ahmed Mohammed

    2018-06-01

    Full Text Available Capsule endoscopy, which uses a wireless camera to take images of the digestive tract, is emerging as an alternative to traditional colonoscopy. The diagnostic values of these images depend on the quality of revealed underlying tissue surfaces. In this paper, we consider the problem of enhancing the visibility of detail and shadowed tissue surfaces for capsule endoscopy images. Using concentric circles at each pixel for random walks combined with stochastic sampling, the proposed method enhances the details of vessel and tissue surfaces. The framework decomposes the image into two detailed layers that contain shadowed tissue surfaces and detail features. The target pixel value is recalculated for the smooth layer using similarity of the target pixel to neighboring pixels by weighting against the total gradient variation and intensity differences. In order to evaluate the diagnostic image quality of the proposed method, we used clinical subjective evaluation with a rank order on selected KID image database and compared it to state-of-the-art enhancement methods. The result showed that the proposed method provides a better result in terms of diagnostic image quality and objective quality contrast metrics and structural similarity index.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  3. SATCAP-C : a program for thermal hydraulic design of pressurized water injection type capsule

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Someya, Hiroyuki; Asoh, Tomokazu; Niimi, Motoji

    1992-10-01

    There are capsules called 'Pressure Water Injection Type Capsule' as a kind of irradiation devices at the Japan Materials Testing Reactor (JMTR). A type of the capsules is a 'Boiling Water Capsule' (usually named BOCA). The other type is a 'Saturated Temperature Capsule' (named SATCAP). When the water is kept at a constant pressure, the water temperature does not become higher than the saturated temperature so far as the water does not fully change to steam. These type capsules are designed on the basis of the conception of applying the water characteristic to the control of irradiation temperature of specimens in the capsules. In designing of the capsules in which the pressurized water is injected, thermal performances have to be understood as exactly as possible. It is not easy however to predict thermal performances such as axially temperature distribution of water injected in the capsule, because there are heat-sinks at both side of inner and outer of capsule casing as the result that the water is fluid. Then, a program (named SATCAP-C) for the BOCA and SATCAP was compiled to grasp the thermal performances in the capsules and has been used the design of the capsules and analysis of the data obtained from some actual irradiation capsules. It was confirmed that the program was effective in thermal analysis for the capsules. The analysis found out the values for heat transfer coefficients at various surfaces of capsule components and some thermal characteristics of capsules. (author)

  4. Magnetically guided capsule endoscopy.

    Science.gov (United States)

    Shamsudhin, Naveen; Zverev, Vladimir I; Keller, Henrik; Pane, Salvador; Egolf, Peter W; Nelson, Bradley J; Tishin, Alexander M

    2017-08-01

    Wireless capsule endoscopy (WCE) is a powerful tool for medical screening and diagnosis, where a small capsule is swallowed and moved by means of natural peristalsis and gravity through the human gastrointestinal (GI) tract. The camera-integrated capsule allows for visualization of the small intestine, a region which was previously inaccessible to classical flexible endoscopy. As a diagnostic tool, it allows to localize the sources of bleedings in the middle part of the gastrointestinal tract and to identify diseases, such as inflammatory bowel disease (Crohn's disease), polyposis syndrome, and tumors. The screening and diagnostic efficacy of the WCE, especially in the stomach region, is hampered by a variety of technical challenges like the lack of active capsular position and orientation control. Therapeutic functionality is absent in most commercial capsules, due to constraints in capsular volume and energy storage. The possibility of using body-exogenous magnetic fields to guide, orient, power, and operate the capsule and its mechanisms has led to increasing research in Magnetically Guided Capsule Endoscopy (MGCE). This work shortly reviews the history and state-of-art in WCE technology. It highlights the magnetic technologies for advancing diagnostic and therapeutic functionalities of WCE. Not restricting itself to the GI tract, the review further investigates the technological developments in magnetically guided microrobots that can navigate through the various air- and fluid-filled lumina and cavities in the body for minimally invasive medicine. © 2017 American Association of Physicists in Medicine.

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

    International Nuclear Information System (INIS)

    Marston, T.U.; Server, W.

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    National Research Council Canada - National Science Library

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

    2008-01-01

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

  10. Thermal characteristic test for saturated temperature type capsule

    International Nuclear Information System (INIS)

    Niimi, Motoji; Someya, Hiroyuki; Kobayashi, Toshiki; Ohuchi, Mitsuo; Harayama, Yasuo

    1989-08-01

    The Japan Material Testing Reactor Project is developing a new type capsule so-called 'Saturated Temperature Capsule', as a part of irradiation technique improvement program. This type capsule, in which the water is supplied and boiled, bases on the conception of keeping the coolant at the saturated temperature and facilitating the temperature setting of specimens heated by gamma-ray in reactor. However, out-pile test was planned, because there were few usable data for design and operation of the capsule into which the coolant was injected. A out-pile apparatus, simulated the capsule with electric heaters, was fabricated and experiments were carried out, to obtain data concerning design and operation for the capsule into which the water was injected. As a structure of simulated capsule, a type of downward coolant supply was adopted. The downward coolant tube type injectes the water in the bottom of capsule by tube through the upper flange. Major objects of experiences were to grasp thermal features under operation and to provide performances of capsule control equipment. Experimental results proved that the temperature of water within the capsule was easily varied by controlling supply water flow rate, and that the control equipment was operated stably and safety. (author)

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

    International Nuclear Information System (INIS)

    Schaaf, B. van der

    2000-01-01

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

  12. 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

  13. Vessel Operating Units (Vessels)

    Data.gov (United States)

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

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

    International Nuclear Information System (INIS)

    Naus, D.J.

    1981-05-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Garrett, G.G.

    1975-01-01

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

  18. NIF capsule performance modeling

    Directory of Open Access Journals (Sweden)

    Weber S.

    2013-11-01

    Full Text Available Post-shot modeling of NIF capsule implosions was performed in order to validate our physical and numerical models. Cryogenic layered target implosions and experiments with surrogate targets produce an abundance of capsule performance data including implosion velocity, remaining ablator mass, times of peak x-ray and neutron emission, core image size, core symmetry, neutron yield, and x-ray spectra. We have attempted to match the integrated data set with capsule-only simulations by adjusting the drive and other physics parameters within expected uncertainties. The simulations include interface roughness, time-dependent symmetry, and a model of mix. We were able to match many of the measured performance parameters for a selection of shots.

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

    International Nuclear Information System (INIS)

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

    2002-10-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    Science.gov (United States)

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

    2016-04-01

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

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

    Science.gov (United States)

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

    2016-11-01

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

  3. 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.

  4. Production, deformation and mechanical investigation of magnetic alginate capsules

    Science.gov (United States)

    Zwar, Elena; Kemna, Andre; Richter, Lena; Degen, Patrick; Rehage, Heinz

    2018-02-01

    In this article we investigated the deformation of alginate capsules in magnetic fields. The sensitivity to magnetic forces was realised by encapsulating an oil in water emulsion, where the oil droplets contained dispersed magnetic nanoparticles. We solved calcium ions in the aqueous emulsion phase, which act as crosslinking compounds for forming thin layers of alginate membranes. This encapsulating technique allows the production of flexible capsules with an emulsion as the capsule core. It is important to mention that the magnetic nanoparticles were stable and dispersed throughout the complete process, which is an important difference to most magnetic alginate-based materials. In a series of experiments, we used spinning drop techniques, capsule squeezing experiments and interfacial shear rheology in order to determine the surface Young moduli, the surface Poisson ratios and the surface shear moduli of the magnetically sensitive alginate capsules. In additional experiments, we analysed the capsule deformation in magnetic fields. In spinning drop and capsule squeezing experiments, water droplets were pressed out of the capsules at elevated values of the mechanical load. This phenomenon might be used for the mechanically triggered release of water-soluble ingredients. After drying the emulsion-filled capsules, we produced capsules, which only contained a homogeneous oil phase with stable suspended magnetic nanoparticles (organic ferrofluid). In the dried state, the thin alginate membranes of these particles were rather rigid. These dehydrated capsules could be stored at ambient conditions for several months without changing their properties. After exposure to water, the alginate membranes rehydrated and became flexible and deformable again. During this swelling process, water diffused back in the capsule. This long-term stability and rehydration offers a great spectrum of different applications as sensors, soft actuators, artificial muscles or drug delivery systems.

  5. Production of 131I gelatin capsules

    International Nuclear Information System (INIS)

    Freud, A.; Hirshfeld, N.; Canfi, A.; Melamud, Y.

    1997-01-01

    Radioiodine ( 131 I) hard-gelatin capsules are widely used for the diagnosis and treatment of various thyroid disorders. Until 1980 radioiodine was supplied by us as a liquid dosage. This proved to be a rather inconvenient form since it resulted in inaccurate dosing by the physicians and caused frequent contamination of the patients and the hospital personnel. In an attempt to overcome these problems we have designed and constructed a production facility for capsules in which 1311 is packaged. Because of the extreme precautions necessary in handling radioactive compounds, encapsulation of radioactive materials requires specifically designed production techniques, special instrumentation and unique quality control procedures that are not encountered in the standard capsule production processes in the pharmaceutical industry

  6. Fabrication, characterization and evaluation of bacterial cellulose-based capsule shells for oral drug delivery

    DEFF Research Database (Denmark)

    Ullah, Hanif; Badshah, Munair; Mäkilä, Ermei

    2017-01-01

    Bacterial cellulose (BC) was investigated for the first time for the preparation of capsule shells for immediate and sustained release of drugs. The prepared capsule shells were characterized using X-ray diffraction, scanning electron microscopy and Fourier transform infrared spectroscopy. The BC...... to gelatin capsules with both immediate and sustained drug release properties depending upon the compositions of the encapsulated materials....

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

    Directory of Open Access Journals (Sweden)

    Vebil Yıldırım

    2017-07-01

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

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

    Directory of Open Access Journals (Sweden)

    Tatjana Gluhak

    2016-10-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Dickson, T.L.

    1993-07-01

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

  11. High temperature radioisotope capsule

    International Nuclear Information System (INIS)

    Bradshaw, G.B.

    1976-01-01

    A high temperature radioisotope capsule made up of three concentric cylinders, with the isotope fuel located within the innermost cylinder is described. The innermost cylinder has hemispherical ends and is constructed of a tantalum alloy. The intermediate cylinder is made of a molybdenum alloy and is capable of withstanding the pressure generated by the alpha particle decay of the fuel. The outer cylinder is made of a platinum alloy of high resistance to corrosion. A gas separates the innermost cylinder from the intermediate cylinder and the intermediate cylinder from the outer cylinder

  12. Comparative assessment of hepatic Glisson's capsule and bovine pericardium in heart valve bioprostheses.

    Science.gov (United States)

    Kagramanov, I I; Kokshenev, I V; Dobrova, N B; Kastava, V T; Serov, R A; Zaets, S B

    1998-05-01

    The optimal material for heart valve bioprostheses remains disputable. This investigation was initiated to compare the properties of hepatic Glisson's capsule, clinical experience of which in cardiovascular surgery is minimal, with those of bovine pericardium. Hepatic Glisson's capsule was harvested from bull calves and used to create composite pulmonary arterial monocusp grafts and bioprostheses. Comparison of the strength and elastic properties of Glisson's capsule and bovine pericardium, as well as the hydrodynamic characteristics of valves made from these materials, was performed. Late results of operations using these materials were estimated echocardiographically. Although Glisson's capsule tissue is thinner than the bovine pericardium, its elasticity modulus is greater. However, the hydrodynamic characteristics of heart valves made from either tissue are similar. Moreover, valves made from Glisson's capsule have a lower systolic pressure gradient on the prosthesis and a higher effective orifice area. Composite pulmonary arterial xenopericardial grafts with a monocusp of Glisson's capsule were used in 30 patients during tetralogy of Fallot repair. Glisson's capsule was also used for tricuspid valve reconstruction and as a bioprosthesis in six patients with Ebstein's anomaly. At 1-2 years after surgery, the Glisson's capsule tissue remained thin and flexible, with no calcification. Although the hydrodynamic properties of hepatic Glisson's capsule and the bovine pericardium are similar, the capsule tissue is thinner and has a greater elasticity modulos. Thus, Glisson's capsule may be used for bioprosthesis construction both independently and in combination with bovine pericardium.

  13. Research vessels

    Digital Repository Service at National Institute of Oceanography (India)

    Rao, P.S.

    The role of the research vessels as a tool for marine research and exploration is very important. Technical requirements of a suitable vessel and the laboratories needed on board are discussed. The history and the research work carried out...

  14. High-gain capsule design for the HIDIF project

    International Nuclear Information System (INIS)

    Honrubia, J.J.; Cerrada, J.A.; Gomez, R.

    2000-01-01

    A high-gain capsule has been designed for the HIDIF project. The goal has been to relax the accelerator requirements by using a radiation pulse with lower peak temperature (220 eV) than previous designs (260 eV). The ablator material is beryllium doped with a very low concentration (0.2 atom %) of copper. The capsule absorbs 1.3 MJ and yields, approximately, 450 MJ in I-D simulations. The effect of the opacity of the ablator on capsule performance has been studied in detail. (authors)

  15. Pressurized Vessel Slurry Pumping

    International Nuclear Information System (INIS)

    Pound, C.R.

    2001-01-01

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

  16. Recent experiences and problems in conducting pressure vessel surveillance examinations

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1979-01-01

    Each of the commercial power reactors in the U.S.A. has a pressure vessel surveillance program. The purpose of the programs is to monitor the effects of radiation on the mechanical properties on the steel pressure vessels. A program for a given reactor includes a series of irradiation capsules containing neutron dosimeters and mechanical property specimens. The capsules are periodically removed during the life of the reactor and evaluated. The surveillance capsule examinations conducted to date have been valuable in assessing the effects of radiation on pressure vessels. However, a number of problems have been observed in the course of capsule examinations which potentially could reduce the maximum value of the data obtained. These problems are related to specimen design and preparation, capsule design and preparation, capsule installation and removal, capsule disassembly, specimen testing and evaluation, program documentation, and quality assurance. Examples of problems encountered in the preceding areas are presented in the present paper, and recommendations are made for minimization or prevention of these problems in future programs. Included in the recommendations is that appropriate ASTM standards, ASME Boiler and Pressure Vessel Code sections, and NRC regulations provide the appropriate framework for prevention of problems

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  18. Aqueous corrosion in static capsule tests representing multi-metal assemblies in the hydraulic circuit of Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Lipa, M. [Association Euratom-CEA, CEA/DSM/DRFC, Centre de Cadarache, 13108 Saint-Paul-Lez-Durance (France)], E-mail: manfred.lipa@cea.fr; Blanchet, J.; Feron, D. [CEA/DEN/SCCME, Centre de Saclay, 91191 Gif sur Yvette (France); Cellier, F. [AREVA ANP, Centre Technique, 71380 Saint Marcel (France)

    2008-12-15

    Tore supra (TS) in vessel components represent a unique combination of metals in the hydraulic circuit. Different materials, e.g. stainless steel, copper alloys, nickel, etc., were joined together by fusion welding, brazing and friction. Since the operation and baking temperature of all in vessel components has been defined to be set at 230 deg. C/40 bars a special water chemistry of the cooling water loop was suggested in order to prevent eventual water leaks due to corrosion at relative high temperatures and pressures in tubes, bellows, coils and coolant plant ancillary equipments. Following experiences with water chemistry in Pressurised Water Reactors, an all volatile chemical treatment (AVT) has been defined for the cooling water quality of TS. Since then, a simplified static (no fluid circulation) corrosion test program at relatively high temperature and pressure has been performed using capsule-type samples made of above mentioned multi-metal assemblies.

  19. H.B. Robinson-2 pressure vessel benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Remec, I.; Kam, F.B.K.

    1998-02-01

    The H. B. Robinson Unit 2 Pressure Vessel Benchmark (HBR-2 benchmark) is described and analyzed in this report. Analysis of the HBR-2 benchmark can be used as partial fulfillment of the requirements for the qualification of the methodology for calculating neutron fluence in pressure vessels, as required by the U.S. Nuclear Regulatory Commission Regulatory Guide DG-1053, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Section 1 of this report describes the HBR-2 benchmark and provides all the dimensions, material compositions, and neutron source data necessary for the analysis. The measured quantities, to be compared with the calculated values, are the specific activities at the end of fuel cycle 9. The characteristic feature of the HBR-2 benchmark is that it provides measurements on both sides of the pressure vessel: in the surveillance capsule attached to the thermal shield and in the reactor cavity. In section 2, the analysis of the HBR-2 benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed with three multigroup libraries based on ENDF/B-VI: BUGLE-93, SAILOR-95 and BUGLE-96. The average ratio of the calculated-to-measured specific activities (C/M) for the six dosimeters in the surveillance capsule was 0.90 {+-} 0.04 for all three libraries. The average C/Ms for the cavity dosimeters (without neptunium dosimeter) were 0.89 {+-} 0.10, 0.91 {+-} 0.10, and 0.90 {+-} 0.09 for the BUGLE-93, SAILOR-95 and BUGLE-96 libraries, respectively. It is expected that the agreement of the calculations with the measurements, similar to the agreement obtained in this research, should typically be observed when the discrete-ordinates method and ENDF/B-VI libraries are used for the HBR-2 benchmark analysis.

  20. Capsule endoscopy: Beyond small bowel

    Directory of Open Access Journals (Sweden)

    Samuel N Adler

    2012-01-01

    Full Text Available In this article the brief and dramatic history of capsule endoscopy of the digestive tract is reviewed. Capsule endoscopy offers a non invasive method to diagnose diseases that affect the esophagus, small bowel and colon. Technological improvements relating to optics, software, data recorders with two way communication have revolutionized this field. These advancements have produced better diagnostic performance.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  2. Thermal analysis of an instrumented capsule using an ANSYS program

    International Nuclear Information System (INIS)

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

    2006-01-01

    An instrumented capsule has been used for an irradiation test of various nuclear materials in the research reactor, HANARO. To obtain the design data of the instrumented capsule, a thermal analysis is performed using a finite element analysis program, ANSYS. The 2-dimensional model for a cross section of the capsule including the specimens is generated, and a gamma-heating rate of the materials for the HANARO power of 24 or 30 MW is considered as an input force. The effect of the gap size and the control rod position on the temperature of the specimens or other components is discussed. From the analysis it is found that the gap between the thermal media and the external tube has a significant effect on the temperature of the specimen. In the case of the material capsule, the maximum temperature for the reactor power of 24 MW is 255degC for an irradiation test and 257degC for a FE analysis at the center stage of the capsule in the axial direction. It is expected that the analysis models using an ANSYS program will be useful in designing the instrumented capsules for an irradiation test and estimating the test results. (author)

  3. Effect of Heat Flux on the Specimen Temperature of an LBE Capsule

    International Nuclear Information System (INIS)

    Kang, Y. H.; Park, S. J.; Cho, M. S.; Choo, K. N.; Lee, Y. S.

    2011-01-01

    For application of high-temperature irradiation tests in the HANARO reactor for Gen IV reactor material development, a number of newly designed LBE capsules have been investigated at KAERI since 2008. Recent study on heat transfer experiment of an LBE capsule with a single heater has shown that the specimen temperature of the mock-up increased linearly with an increase of heat input. The work highlighted only the heat transfer capability of an LBE capsule with a single heater as a simulated specimen in a liquid metal medium. Hence, a new LBE capsule with multi specimen sets has been designed and fabricated for the heat transfer experiment of an LBE capsule of 11M-01K. In this paper, a series of thermal analyses and heat transfer experiments for a newly designed LBE capsule was implemented to study the effect of an increase in the value of heat input and its influence on temperature distribution in the capsule mock-up

  4. Fabrication of Non-instrumented capsule for DUPIC simulated fuel irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.G.; Kang, Y.H.; Park, S.J.; Shin, Y.T. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    In order to develope DUPIC nuclear fuel, the irradiation test for simulated DUPIC fuel was planed using a non-instrumented capsule in HANARO. Because DUPIC fuel is highly radioactive material the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO was designed to remotely assemble and disassemble in hot cell. And then, according to the design requirements the non-instrumented DUPIC capsule was successfully manufactured. Also, the manufacturing technologies of the non-instrumented capsule for irradiating the nuclear fuel in HANARO were established, and the basic technology for the development of the instrumented capsule technology was accumulated. This report describes the manufacturing of the non-instrumented capsule for simulated DUPIC fuel. And, this report will be based to develope the instrumented capsule, which will be utilized to irradiate the nuclear fuel in HANARO. 26 refs., 4 figs. (Author)

  5. Collapse of experimental capsules under external pressure

    International Nuclear Information System (INIS)

    Simonen, F.A.; Shippell, R.J. Jr.

    1980-01-01

    Stress analyses and developmental tests of capsules fabricated from thick-walled tubing were performed for an external pressure design condition. In the design procedure no credit was taken for the expected margin in pressure between yielding of the capsule wall and catastrophic collapse or flattening. In tests of AISI-1018 low carbon steel capsules, a significant margin was seen between yield and collapse pressure. However, the experimental yield pressures were significantly below predictions, essentially eliminating the safety margin present in the conservative design approach. The differences between predictions and test results are attributed to deficiencies in the plasticity theories commonly in use for engineering stress analyses. The results of this study show that the von Mises yield condition does not accurately describe the yield behavior of the AISI-1018 steel tubing material for the triaxial stress conditions of interest. Finite element stress analyses successfully predicted the transition between uniform inward plastic deformation and ovalization that leads to catastrophic collapse. After adjustments to correct for the unexpected yield behavior of the tube material, the predicted pressure-deflection trends were found to follow the experimental data

  6. Irradiation of reactor materials within projects VISA-2 and 3, 3. Procedure for construction and testing the capsules and test-tubes - Phase I (Parts I and II) Part II; Ozracivanje reaktorskih materijala po projektima VISA-2 i 3, 3. Osvajanje postupka izrade i ispitivanja kapsula i kenera VISA - I faza (I i II deo), II deo

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1964-02-15

    Experiments concerned with Projects VISA-2 and 3 demand construction of hermetization test-tubes, irradiation capsules, experimental devices and reactor channels as well as welding of fuel element claddings. For this purpose special materials as stainless steels, aluminium alloys, pure aluminium, magnesium, zirconium were chosen. these materials demand special procedure for welding. This report includes design and construction data with drawings of the special device for semiautomated circular welding.

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

    International Nuclear Information System (INIS)

    1983-01-01

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

  8. Design and fabrication of non-instrumented capsule

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Sung; Lee, Jeong Young; Kim, Joon Yeon; Lee, Sung Ho; Ji, Dae Young; Kim, Suk Hoon; Ahn, Sung Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-04-01

    The use of non-instrumented capsule designed and fabricated in this time is for the evaluation of material irradiation performance, it is to be installed in the inner core of HANARO. The design process of non-instrumented capsule was accomplished by the decision of the quality of material and the shape, thermal analysis, structural analysis. The temperature of the specimen and the stress in capsule during irradiation test was calculated by the thermal analysis and the structural analysis. GGENGTC code and ABAQUS code were used for the calculation of non-instrumented capsule. In case of installing the capsule in irradiation hole, the coolant flow rate and the pressure drop in the hole is changed, which will affect the coolant flow rate of the fuel region. Eventually the coolant flow rate outside capsule have to be restricted to the allowable range. In order to obtain the required pressure drop, the flow rate control mechanism, end plate and orifice ring were used in this test. The test results are compared with 36-element fuel pressure drop data which AECL performed by the SCTR facility.

  9. Failure of the capsule for coated particles irradiation

    International Nuclear Information System (INIS)

    Yamaki, Jikei; Nomura, Yasushi; Nagamatsuya, Takaaki; Yamahara, Takeshi; Sakai, Haruyuki

    1975-10-01

    During operation cycle No. 27 of the JMTR (Japan Material Testing Reactor) on May 20, 1974, leakage of the fission product gas occurred from the capsule 72F-7A, which contained coated particles for the irradiation; the coated particles are for the development of a multi-purpose high temperature gas cooled reactor. The capsule was designed for heat 1600 0 C. Three nickel plates as the heat reflector were sandwiched in between the plates of titanium and zirconium, which were adsorbents for the impurity gases in the cladding tube (Nb-1%Zr). Temperatures of the plates were about 1000 0 C under the irradiation, so one metal diffused into the other metal through interfaces, resulting in the formation of an alloy. Its melting point was lower than those of metals in the capsule. The cladding material Nb-1%Zr was melted by the alloy and finally a pin hole developed through the cladding. The process of failure, design of the capsule, post-irradiation test of the capsule and the failure-reproducing experiment with a mock-up capsule are described. (auth.)

  10. Design and fabrication of non-instrumented capsule

    International Nuclear Information System (INIS)

    Kim, Yong Sung; Lee, Jeong Young; Kim, Joon Yeon; Lee, Sung Ho; Ji, Dae Young; Kim, Suk Hoon; Ahn, Sung Ho

    1995-04-01

    The use of non-instrumented capsule designed and fabricated in this time is for the evaluation of material irradiation performance, it is to be installed in the inner core of HANARO. The design process of non-instrumented capsule was accomplished by the decision of the quality of material and the shape, thermal analysis, structural analysis. The temperature of the specimen and the stress in capsule during irradiation test was calculated by the thermal analysis and the structural analysis. GGENGTC code and ABAQUS code were used for the calculation of non-instrumented capsule. In case of installing the capsule in irradiation hole, the coolant flow rate and the pressure drop in the hole is changed, which will affect the coolant flow rate of the fuel region. Eventually the coolant flow rate outside capsule have to be restricted to the allowable range. In order to obtain the required pressure drop, the flow rate control mechanism, end plate and orifice ring were used in this test. The test results are compared with 36-element fuel pressure drop data which AECL performed by the SCTR facility

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

    Directory of Open Access Journals (Sweden)

    Tae-Sik Cho

    2012-08-01

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

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

    Science.gov (United States)

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

    2012-01-01

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

  13. Radioactive waste processing vessel

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  14. Tempest in a vessel

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-01-01

    As the ASN made some statements about anomalies of carbon content in the EPR vessel bottom and top, the author recalls and comments some technical issues to better understand the information published on this topic. He notably addresses the role of the vessel, briefly indicates its operating conditions, shape and structure, and mechanical components for the top, its material and mechanical properties, and test samples used to assess mechanical properties. He also comments the phenomenon of radio-induced embrittlement, the vessel manufacturing process, and evokes the applicable regulations. He quotes and comments statements made by the ASN and Areva which evoke further assessments of the concerned components

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

    International Nuclear Information System (INIS)

    Scheuer, A.; Gutsmiedl, E.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Gutsmiedl, Erwin

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  18. Soft gelatin capsules (softgels).

    Science.gov (United States)

    Gullapalli, Rampurna Prasad

    2010-10-01

    It is estimated that more than 40% of new chemical entities (NCEs) coming out of the current drug discovery process have poor biopharmaceutical properties, such as low aqueous solubility and/or permeability. These suboptimal properties pose significant challenges for the oral absorption of the compounds and for the development of orally bioavailable dosage forms. Development of soft gelatin capsule (softgel) dosage form is of growing interest for the oral delivery of poorly water soluble compounds (BCS class II or class IV). The softgel dosage form offers several advantages over other oral dosage forms, such as delivering a liquid matrix designed to solubilize and improve the oral bioavailability of a poorly soluble compound as a unit dose solid dosage form, delivering low and ultra-low doses of a compound, delivering a low melting compound, and minimizing potential generation of dust during manufacturing and thereby improving the safety of production personnel. However, due to the very dynamic nature of the softgel dosage form, its development and stability during its shelf-life are fraught with several challenges. The goal of the current review is to provide an in-depth discussion on the softgel dosage form to formulation scientists who are considering developing softgels for therapeutic compounds.

  19. Status of irradiation capsule design

    International Nuclear Information System (INIS)

    Nagata, Hiroshi; Yamaura, Takayuki; Nagao, Yoshiharu

    2013-01-01

    For the irradiation test after the restart of JMTR, further precise temperature control and temperature prediction are required. In the design of irradiation capsule, particularly sophisticated irradiation temperature prediction and evaluation are urged. Under such circumstance, among the conventional design techniques of irradiation capsule, the authors reviewed the evaluation method of irradiation temperature. In addition, for the improvement of use convenience, this study examined and improved FINAS/STAR code in order to adopt the new calculation code that enables a variety of analyses. In addition, the study on the common use of the components for radiation capsule enabled the shortening of design period. After the restart, the authors will apply this improved calculation code to the design of irradiation capsule. (A.O.)

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

    International Nuclear Information System (INIS)

    Vlasichev, G.N.

    1994-01-01

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

  1. Welding of iridium heat source capsule components

    International Nuclear Information System (INIS)

    Mustaleski, T.M.; Yearwood, J.C.; Burgan, C.E.; Green, L.A.

    1991-01-01

    Interplanetary spacecraft have long used radioisotope thermoelectric generators (RTG) to produce power for instrumentation. These RTG produce electrical energy from the heat generated through the radioactive decay of plutonium-238. The plutonium is present as a ceramic pellet of plutonium oxide. The pellet is encapsulated in a containment shell of iridium. Iridium is the material of choice for these capsules because of its compatibility with the plutonium dioxide. The high-energy beam welding (electron beam and laser) processes used in the fabrication of the capsules has not been published. These welding procedures were originally developed at the Mound Laboratories and have been adapted for use at the Oak Ridge Y-12 Plant. The work involves joining of thin material in small sizes to exacting tolerances. There are four different electron beam welds on each capsule, with one procedure being used in three locations. There is also a laser weld used to seal the edges of a sintered frit assembly. An additional electron beam weld is also performed to seal each of the iridium blanks in a stainless steel waster sheet prior to forming. In the transfer of these welding procedures from one facility to another, a number of modifications were necessary. These modifications are discussed in detail, as well as the inherent problems in making welds in material which is only 0.005 in. thick. In summary, the paper discusses the welding of thin components of iridium using the high energy beam processes. While the peculiarities of iridium are pertinent to the discussion, much of the information is of general interest to the users of these processes. This is especially true of applications involving thin materials and high-precision assemblies

  2. Spheroidal and conical shapes of ferrofluid-filled capsules in magnetic fields

    Science.gov (United States)

    Wischnewski, Christian; Kierfeld, Jan

    2018-04-01

    We investigate the deformation of soft spherical elastic capsules filled with a ferrofluid in external uniform magnetic fields at fixed volume by a combination of numerical and analytical approaches. We develop a numerical iterative solution strategy based on nonlinear elastic shape equations to calculate the stretched capsule shape numerically and a coupled finite element and boundary element method to solve the corresponding magnetostatic problem and employ analytical linear response theory, approximative energy minimization, and slender-body theory. The observed deformation behavior is qualitatively similar to the deformation of ferrofluid droplets in uniform magnetic fields. Homogeneous magnetic fields elongate the capsule and a discontinuous shape transition from a spheroidal shape to a conical shape takes place at a critical field strength. We investigate how capsule elasticity modifies this hysteretic shape transition. We show that conical capsule shapes are possible but involve diverging stretch factors at the tips, which gives rise to rupture for real capsule materials. In a slender-body approximation we find that the critical susceptibility above which conical shapes occur for ferrofluid capsules is the same as for droplets. At small fields capsules remain spheroidal and we characterize the deformation of spheroidal capsules both analytically and numerically. Finally, we determine whether wrinkling of a spheroidal capsule occurs during elongation in a magnetic field and how it modifies the stretching behavior. We find the nontrivial dependence between the extent of the wrinkled region and capsule elongation. Our results can be helpful in quantitatively determining capsule or ferrofluid material properties from magnetic deformation experiments. All results also apply to elastic capsules filled with a dielectric liquid in an external uniform electric field.

  3. Development of a reconstitution system of Charpy probes for the surveillance of vessels in nucleo electric plants

    International Nuclear Information System (INIS)

    Romero C, J.; Hernandez, R.; Fernandez, F.; Gonzalez M, A.

    2007-01-01

    This work describes the development of a welding system, for the rebuilding of halves of Charpy test tubes, the rebuilding consists on welding two implants in those ends of these halves of test tubes, in these welding the main requirement is not to alter the mechanical properties in a minimum volume of 1 cm 3 , the rebuilding is medullary in the surveillance programs of the reactor vessel. In these programs, the mechanical state of the vessel is evaluated, for it there are surveillance capsules with a Charpy witness test tubes series, subjected to a neutron flow similar or bigger to that of the vessel. The objective is to evaluate in advance on the vessel fragilization grade its life design. However the number of capsules with the witness test tubes it is only for the plant design life and at the moment the nucleo electric, negotiates an extension of life of these, until for 20 more years, of there the importance of this material witness's that stores the information of the damage accumulated by the neutron flow. This material requires to be taken advantage it after being rehearsed and the normative one settles down as obligatory to qualify the rebuilding process with all the requirements settled down in the ASTM Designation: E 1253-99 'Standard Guide for Reconstitution of irradiated Charpy-Sized Specimens', to obtain other reconstituted Charpy test tubes that are again introduced in the reactor. When being reconstituted the halves of the original test tubes it is obtained double reconstituted Charpy test tubes. Half of the test tubes they are used in the surveillance program of the vessel, with the surpluses test tubes, it can determine the fracture toughness, property of the material used in the extension methodology of life of vessel. (Author)

  4. Some views on mechanical safety of capsules for radioactive waste

    International Nuclear Information System (INIS)

    Nilsson, F.

    1978-02-01

    Three different concepts for encapsulation of used nuclear fuel are discussed with respect to their mechanical safety. In the first concept the burnt out fuel elements are encapsulated into a copper capsule. The material properties of copper are discussed especially with reference to toughness and creep. A simple fracture mechanical analysis shows that the risk for direct fracture is negligible at the actual stress levels. The loads on the capsule are studied and are found to be normally less than 40 MPa (residual stresses). Transient loads that might arise in the handling of the capsule might however be dangerous to its integrity. The next concept is encapsulation of the fuel elements into a sintered aluminium oxide capsule. A fracture probability analysis based on Weibull's statistical fracture theory gives fracture probabilities that are acceptable. Extended studies of this concept, especially of the risk for delayed fracture, is recommended. The last concept is a Pb-Ti capsule for glassed refined fuel. An analysis of the relaxation of internal stresses is performed. The critical point of these capsules appears to be the welds on the titanium shell where the risk for a direct fracture is not neglibigle

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  6. In-vessel tritium

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  7. C5 capsule operation modes analysis

    International Nuclear Information System (INIS)

    Negut, Gh.; Ancuta, Mirela; Stefan, Violeta

    2008-01-01

    This paper is part of the Nuclear Research Institute Program 13 dedicated to 'TRIGA Research Reactor performance enhancing' and its objective is improving the engineering of the structural materials irradiation. The paper raises the knowledge level on C5 capsule irradiation modes and utilizes previous results in order to increase C5 performances. In the paper the irradiation modes to test zirconium yttrium sample are assessed. These tests are proposed by AECL. There are presented the C5 initial conditions and models. Also. there are presented the thermal hydraulic conditions during normal and accidental operation. The results will be used in the C5 safety report. (authors)

  8. Summary Report for Capsule Dry Storage Project

    Energy Technology Data Exchange (ETDEWEB)

    JOSEPHSON, W S

    2003-09-04

    There are 1.936 cesium (Cs) and strontium (Sr) capsules stored in pools at the Waste Encapsulation and Storage Facility (WESF). These capsules will be moved to dry storage on the Hanford Site as an interim measure to reduce risk. The Cs/Sr Capsule Dry Storage Project (CDSP) is conducted under the assumption the capsules will eventually be moved to the repository at Yucca Mountain, and the design criteria include requirements that will facilitate acceptance at the repository. The storage system must also permit retrieval of capsules in the event vitrification of the capsule contents is pursued. A cut away drawing of a typical cesium chloride (CsCI) capsule and the capsule property and geometry information are provided in Figure 1.1. Strontium fluoride (SrF{sub 2}) capsules are similar in design to CsCl capsules. Further details of capsule design, current state, and reference information are given later in this report and its references. Capsule production and life history is covered in WMP-16938, Capsule Characterization Report for Capsule Dry Storage Project, and is briefly summarized in Section 5.2 of this report.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    International Nuclear Information System (INIS)

    Oberpichler, R.; Schnellenbach, G.

    1975-01-01

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

  12. Equilibrium ignition for ICF capsules

    International Nuclear Information System (INIS)

    Lackner, K.S.; Colgate, S.A.; Johnson, N.L.; Kirkpatrick, R.C.; Menikoff, R.; Petschek, A.G.

    1993-01-01

    There are two fundamentally different approaches to igniting DT fuel in an ICF capsule which can be described as equilibrium and hot spot ignition. In both cases, a capsule which can be thought of as a pusher containing the DT fuel is imploded until the fuel reaches ignition conditions. In comparing high-gain ICF targets using cryogenic DT for a pusher with equilibrium ignition targets using high-Z pushers which contain the radiation. The authors point to the intrinsic advantages of the latter. Equilibrium or volume ignition sacrifices high gain for lower losses, lower ignition temperature, lower implosion velocity and lower sensitivity of the more robust capsule to small fluctuations and asymmetries in the drive system. The reduction in gain is about a factor of 2.5, which is small enough to make the more robust equilibrium ignition an attractive alternative

  13. Endurance test for DUPIC capsule

    International Nuclear Information System (INIS)

    Chung, Heung June; Bae, K. K.; Lee, C. Y.; Park, J. M.; Ryu, J. S.

    1999-07-01

    This report presents the pressure drop, vibration and endurance test results for mini-plate fuel rig which were designed fabricately by KAERI. From the pressure drop test results, it is noted that the flow rate across the capsule corresponding to the pressure drop of 200 kPa is measured to be about 9.632 kg/sec. Vibration frequency for the capsule ranges from 14 to 18.5 Hz. RMS (Root Mean Square) displacement for the fuel rig is less than 14 μm, and the maximum displacement is less than 54 μm. Based on the endurance test results, the appreciable fretting wear for the DUPIC capsule was not detected. Oxidation on the support tube is observed, also tiny trace of wear between contact points observed. (author). 4 refs., 10 tabs., 45 figs

  14. TEMPO-oxidized Konjac glucomannan as appliance for the preparation of hard capsules

    NARCIS (Netherlands)

    Chen, Yuying; Zhao, Huiying; Liu, Xianwu; Li, Zusen; Liu, Bin; Wu, Jiande; Shi, Mengxuan; Norde, Willem; Li, Yuan

    2016-01-01

    TEMPO-oxidized Konjac glucomannan (OKGM) was developed as new material for preparing vegetarian hard capsules. OKGM of different degrees of oxidation: DO30%, DO50%, and DO80% were prepared to select optimum DO for capsule formation. FT-IR results proved that the primary alcohol groups on KGM were

  15. CAPSULE REPORT: EVAPORATION PROCESS

    Science.gov (United States)

    Evaporation has been an established technology in the metal finishing industry for many years. In this process, wastewaters containing reusable materials, such as copper, nickel, or chromium compounds are heated, producing a water vapor that is continuously removed and condensed....

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  17. LANL Robotic Vessel Scanning

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-25

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

  18. Capsule HRB-15B postirradiation examination report

    International Nuclear Information System (INIS)

    Ketterer, J.W.; Bullock, R.E.

    1981-06-01

    Capsule HRB-15B design tested 184 thin graphite trays containing unbonded fuel particles to peak exposures of 6.6 x 10 25 n/m 2 (E > 29 fJ)/sub HTGR/ fast fluence, approx. 27% fissions per initial metal atom (FIMA) fissile burnup, and 6% FIMA fertile burnup at nominal time-averaged temperatures of 815 to 915 0 C. The capsule tested a variety of low-enriched uranium (approx. 19.5% U-235) fissile particle types, including UC 2 , UC/sub x/O/sub y/, UO 2 , zirconium-buffered UO 2 (referred to in this report as UO 2 /sup *), and 1:1(Th,U)O 2 with both TRISO and silicon-BISO coatings. All fertile particles were ThO 2 with BISO, silicon-BISO, or TRISO coatings. The findings indicated that all TRISO particles retained virtually all of their fission product inventories, except small quantities of silver, at these irradiation temperatures, while some of the silicon-BISO particles released significant amounts of both silver and cesium. No kernel migration, pressure vessel, or outer pyrolytic carbon (OPyC) failures were observed in the fuel particles, which had total diameters of 2 /sup */ particles exhibited no detrimental irradiation effects, but they contained pure carbon precipitates in the kernels after irradiation which were not observed in the undoped UO 2 particles. Postirradiation examination revealed no differences in the irradiation performance of three UC/sub x/O/sub y/ kernel types with varying oxygen/uranium ratios

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

    International Nuclear Information System (INIS)

    Krompholz, K.; Ullrich, G.

    1985-01-01

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

  20. Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

    International Nuclear Information System (INIS)

    Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

    2011-01-01

    Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  2. An O antigen capsule modulates bacterial pathogenesis in Shigella sonnei.

    Science.gov (United States)

    Caboni, Mariaelena; Pédron, Thierry; Rossi, Omar; Goulding, David; Pickard, Derek; Citiulo, Francesco; MacLennan, Calman A; Dougan, Gordon; Thomson, Nicholas R; Saul, Allan; Sansonetti, Philippe J; Gerke, Christiane

    2015-03-01

    Shigella is the leading cause for dysentery worldwide. Together with several virulence factors employed for invasion, the presence and length of the O antigen (OAg) of the lipopolysaccharide (LPS) plays a key role in pathogenesis. S. flexneri 2a has a bimodal OAg chain length distribution regulated in a growth-dependent manner, whereas S. sonnei LPS comprises a monomodal OAg. Here we reveal that S. sonnei, but not S. flexneri 2a, possesses a high molecular weight, immunogenic group 4 capsule, characterized by structural similarity to LPS OAg. We found that a galU mutant of S. sonnei, that is unable to produce a complete LPS with OAg attached, can still assemble OAg material on the cell surface, but a galU mutant of S. flexneri 2a cannot. High molecular weight material not linked to the LPS was purified from S. sonnei and confirmed by NMR to contain the specific sugars of the S. sonnei OAg. Deletion of genes homologous to the group 4 capsule synthesis cluster, previously described in Escherichia coli, abolished the generation of the high molecular weight OAg material. This OAg capsule strongly affects the virulence of S. sonnei. Uncapsulated knockout bacteria were highly invasive in vitro and strongly inflammatory in the rabbit intestine. But, the lack of capsule reduced the ability of S. sonnei to resist complement-mediated killing and to spread from the gut to peripheral organs. In contrast, overexpression of the capsule decreased invasiveness in vitro and inflammation in vivo compared to the wild type. In conclusion, the data indicate that in S. sonnei expression of the capsule modulates bacterial pathogenesis resulting in balanced capabilities to invade and persist in the host environment.

  3. An O antigen capsule modulates bacterial pathogenesis in Shigella sonnei.

    Directory of Open Access Journals (Sweden)

    Mariaelena Caboni

    2015-03-01

    Full Text Available Shigella is the leading cause for dysentery worldwide. Together with several virulence factors employed for invasion, the presence and length of the O antigen (OAg of the lipopolysaccharide (LPS plays a key role in pathogenesis. S. flexneri 2a has a bimodal OAg chain length distribution regulated in a growth-dependent manner, whereas S. sonnei LPS comprises a monomodal OAg. Here we reveal that S. sonnei, but not S. flexneri 2a, possesses a high molecular weight, immunogenic group 4 capsule, characterized by structural similarity to LPS OAg. We found that a galU mutant of S. sonnei, that is unable to produce a complete LPS with OAg attached, can still assemble OAg material on the cell surface, but a galU mutant of S. flexneri 2a cannot. High molecular weight material not linked to the LPS was purified from S. sonnei and confirmed by NMR to contain the specific sugars of the S. sonnei OAg. Deletion of genes homologous to the group 4 capsule synthesis cluster, previously described in Escherichia coli, abolished the generation of the high molecular weight OAg material. This OAg capsule strongly affects the virulence of S. sonnei. Uncapsulated knockout bacteria were highly invasive in vitro and strongly inflammatory in the rabbit intestine. But, the lack of capsule reduced the ability of S. sonnei to resist complement-mediated killing and to spread from the gut to peripheral organs. In contrast, overexpression of the capsule decreased invasiveness in vitro and inflammation in vivo compared to the wild type. In conclusion, the data indicate that in S. sonnei expression of the capsule modulates bacterial pathogenesis resulting in balanced capabilities to invade and persist in the host environment.

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

    International Nuclear Information System (INIS)

    Kussmaul, K.; Ewald, J.

    1977-01-01

    The fracture safe assessment of nuclear pressure vessels (RPV) is based upon: (i) an adequate stress analysis, (ii) reliable material characteristics, (iii) acceptable defects sizes. There may arise problems which are related to the inhomogeneity, low toughness and crack phenomena sometimes observed in the base material and heat affected zone (HAZ). Due to this it is difficult and in some respects even impossible to measure the decisive values of (fracture-) toughness and defects. Apart from the significance of those facts for existing RPVs, all efforts were directed to provide a steel which should be non-susceptible to embrittlement and/or cracking in the HAZ and simultaneously yielding in a higher upper shelf toughness of base and HAZ material. These objections were pursued in cooperation with manufacturers, vendors and inspection authorities by the following activities. (i) Detailed investigations to obtain information on: occurrence and size of inhomogeneities and defects, especially stress relief cracking (SCR), toughness properties adjacent to defects; (ii) improvement of: chemical composition, steel making processes, welding procedures, optimum temperature cycle and level for stress relief heat-treatment. In order to solve these tasks it was necessary to develop additional tools and to correlate all partial results which were newly elaborated. (Auth.)

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  6. Simulations of indirectly driven gas-filled capsules at the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Weber, S. V.; Casey, D. T.; Eder, D. C.; Pino, J. E.; Smalyuk, V. A.; Remington, B. A.; Rowley, D. P.; Yeamans, C. B.; Tipton, R. E.; Barrios, M.; Benedetti, R.; Berzak Hopkins, L.; Bleuel, D. L.; Bond, E. J.; Bradley, D. K.; Caggiano, J. A.; Callahan, D. A.; Cerjan, C. J.; Clark, D. S.; Divol, L. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); and others

    2014-11-15

    Gas-filled capsules imploded with indirect drive on the National Ignition Facility have been employed as symmetry surrogates for cryogenic-layered ignition capsules and to explore interfacial mix. Plastic capsules containing deuterated layers and filled with tritium gas provide a direct measure of mix of ablator into the gas fuel. Other plastic capsules have employed DT or D{sup 3}He gas fill. We present the results of two-dimensional simulations of gas-filled capsule implosions with known degradation sources represented as in modeling of inertial confinement fusion ignition designs; these are time-dependent drive asymmetry, the capsule support tent, roughness at material interfaces, and prescribed gas-ablator interface mix. Unlike the case of cryogenic-layered implosions, many observables of gas-filled implosions are in reasonable agreement with predictions of these simulations. Yields of TT and DT neutrons as well as other x-ray and nuclear diagnostics are matched for CD-layered implosions. Yields of DT-filled capsules are over-predicted by factors of 1.4–2, while D{sup 3}He capsule yields are matched, as well as other metrics for both capsule types.

  7. Safety Analysis Report for Primary Capsule of Ir-192 Radiation Source

    International Nuclear Information System (INIS)

    Lee, J. C.; Bang, K. S.; Choi, W. S.; Seo, K. S.; Son, K. J.; Park, W. J.

    2008-12-01

    All of the source capsules to transport a special form radioactive material should be designed and fabricated in accordance with the design criteria prescribed in IAEA standards and domestic regulations. The objective of this project is to prove the safety of a primary capsule for Ir-192 radiation source which produced in the HANARO. The safety tests of primary capsules were carried out for the impact, percussion and heat conditions. And leakage tests were carried out before and after the each tests. The capsule showed slight scratches and their deformations were not found after each tests. It also met the allowable limits of leakage rate after each test. Therefore, it has been verified that the capsule was designed and fabricated to meet all requirements for the special form radioactive materials

  8. Improvement and utilization of irradiation capsule technology in HANARO

    International Nuclear Information System (INIS)

    Choo, Kee-Nam; Cho, Man-Soon; Kim, Bong-Goo; Lee, Cheol-Yong; Yang, Sung-Woo; Shin, Yoon-Taek; Park, Seng-Jae; Jung, Hoan-Sung

    2012-01-01

    Several improvements of irradiation capsule technology regarding irradiation test parameters, such as temperature and neutron flux/fluence, and regarding instrumentation have progressed at HANARO since the last KAERI-JAERI joint seminar held in 2008. The standard HANARO capsule technology that was developed for use in a commercial power plant temperature of about 300degC was improved to apply to a temperature range of 100-1000degC for the irradiation test of materials of new research reactors and future nuclear systems. Low-flux and long-term irradiation technologies have been developed at HANARO. As a beginning step of the localization of capsule instrumentation technology, the irradiation performance of a domestically produced thermocouple and LVDT will be examined at HANARO. The accuracy of an evaluation of neutron fluence and precise welding technology are also being examined at HANARO. Based on these accumulated capsule technologies, a HANARO irradiation capsule system is being actively utilized for the national R and D programme on commercial nuclear reactors and nuclear fuel cycle technology in Korea. HANARO has recently started the irradiation support of R and D relevant to future nuclear systems including SMART, VHTR, and SFR, and HANARO is preparing new support relevant to new research and Fusion reactors. (author)

  9. Sensitive Detection of Deliquescent Bacterial Capsules through Nanomechanical Analysis.

    Science.gov (United States)

    Nguyen, Song Ha; Webb, Hayden K

    2015-10-20

    Encapsulated bacteria usually exhibit strong resistance to a wide range of sterilization methods, and are often virulent. Early detection of encapsulation can be crucial in microbial pathology. This work demonstrates a fast and sensitive method for the detection of encapsulated bacterial cells. Nanoindentation force measurements were used to confirm the presence of deliquescent bacterial capsules surrounding bacterial cells. Force/distance approach curves contained characteristic linear-nonlinear-linear domains, indicating cocompression of the capsular layer and cell, indentation of the capsule, and compression of the cell alone. This is a sensitive method for the detection and verification of the encapsulation status of bacterial cells. Given that this method was successful in detecting the nanomechanical properties of two different layers of cell material, i.e. distinguishing between the capsule and the remainder of the cell, further development may potentially lead to the ability to analyze even thinner cellular layers, e.g. lipid bilayers.

  10. SrF2 capsule design for heat engine applications

    International Nuclear Information System (INIS)

    Lester, D.H.

    1976-04-01

    A number of design changes were considered to improve heat transfer characteristics of the WESF capsule. This capsule was evaluated in a design concept for use as a heat source in a helium-working fluid, Stirling heat engine. Throughout the study a heat block concept was used. The helium was assumed to be at 1200 0 F and 200 atm. The upper temperature limit at the fuel-metal interface was assumed to be 800 0 C because of material compatibility considerations. A 0.6-in. thick outer can was considered since it may be required for impact resistance and high pressure accident environments. The modifications considered were: (1) filling all gaps with helium rather than air, (2) filling gaps with powdered metal, and (3) adding a third can to the existing capsule. Also, enhancement of emissivity on metal surfaces was considered as a possible modification

  11. An effective surveillance strategy for reactor pressure vessel assessment in the long term operation perspective

    International Nuclear Information System (INIS)

    Chaouadi, R.; Gerard, R.

    2015-01-01

    The reactor pressure vessel (RPV) irradiation embrittlement is monitored by means of surveillance capsules containing the RPV belt-line materials, inserted inside the reactor pressure vessel (RPV) before the start of operation. These capsules are placed at location where they receive a higher neutron flux than the vessel wall, by a factor of the order of 2 to 3. They are regularly retrieved and tested to evaluate the RPV irradiation embrittlement according to specific regulatory procedures and standards, in order to guarantee the safe operation of the RPV throughout its lifetime. These procedures are often relying on empirical but conservative concepts. In parallel, material research reactor (MTR) irradiations are often used to support the surveillance data and to develop a better understanding of irradiation effects, not only qualitatively but also quantitatively. Taking advantage of the increased understanding of irradiation effects, analytical tools were developed to improve the evaluation embrittlement and quality assurance of the RPV embrittlement assessment. In this framework, an alternative but complementary surveillance program assessment was developed in Belgium, the so-called enhanced surveillance, in order to benefit from the latest developments in the area of materials science and irradiation effects. The neutron flux and fracture properties of the surveillance materials can be reliably characterized and correlated to each other using physically-based rather than empirical concepts. The enhanced surveillance approach is complementary to the mandatory regulatory procedure and allows quantifying the conservatism of the regulatory approach. The enhanced surveillance approach that uses the reconstitution technology to fabricate additional small size specimens, appropriate modeling tools and microstructural examination when required, makes it possible to rationalize all available information in a physically-based way

  12. Development and application of a material law for steel-fibre-reinforced concrete with regard to its use for pre-stressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.; Borgerhoff, M.

    1995-01-01

    On the basis of the evaluation of many publications on the mechanical behaviour of steel fibre reinforced concrete (SFRC) and on the results of experiments using an SFRC especially developed for pre-stressed concrete reactor vessels (PCRVs), a material law for SFRC including general multiaxial stress conditions has been developed. From fibre pull-out tests described in the literature and by use of the experimental results, relations describing the capable tensile stress in SFRC after cracking, as a function of crack width, have been derived. There is a significant increase in the biaxial compressive strength of SFRC compared with plain concrete. The improved behaviour under multiaxial stress conditions, with one of the principal stresses being tensile, is outlined in comparison with different formulations of failure envelopes of plain concrete. For the purpose of verifying the material law implemented in the computer program used, analyses have been carried out for experiments with SFRC beams. After some modification concerning the shear behaviour, load-displacement curves and realistic crack propagations which correspond well have been obtained. In the stand-tube area in the centre of a PCRV top cap the use of SFRC is advantageous because of the difficulties concerning the arrangement of reinforcement in the concrete between the tubes. (orig.)

  13. Capsule safety analysis of PRTF irradiation facility

    International Nuclear Information System (INIS)

    Suwarto

    2013-01-01

    Power Ramp Test Facility (PRTF) is an irradiation facility used for fuel testing of power reactor. PRTF has a capsule which is a test fuel rod container. During operation, pressurized water of 160 bars flows through in the capsule. Due to the high pressure it should be analyzed the impact of the capsule on reactor core safety. This analysis has purpose to calculate the ability of capsule pressure capacity. The analysis was carried out by calculating pressure capacity. From the calculating results it can be concluded that the capsule with pressure capacity of 438 bars will be safe to prevent the operation pressure of PRTF. (author)

  14. Probing cell internalisation mechanics with polymer capsules.

    Science.gov (United States)

    Chen, Xi; Cui, Jiwei; Ping, Yuan; Suma, Tomoya; Cavalieri, Francesca; Besford, Quinn A; Chen, George; Braunger, Julia A; Caruso, Frank

    2016-10-06

    We report polymer capsule-based probes for quantifying the pressure exerted by cells during capsule internalisation (P in ). Poly(methacrylic acid) (PMA) capsules with tuneable mechanical properties were fabricated through layer-by-layer assembly. The P in was quantified by correlating the cell-induced deformation with the ex situ osmotically induced deformation of the polymer capsules. Ultimately, we found that human monocyte-derived macrophage THP-1 cells exerted up to approximately 360 kPa on the capsules during internalisation.

  15. Light Water Reactor-Pressure Vessel Surveillance project computer system

    International Nuclear Information System (INIS)

    Merriman, S.H.

    1980-10-01

    A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes

  16. Containment vessel

    International Nuclear Information System (INIS)

    Zbirohowski-Koscia, K.F.; Roberts, A.C.

    1980-01-01

    A concrete containment vessel for nuclear reactors is disclosed that is spherical and that has prestressing tendons disposed in first, second and third sets, the tendons of each set being all substantially concentric and centred around a respective one of the three orthogonal axes of the sphere; the tendons of the first set being anchored at each end at a first anchor rib running around a circumference of the vessel, the tendons of the second set being anchored at each end at a second anchor rib running around a circumference of the sphere and disposed at 90 0 to the first rib, and the tendons of the third set being anchored some to the first rib and the remainder to the second rib. (author)

  17. Status for development of a capsule and instruments for high-temperature irradiation in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Choo, Kee Nam; Lee, Chul Yong; Yang, Seong Woo; Shim, Kyue Taek; Chung, Hwan-Sung [Korea Atomic Energy Research Institute, Taejeon (Korea, Republic of)

    2012-03-15

    As the reactors planned in the Gen-IV program will be operated at high temperature and under high neutron flux, the requirements for irradiation of materials at high temperature are recently being gradually increased. The irradiation tests of materials in HANARO up to the present have been performed usually at temperatures below 300degC at which the RPV materials of the commercial reactors are being operated. To overcome the restriction for high-temperature use of Al thermal media of the existing standard capsule, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated as a more advanced capsule than the single thermal media capsule. (author)

  18. Contraindications for video capsule endoscopy

    OpenAIRE

    Bandorski, Dirk; Kurniawan, Niehls; Baltes, Peter; Hoeltgen, Reinhard; Hecker, Matthias; Stunder, Dominik; Keuchel, Martin

    2016-01-01

    Video capsule endoscopy (VCE) has been applied in the last 15 years in an increasing field of applications. Although many contraindications have been put into perspective, some precautions still have to be considered. Known stenosis of the gastrointestinal tract is a clear contraindication for VCE unless surgery is already scheduled or at least has been considered as an optional treatment modality. In patients with a higher incidence of stenosis, as in an established diagnosis of Crohn?s dise...

  19. Computational evaluation of the constraint loss on the fracture toughness of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Serrano Garcia, M.

    2007-01-01

    The Master Curve approach is included on the ASME Code through some Code Cases to assess the reactor pressure vessel integrity. However, the margin definition to be added is not defined as is the margin to be added when the Master Curve reference temperature T 0 is obtained by testing pre-cracked Charpy specimens. The reason is that the T 0 value obtained with this specimen geometry is less conservative than the value obtained by testing compact tension specimens possible due to a loss of constraint. The two parameter fracture mechanics, considered as an extension of the classical fracture mechanics, coupled to a micromechanical fracture models is a valuable tool to assess the effect of constraint loss on fracture toughness. The definition of a parameter able to connect the fracture toughens value to the constraint level on the crack tip will allow to quantify margin to be added to the T 0 value when this value is obtained testing the pre-cracked Charpy specimens included in the surveillance capsule of the reactor pressure vessel. The Nuclear Regulatory Commission (NRC) define on the To value obtained by testing compact tension specimens and ben specimens (as pre-cracked Charpy are) bias. the NRC do not approved any of the direct applications of the Master Curve the reactor pressure vessel integrity assessment until this bias will be quantified in a reliable way. the inclusion of the bias on the integrity assessment is done through a margin to be added. In this thesis the bias is demonstrated an quantified empirical and numerically and a generic value is suggested for reactor pressure vessel materials, so that it can be used as a margin to be added to the T 0 value obtained by testing the Charpy specimens included in the surveillance capsules. (Author) 111 ref

  20. Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

    International Nuclear Information System (INIS)

    Haggag, F.M.; Nanstad, R.K.; Byrne, S.T.

    1991-01-01

    The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low [121 to 210 degrees C (250--410 degrees F)] compared to those for commercial light-water reactors (LWRs) [∼288 degrees C (550 degrees F)]. The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 x10 18 neutron/cm 2 [0.68 x 10 18 neutron/cm 2 (>1 MeV)] at temperatures of 288, 204, 163, and 121 degrees C (550, 400, 325, and 250 degrees F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs

  1. Anterior Lens Capsule and Iris Thicknesses in Pseudoexfoliation Syndrome.

    Science.gov (United States)

    Batur, Muhammed; Seven, Erbil; Tekin, Serek; Yasar, Tekin

    2017-11-01

    The aim of this study was to evaluate anatomic properties of the lens capsule and iris by anterior segment optical coherence tomography (AS-OCT) in patients with pseudoexfoliation (PEX). This prospective study included 62 eyes of 62 patients with PEX syndrome and 43 eyes of 43 age- and gender-matched controls. All subjects underwent full ophthalmologic examinations including AS-OCT. Pupillary diameter, midperipheral stromal iris thickness, central and temporal lens capsule thicknesses, and peripheral pseudoexfoliation material thickness on the anterior lens capsule surface were measured and recorded. Mean age was 66.8 ± 9.3 years in the PEX group and 65.5 ± 8.9 years in the control group (p = 0.44). The PEX group consisted of 62 patients: 38 men (61.3%) and 24 women (38.7%); the control group included 43 subjects: 25 men (58.1%) and 18 women (41.9%). Pupillary diameter after pharmacologic mydriasis was 21% smaller in the PEX group than controls. Mean midperipheral iris thickness was 36 ± 7.2 μm (7.8%) thinner in the PEX group than that of control group (p = 0.047). The central anterior capsule was a mean of 3.40 ± 0.51 μm (18%) thicker in the PEX group compared to the control group (p = 0.0001). The temporal anterior lens capsule was a mean of 0.17 ± 0.15 μm thicker in the PEX group compared to the control group (p = 0.81). With high-resolution OCT imaging, it has become possible to evaluate the anterior lens capsule without histologic examination and demonstrate that it is thicker than normal in PEX patients.

  2. First beryllium capsule implosions on the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kline, J. L.; Yi, S. A.; Simakov, A. N.; Olson, R. E.; Wilson, D. C.; Kyrala, G. A.; Perry, T. S.; Batha, S. H.; Zylstra, A. B. [Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States); Dewald, E. L.; Tommasini, R.; Ralph, J. E.; Strozzi, D. J.; MacPhee, A. G.; Callahan, D. A.; Hinkel, D. E.; Hurricane, O. A.; Milovich, J. L.; Rygg, J. R.; Khan, S. F. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); and others

    2016-05-15

    The first indirect drive implosion experiments using Beryllium (Be) capsules at the National Ignition Facility confirm the superior ablation properties and elucidate possible Be-ablator issues such as hohlraum filling by ablator material. Since the 1990s, Be has been the preferred Inertial Confinement Fusion (ICF) ablator because of its higher mass ablation rate compared to that of carbon-based ablators. This enables ICF target designs with higher implosion velocities at lower radiation temperatures and improved hydrodynamic stability through greater ablative stabilization. Recent experiments to demonstrate the viability of Be ablator target designs measured the backscattered laser energy, capsule implosion velocity, core implosion shape from self-emission, and in-flight capsule shape from backlit imaging. The laser backscatter is similar to that from comparable plastic (CH) targets under the same hohlraum conditions. Implosion velocity measurements from backlit streaked radiography show that laser energy coupling to the hohlraum wall is comparable to plastic ablators. The measured implosion shape indicates no significant reduction of laser energy from the inner laser cone beams reaching the hohlraum wall as compared with plastic and high-density carbon ablators. These results indicate that the high mass ablation rate for beryllium capsules does not significantly alter hohlraum energetics. In addition, these data, together with data for low fill-density hohlraum performance, indicate that laser power multipliers, required to reconcile simulations with experimental observations, are likely due to our limited understanding of the hohlraum rather than the capsule physics since similar multipliers are needed for both Be and CH capsules as seen in experiments.

  3. Capsule of parotid gland tumor: evaluation by 3.0 T magnetic resonance imaging using surface coils

    International Nuclear Information System (INIS)

    Ishibashi, Mana; Fujii, Shinya; Nishihara, Keisuke; Matsusue, Eiji; Kodani, Kazuhiko; Kaminou, Toshio; Ogawa, Toshihide; Kawamoto, Katsuyuki

    2010-01-01

    Background: Magnetic resonance (MR) imaging of parotid gland tumors has been widely reported, although few reports have evaluated the capsule of parotid gland tumors in detail. Purpose: To evaluate the diagnostic usefulness of 3.0 T MR imaging with surface coils for detection of the parotid gland tumor capsule, and to clarify the characteristics of the capsules. Material and Methods: Seventy-eight patients with parotid gland tumors (63 benign and 15 malignant) were evaluated. Axial and coronal T2-weighted and contrast-enhanced T1-weighted images were obtained using a 3.0 T MR scanner with 70 mm surface coils. It was retrospectively assessed whether each parotid gland tumor was completely surrounded by a capsule. The capsule was classified as regular or irregular in terms of capsular thickness, and as none, mildly, or strongly enhancing in terms of contrast enhancement. Visual interpretations were compared with histopathological findings to evaluate the diagnostic ability of MR imaging to detect parotid gland tumor capsules. Statistical evaluation was conducted concerning the presence of capsules, capsular irregularity, and the difference in contrast enhancement between benign and malignant tumors, and that between pleomorphic adenomas and Warthin's tumors. Results: Capsules completely surrounding the tumor on MR imaging yielded a sensitivity of 87.7% (50/57), specificity of 90.5% (19/21), and accuracy of 88.5% (69/78). Benign tumors had a capsule completely surrounding the tumor significantly more often than malignant tumors (P = 0.009). Concerning capsular irregularity, malignant tumors tended to have more irregular capsules than benign tumors, although there were no significant differences. The capsules of malignant tumors enhanced significantly more strongly than those of benign tumors (P = 0.018). Conclusion: 3.0 T MR imaging using surface coils could correctly depict parotid gland tumor capsules in most cases. Most benign and some malignant tumors had capsules

  4. Method and apparatus for dispensing small quantities of mercury from evacuated and sealed glass capsules

    Science.gov (United States)

    Grossman, Mark W.; George, William A.; Pai, Robert Y.

    1985-01-01

    A technique for opening an evacuated and sealed glass capsule containing a material that is to be dispensed which has a relatively high vapor pressure such as mercury. The capsule is typically disposed in a discharge tube envelope. The technique involves the use of a first light source imaged along the capsule and a second light source imaged across the capsule substantially transversely to the imaging of the first light source. Means are provided for constraining a segment of the capsule along its length with the constraining means being positioned to correspond with the imaging of the second light source. These light sources are preferably incandescent projection lamps. The constraining means is preferably a multiple looped wire support.

  5. Acoustoelastic evaluation of the 20 MnMoNi 55 structural material of the pressure vessels of Angra 2 and 3 nuclear reactors

    International Nuclear Information System (INIS)

    Dutra, Marco Aurelio Monteiro

    2009-01-01

    The pressure vessels of the Angra II and Angra III nuclear power plants, which are of large thickness, have as one of their structural components the 20 MnMo Ni 55 steel. The acoustoelastic evaluation carried out through non-destructive ultrasonic tests was based on the variation of the speed of ultrasonic shear waves of normal incidence as a function of the stress applied in the material using the acoustic birefringence technique, which considers the fractional difference of the speeds of two ultrasonic waves propagating in orthogonal directions. In this work two experiments were carried out. In the first one, a new method for acquisition of ultrasonic signals for joint application with this technique of stress analysis was evaluated by comparison with the conventional one. The aim was to determine its potential and limitations for application in materials up to 120 mm of thickness. In the second experiment, the acoustoelastic behavior of the 20 MnMo Ni 55 steel was studied. The new method of acquiring ultrasonic signals led to satisfactory results and was used in the study of the acoustoelastic behavior of the 20 MnMoNi 55 steel. The acoustoelastic analysis indicated that the material has an anisotropic and heterogeneous behavior. The acoustoelastic constant of this material, obtained experimentally from the analysis of the graphs of the material behavior (birefringence x stress) under compression tests performed on specimens of the 20 MnMo Ni 55 steel, was used for the quantitative stress analysis. A specimen of the 20 MnMo Ni 55 steel was subjected to a bending test to carry out a qualitative stress analysis through the difference between the values of the birefringence obtained during loading (B) and before loading (Bo) in every point studied. With the obtained results it was possible to identify qualitative and quantitatively the regions of the specimen under compressive and tensile stresses. The execution of these tests allowed to verify the efficiency of

  6. Reactor pressure vessel status report

    International Nuclear Information System (INIS)

    Strosnider, J.; Wichman, K.; Elliot, B.

    1994-12-01

    This report gives a brief description of the reactor pressure vessel (RPV), followed by a discussion of the radiation embrittlement of RPV beltline materials and the two indicators for measuring embrittlement, the end-of-license (EOL) reference temperature and the EOL upper-shelf energy. It also summarizes the GL 92-01 effort and presents, for all 37 boiling water reactor plants and 74 pressurized water reactor plants in the United States, the current status of compliance with regulatory requirements related to ensuring RPV integrity. The staff has evaluated the material data needed to predict neutron embrittlement of the reactor vessel beltline materials. These data will be stored in a computer database entitled the reactor vessel integrity database (RVID). This database will be updated annually to reflect the changes made by the licensees in future submittals and will be used by the NRC staff to assess the issues related to vessel structural integrity

  7. Sealing method and sealing device for radioactive waste containing vessel

    International Nuclear Information System (INIS)

    Ishiwatari, Koji; Otsuki, Akira

    1998-01-01

    A radioactive waste-containing body is hoisted down into a strong-material vessel opened upwardly, and a strong-material lid is hoisted down to the opening of the strong-material-vessel and welded. The strong material vessel is hoisted up and loaded on a corrosion resistant-material bottom plate placed horizontally. A corrosion resistant-material vessel having one opening end and having a corrosion resistant-material flange on the other end and previously agreed with the strong material-vessel main body is hoisted up by a hoisting device having an inserting device so that the opening of the corrosion resistant vessel is directed downwardly. The corrosion resistant vessel is press-fitted to the outside of the strong material-vessel by the inserting device while being heated by a preheater to shrink. Subsequently, the lower end of the corrosion resistant-material vessel and the corrosion resistant-material bottom plate are welded to constitute a corrosion resistant-material vessel. Then, the radioactive waste containing body can be sealed in a sealing vessel comprising the strong-material vessel and the corrosion resistant-material vessel. (N.H.)

  8. [Human joint capsule in osteoarthrosis (morpholocical changes) (author's transl)].

    Science.gov (United States)

    Dettmer, N; Barz, B

    1977-07-29

    Morphological investigations of the joint capsules in osteoarthrotic-changed joints have given rise to doubts about the present theory of the causal aetiology of the osteoarthrosis. In every inspected and demonstrated illustration beside the partly normal capsules segments could be found every transition between mild regressive alterations and most massive proliferative changes of the conective tissue and the lining cell layer. It was extraordinary, that the strongly dilated vessels were filled with red blood cells. In another part of the same case was found a massive stricture caused by concentrically deposited substances, which were impregnated with collagenous fibers. Regeneration of the vessels frequently happened adjacent totally obstructed ones. Round-cell infiltrations, granulocytes or other indications of an inflammatory synovitis are found only in a few cases. The intracartilaginous enzymatic reactions, which have been much talked of and which were explained as characteristic of the osteoarthrosis cannot be the cause of the degradation of the cartilage, particularly, because of the normal cell count which is to be found in the synovial fluid. We can answer this problem, if we can prove that the substrates of the chondral metabolism themselves exert a direct or indirect influence on the interstitial connective tissue with induction of the powerful proliferation of the same tissue. The changes in the transit zone would be secondary and their effect on the lining cell layer would increase the progression of the arthrotic events.

  9. Utilization of the capsule out-pile test facilities(2000-2003)

    Energy Technology Data Exchange (ETDEWEB)

    Cho, M. S.; Oh, J. M.; Cho, Y. G. and others

    2003-06-01

    Two out-pile test facilities were installed and being utilized for the non-irradiation tests outside the HANARO. The names of the facilities are the irradiation equipment design verification test facilities and the one-channel flow test device. In these facilities, the performance test of all capsules manufactured before loading in the HANARO and the design verification test for newly developed capsules were performed. The tests in these facilities include loading/unloading, pressure drop, endurance and vibration test etc. of capsules. In the period 2000{approx}2003, the performance tests for 8 material capsules of 99M-01K{approx}02M-05U were carried out, and the design verification tests of creep and fuel capsules developed newly were performed. For development of the creep capsule, pressure drop measurement, operation test of heater, T/C, LVDT and stress loading test were performed. In the design stage of the fuel capsule, the endurance and vibration test besides the above mentioned tests were carried out for verification of the safe operation during irradiation test in the HANARO. And in-chimeny bracket and the capsule supporting system were fixed and the flow tubes and the handling tools were manufactured for use at the facilities.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  11. Photophysics Applied to Cavitands and Capsules.

    Science.gov (United States)

    Berryman, Orion B; Dube, Henry; Rebek, Julius

    2011-07-01

    The use of light as a stimulus to control functional materials or nano-devices is appealing as it provides convenient control of triggering events where and when they are desired without introducing extra components to the system. Many photophysical and photochemical processes are extremely fast, giving rise to nearly instantaneous onset of events. However, these fast processes can be challenging to engineer into chemical systems. Supramolecular chemistry offers a convenient way to study and control photoprocesses. Given the reversible and self-programmed nature of modern host-guest systems, a modular approach can be considered in which different photoprocesses are coupled to obtain complex functions that emerge and are controlled solely by light inputs. In this review, we highlight recent examples of photoswitching and photophysics applied in the context of supramolecular host-guest systems, with a particular emphasis on resorcinarene based cavitands and hydrogen bonded capsules.

  12. Review of the International Atomic Energy Agency International database on reactor pressure vessel materials and US Nuclear Regulatory Commission/Oak Ridge National Laboratory embrittlement data base

    International Nuclear Information System (INIS)

    Wang, J.A.; Kam, F.B.K.

    1998-02-01

    The International Atomic Energy Agency (IAEA) has supported neutron radiation effects information exchange through meetings and conferences since the mid-1960s. Through an International Working Group on Reliability of Reactor Pressure Components, information exchange and research activities were fostered through the Coordinated Research Program (CRP) sponsored by the IAEA. The final CRP meeting was held in November 1993, where it was recommended that the IAEA coordinate the development of an International Database on Reactor Pressure Vessel Material (IDRPVM) as the first step in generating an International Database on Aging Management. The purpose of this study was to provide special technical assistance to the NRC in monitoring and evaluating the IAEA activities in developing the IAEA IDRPVM, and to compare the IDRPVM with the Nuclear Regulatory Commission (NRC) - Oak Ridge National Laboratory (ORNL) Power Reactor Embrittlement Data Base (PR-EDB) and provide recommendations for improving the PR-EDB. A first test version of the IDRPVM was distributed at the First Meeting of Liaison Officers to the IAEA IDRPVM, in November 1996. No power reactor surveillance data were included in this version; the testing data were mainly from CRP Phase III data. Therefore, because of insufficient data and a lack of power reactor surveillance data received from the IAEA IDRPVM, the comparison is made based only on the structure of the IDRPVM. In general, the IDRPVM and the EDB have very similar data structure and data format. One anticipates that because the IDRPVM data will be collected from so many different sources, quality assurance of the data will be a difficult task. The consistency of experimental test results will be an important issue. A very wide spectrum of material characteristics of RPV steels and irradiation environments exists among the various countries. Hence the development of embrittlement prediction models will be a formidable task. 4 refs., 2 figs., 4 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-11-01

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

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

    Science.gov (United States)

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

    2010-11-01

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

  15. The Change of the Seebeck Coefficient Due to Neutron Irradiation and Thermal Fatigue of Nuclear Reactor Pressure Vessel Steel and its Application to the Monitoring of Material Degradation

    International Nuclear Information System (INIS)

    Niffenegger, M.; Reichlin, K.; Kalkhof, D.

    2002-05-01

    The monitoring of material degradation, that might be caused by neutron irradiation and thermal fatigue, is an important topic in lifetime extension of nuclear power plants. We therefore investigated the application of the Seebeck effect for determining material degradation of common reactor pressure vessel steel. The Seebeck coefficient (SC) of several irradiated Charpy specimens made from Japanese JRQ-steel were measured. The specimens suffered a fluence from 0 up to 4.5 x 10 19 neutrons per cm 2 with energies higher than 1 MeV. The measured changes of the SC within this range were about 500 nV, increasing continuously in the range under investigation. Some indications of saturation appeared at fluencies larger than 4.55 x 10 19 neutrons per cm 2 . We obtained a linear dependency between the SC and the temperature shift ΔT 41 of the Charpy-Energy- Temperature curve which is widely used to characterize material embrittlement. Similar measurements were performed on specimens made from the widely used austenitic steel X6CrNiTi18-10 (according to DIN 1.4541) that were fatigued by applying a cyclic strain amplitude of 0.28%. For this kind of fatigue the observed change of SC was somewhat smaller than for the irradiated specimens. Further investigations were made to quantify the size of the gage volume in which the thermoelectric power is generated. It appeared that the information gathered from a Thermo Electric Power (TEP) measurement is very local. To overcome this problem we propose a novel TEP-method using a Thermoelectric Scanning Microscope (TSM). We finally conclude that the change of the SC has a potential for monitoring of material degradation due to neutron irradiation and thermal fatigue, but it has to be taken into account that several influencing parameters could contribute to the TEP in either an additional or extinguishing manner. A disadvantage of the method is the requirement of a clean surface without any oxide layer. A part of this disadvantage can

  16. Uncertainty study of the PWR pressure vessel fluence. Adjustment of the nuclear data base

    International Nuclear Information System (INIS)

    Kodeli, I.A.

    1994-01-01

    The code system devoted to the calculation of the sensitivity and uncertainty of of the neutron flux and reaction rates calculated by the transport codes, has been developed. Adjustment of the basic data to experimental results can be performed as well. Various sources of uncertainties can be taken into account, such as those due to the uncertainties in the cross-sections, response functions, fission spectrum and space distribution of neutron source, geometry and material composition uncertainties... One -As well as two- dimensional analysis can be performed. Linear perturbation theory is applied. The code system is sufficiently general to be used for various analysis in the fields of fission and fusion. The principal objective of our studies concerns the capsule dosimetry study realized in the framework of the 900 MWe PWR pressure vessel surveillance program. The analysis indicates that the present calculations, performed by the code TRIPOLI-2, using the ENDF/B-IV based, non-perturbed neutron cross-section library in 315 energy groups, allows to estimate the neutron flux and the reaction rates in the surveillance capsules and in the most calculated and measured reaction rates permits to reduce these uncertainties. The results obtained with the adjusted iron cross-sections, response functions and fission spectrum show that the agreement between the calculation and the experiment was improved to become within 10% approximately. The neutron flux deduced from the experiment is then extrapolated from the capsule to the most exposed pressure vessel location using the calculated lead factor. The uncertainty in this factor was estimated to be about 7%. (author). 39 refs., 52 figs., 30 tabs

  17. Pressure vessel integrity 1991

    International Nuclear Information System (INIS)

    Bhandari, S.; Doney, R.O.; McDonald, M.S.; Jones, D.P.; Wilson, W.K.; Pennell, W.E.

    1991-01-01

    This volume contains papers relating to the structural integrity assessment of pressure vessels and piping, with special emphasis on nuclear industry applications. The papers were prepared for technical sessions developed under the sponsorship of the ASME Pressure Vessels and Piping Division Committees for Codes and Standards, Computer Technology, Design and Analysis, and Materials Fabrication. They were presented at the 1991 Pressure Vessels and Piping Division Conference in San Diego, California, June 23-27. The primary objective of the sponsoring organization is to provide a forum for the dissemination and discussion of information on development and application of technology for the structural integrity assessment of pressure vessels and piping. This publication includes contributions from authors from Australia, France, Japan, Sweden, Switzerland, the United Kingdom, and the United States. The papers here are organized in six sections, each with a particular emphasis as indicated in the following section titles: Fracture Technology Status and Application Experience; Crack Initiation, Propagation and Arrest; Ductile Tearing; Constraint, Stress State, and Local-Brittle-Zones Effects; Computational Techniques for Fracture and Corrosion Fatigue; and Codes and Standards for Fatigue, Fracture and Erosion/Corrosion

  18. NASPGHAN Capsule Endoscopy Clinical Report.

    Science.gov (United States)

    Friedlander, Joel A; Liu, Quin Y; Sahn, Benjamin; Kooros, Koorosh; Walsh, Catharine M; Kramer, Robert E; Lightdale, Jenifer R; Khlevner, Julie; McOmber, Mark; Kurowski, Jacob; Giefer, Matthew J; Pall, Harpreet; Troendle, David M; Utterson, Elizabeth C; Brill, Herbert; Zacur, George M; Lirio, Richard A; Lerner, Diana G; Reynolds, Carrie; Gibbons, Troy E; Wilsey, Michael; Liacouras, Chris A; Fishman, Douglas S

    2017-03-01

    Wireless capsule endoscopy (CE) was introduced in 2000 as a less invasive method to visualize the distal small bowel in adults. Because this technology has advanced it has been adapted for use in pediatric gastroenterology. Several studies have described its clinical use, utility, and various training methods but pediatric literature regarding CE is limited. This clinical report developed by the Endoscopic and Procedures Committee of the North American Society of Pediatric Gastroenterology, Hepatology and Nutrition outlines the current literature, and describes the recommended current role, use, training, and future areas of research for CE in pediatrics.

  19. Herniation of the anterior lens capsule

    Directory of Open Access Journals (Sweden)

    Pereira Nolette

    2007-01-01

    Full Text Available Herniation of the anterior lens capsule is a rare abnormality in which the capsule bulges forward in the pupillary area. This herniation can be mistaken for an anterior lenticonus where both the capsule and the cortex bulge forward. The exact pathology behind this finding is still unclear. We report the clinical, ultrasound biomicroscopy (UBM and histopathological findings of a case of herniation of the anterior lens capsule. UBM helped to differentiate this entity from anterior lenticonus. Light microscopy revealed capsular splitting suggestive of capsular delamination and collection of fluid (aqueous in the area of herniation giving it a characteristic appearance.

  20. A disposable and multifunctional capsule for easy operation of microfluidic elastomer systems

    International Nuclear Information System (INIS)

    Thorslund, Sara; Läräng, Thomas; Kreuger, Johan; Nguyen, Hugo; Barkefors, Irmeli

    2011-01-01

    The global lab-on-chip and microfluidic markets for cell-based assays have been predicted to grow considerably, as novel microfluidic systems enable cell biologists to perform in vitro experiments at an unprecedented level of experimental control. Nevertheless, microfluidic assays must, in order to compete with conventional assays, be made available at easily affordable costs, and in addition be made simple to operate for users having no previous experience with microfluidics. We have to this end developed a multifunctional microfluidic capsule that can be mass-produced at low cost in thermoplastic material. The capsule enables straightforward operation of elastomer inserts of optional design, here exemplified with insert designs for molecular gradient formation in microfluidic cell culture systems. The integrated macro–micro interface of the capsule ensures reliable connection of the elastomer fluidic structures to an external perfusion system. A separate compartment in the capsule filled with superabsorbent material is used for internal waste absorption. The capsule assembly process is made easy by integrated snap-fits, and samples within the closed capsule can be analyzed using both inverted and upright microscopes. Taken together, the capsule concept presented here could help accelerate the use of microfluidic-based biological assays in the life science sector. (technical note)

  1. Effects of lipophilic components on the compatibility of lipid-based formulations with hard gelatin capsules.

    Science.gov (United States)

    Chen, Feng-Jing; Etzler, Frank M; Ubben, Johanna; Birch, Amy; Zhong, Li; Schwabe, Robert; Dudhedia, Mayur S

    2010-01-01

    the capsules took much longer time (7 weeks) for ranking the predicted capsule deformation at 40 degrees C. Asides from the time savings, activity determination can be considered to be material sparing by offering capsule-fill compatibility assessment even without the need for preparing liquid-filled capsules once appropriate positive and negative references are established. With further optimization, this approach should enable high throughput screening of LBDDS for capsule-fill compatibility in liquid-filled capsule development.

  2. Capsule physics comparison of different ablators for NIF implosion designs

    Science.gov (United States)

    Clark, Daniel; Kritcher, Andrea; Yi, Austin; Zylstra, Alex; Haan, Steven; Ralph, Joseph; Weber, Christopher

    2017-10-01

    Indirect drive implosion experiments on the Naitonal Ignition Facility (NIF) have now tested three different ablator materials: glow discharge polymer (GDP) plastic, high density carbon (HDC), and beryllium. How do these different ablator choices compare in current and future implosion experiments on NIF? What are the relative advantages and disadvantages of each? This talk compares these different ablator options in capsule-only simulations of current NIF experiments and proposed future designs. The simulations compare the impact of the capsule fill tube, support tent, and interface surface roughness for each case, as well as all perturbations in combination. According to the simulations, each ablator is impacted by the various perturbation sources differently, and each material poses unique challenges in the pursuit of ignition. This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

  3. Preparation of functions of computer code GENGTC and improvement for two-dimensional heat transfer calculations for irradiation capsules

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Someya, Hiroyuki; Ito, Haruhiko.

    1992-11-01

    Capsules for irradiation tests in the JMTR (Japan Materials Testing Reactor), consist of irradiation specimens surrounded by a cladding tube, holders, an inner tube and a container tube (from 30mm to 65mm in diameter). And the annular gaps between these structural materials in the capsule are filled with liquids or gases. Cooling of the capsule is done by reactor primary coolant flowing down outside the capsule. Most of the heat generated by fission in fuel specimens and gamma absorption in structural materials is directed radially to the capsule container outer surface. In thermal performance calculations for capsule design, an one(r)-dimensional heat transfer computer code entitled (Generalyzed Gap Temperature Calculation), GENGTC, originally developed in Oak Ridge National Laboratory, U.S.A., has been frequently used. In designing a capsule, are needed many cases of parametric calculations with respect to changes materials and gap sizes. And in some cases, two(r,z)-dimensional heat transfer calculations are needed for irradiation test capsules with short length fuel rods. Recently the authors improved the original one-dimensional code GENGTC, (1) to simplify preparation of input data, (2) to perform automatic calculations for parametric survey based on design temperatures, ect. Moreover, the computer code has been improved to perform r-z two-dimensional heat transfer calculation. This report describes contents of the preparation of the one-dimensional code GENGTC and the improvement for the two-dimensional code GENGTC-2, together with their code manuals. (author)

  4. Phenomenological vessel burst investigations

    International Nuclear Information System (INIS)

    Hippelein, K.W.; Julisch, P.; Muz, J.; Schiedermaier, J.

    1985-07-01

    Fourteen burst experiments have been carried out using vessels with circumferential and longitudinal flaws, for investigation of the fracture behaviour, i.e. the time-related fracture opening. The vessels had dimensions (outer diameter x wall thickness = 800 x 47 mm) which correspond to the dimensions of the main coolant piping of a 1300 MW e PWR. The test specimens had been made of the base-safe material 20 MnMoNi 55 and of a special, 22 NiMoCr 37 base alloy. The experimental conditions with regard to pressure and temperature have been chosen so as to correspond to normal operating conditions of a PWR (p∝17.5 MPa, T∝300 0 C), i.e. the flaws have been so dimensioned that failure was to be expected at a pressure of p∝17.5 MPa. As a rule, water has been used as the pressure medium, or in some cases air, in order to influence the time-dependent pressure decrease. Fluid and structural dynamics calculations have also been made. In order to determine the impact of a fast propagating crack on the leak-to-fracture curve, which normally is defined by quasistationary experiments, suitable tests have been made with large-volume, cylindrical vessels (outer diameter x wall thickness x length = 3000 x 21 x 14000 mm) made of the material WSt E 43. The leak-before-fracture criterion has been confirmed. (orig./HP) [de

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  6. Validation of the quality control method for sodium dicloxacillin in Dicloxen capsules

    International Nuclear Information System (INIS)

    Rodriguez Hernandez, Yaslenis; Perez Navarro, Maikel; Suarez Perez, Yania

    2014-01-01

    Sodium dicloxacillin is a semi synthetic derivative of the isoxasocyl penicillin group that may appear in oral suspension form and in caplets. For the analysis of the raw materials and the finished products, it is recommended to use high performance liquid chromatography that is an unavailable method at the dicloxen capsule manufacturing lab for the routine analysis of the drug. To develop and to validate a useful ultraviolet spectrophotometry method for the quality control of sodium dicloxacillin in Dicloxen capsules

  7. Passive sorting of capsules by deformability

    Science.gov (United States)

    Haener, Edgar; Juel, Anne

    We study passive sorting according to deformability of liquid-filled ovalbumin-alginate capsules. We present results for two sorting geometries: a straight channel with a half-cylindrical obstruction and a pinched flow fractioning device (PFF) adapted for use with capsules. In the half-cylinder device, the capsules deform as they encounter the obstruction, and travel around the half-cylinder. The distance from the capsule's centre of mass to the surface of the half-cylinder depends on deformability, and separation between capsules of different deformability is amplified by diverging streamlines in the channel expansion downstream of the obstruction. We show experimentally that capsules can be sorted according to deformability with their downstream position depending on capillary number only, and we establish the sensitivity of the device to experimental variability. In the PFF device, particles are compressed against a wall using a strong pinching flow. We show that capsule deformation increases with the intensity of the pinching flow, but that the downstream capsule position is not set by deformation in the device. However, when using the PFF device like a T-Junction, we achieve improved sorting resolution compared to the half-cylinder device.

  8. Intrauterine fertilization capsules--a clinical trial

    DEFF Research Database (Denmark)

    Lenz, S; Lindenberg, S; Sundberg, K

    1991-01-01

    Treatment of 26 women with tubal infertility was attempted using intrauterine capsules loaded with oocytes and spermatozoa. The stimulation protocol was as used for in vitro fertilization and embryo transfer and consisted of short-term use of Buserelin, human menopausal gonadotropin, and human...... and piston from an intrauterine device. Six complete capsules and parts of two other capsules were expelled. None of the women became pregnant, compared with a pregnancy rate of 21% per aspiration following in vitro fertilization and embryo transfer during the same period....... chorionic gonadotropin. Oocytes were collected by ultrasonically guided transvaginal aspiration, and spermatozoa were prepared by swim-up technique. The gametes were placed in agar capsules 4 hr after oocyte collection, and the capsules were introduced to the uterine fundus using an insertion tube...

  9. Adhesive capsulitis of the shoulder: MR arthrography

    International Nuclear Information System (INIS)

    Kim, Hyun Jeong; Han, Tae Il; Lee, Kwang Won; Choi, Youn Seon; Kim, Dae Hong; Han, Hyun Young; Song, Mun Kab; Kwon, Soon Tae

    2001-01-01

    Adhesive capsulitis is a clinical syndrome involving pain and decreased joint motion caused by thickening and contraction of the joint capsule. The purpose of this study is to describe the MR arthrographic findings of this syndrome. Twenty-nine sets of MR arthrographic images were included in the study. Fourteen patients had adhesive capsulitis diagnosed by physical examination and arthrography, and their MR arthrographic findings were compared with those of 15 subjects in the control group. The images were retrospectively reviewed with specific attention to the thickness of the joint capsule, volume of the axillary pouch (length, width, height(depth)), thinkness of the coracohumeral ligament, presence of extra-articular contrast extravasation, and contrst filling of the subcoracoid bursa. Mean capsular thickness measured at the inferior portion of the axillary pouch was 4.1 mm in patients with adhesive capsulitis and 1.5 mm in the control group. The mean width of the axillary pouch was 2.5 mm in patients and 9.5 mm in controls. In patients, the capsule was significantly thicker and the axillary pouch significantly narrower than in controls (p<0.05). Capsule thickness greater than 2.5 mm at the inferior portion of the axillary pouch (sensitivity 93%, specificity 80%) and a pouch narrower than 3.5 mm (sensitivity 93%, specificity 100%) were useful criteria for the diagnosis of adhesive capsulitis. In patients with this condition, extra-articular contrast extravasation was noted in six patients (43%) and contrast filling of the subcoracoid bursa in three (21%). The MR arthrographic findings of adhesive capsulitis are capsular thickening, a low-volume axillary pouch, extra-articular contrast extravasation, and contrast filling of the subcoracoid bursa. Capsule thickness greater than 2.5 mm at the inferior portion of the axillary pouch and a pouch width of less than 3.5 mm are useful diagnostic imaging characteristics

  10. Thermoregulation of Capsule Production by Streptococcus pyogenes

    Science.gov (United States)

    Kang, Song Ok; Wright, Jordan O.; Tesorero, Rafael A.; Lee, Hyunwoo; Beall, Bernard; Cho, Kyu Hong

    2012-01-01

    The capsule of Streptococcus pyogenes serves as an adhesin as well as an anti-phagocytic factor by binding to CD44 on keratinocytes of the pharyngeal mucosa and the skin, the main entry sites of the pathogen. We discovered that S. pyogenes HSC5 and MGAS315 strains are further thermoregulated for capsule production at a post-transcriptional level in addition to the transcriptional regulation by the CovRS two-component regulatory system. When the transcription of the hasABC capsular biosynthetic locus was de-repressed through mutation of the covRS system, the two strains, which have been used for pathogenesis studies in the laboratory, exhibited markedly increased capsule production at sub-body temperature. Employing transposon mutagenesis, we found that CvfA, a previously identified membrane-associated endoribonuclease, is required for the thermoregulation of capsule synthesis. The mutation of the cvfA gene conferred increased capsule production regardless of temperature. However, the amount of the capsule transcript was not changed by the mutation, indicating that a post-transcriptional regulator mediates between CvfA and thermoregulated capsule production. When we tested naturally occurring invasive mucoid strains, a high percentage (11/53, 21%) of the strains exhibited thermoregulated capsule production. As expected, the mucoid phenotype of these strains at sub-body temperature was due to mutations within the chromosomal covRS genes. Capsule thermoregulation that exhibits high capsule production at lower temperatures that occur on the skin or mucosal surface potentially confers better capability of adhesion and invasion when S. pyogenes penetrates the epithelial surface. PMID:22615992

  11. Hydrodynamick instabilities on ICF capsules

    International Nuclear Information System (INIS)

    Haan, S.W.

    1991-01-01

    This article summarizes our current understanding of hydrodynamic instabilities as relevant to ICF. First we discuss classical, single mode Rayleigh-Taylor instability, and nonlinear effects in the evolution of a single mode. Then we discuss multimode systems, considering: (1) the onset of nonlinearity; (2) a second order mode coupling theory for weakly nonlinear effects, and (3) the fully nonlinear regime. Two stabilization mechanisms relevant to ICF are described next: gradient scale length and convective stabilization. Then we describe a model which is meant to estimate the weakly nonlinear evolution of multi-mode systems as relevant to ICF, given the short-wavelength stabilization. Finally, we discuss the relevant code simulation capability, and experiments. At this time we are quite optimistic about our ability to estimate instability growth on ICF capsules, but further experiments and simulations are needed to verify the modeling. 52 refs

  12. Small bowel involvement documented by capsule endoscopy in Churg-Strauss syndrome.

    Science.gov (United States)

    Beye, Birane; Lesur, Gilles; Claude, Pierre; Martzolf, Lionel; Kieffer, Pierre; Sondag, Daniel

    2015-01-01

    Churg-Strauss syndrome is a small and medium vessel vasculitis and is also known as allergic granulomatous angiitis. Gastrointestinal involvement is common in patients with Churg-Strauss syndrome (20-50%). The most common symptoms are abdominal pain, diarrhoea and occasionally gastrointestinal bleeding and perforation. We present a case of Churg-Strauss syndrome with small bowel lesions documented by video capsule endoscopy.

  13. Effect of sodium lauryl sulfate in dissolution media on dissolution of hard gelatin capsule shells.

    Science.gov (United States)

    Zhao, Fang; Malayev, Vyacheslav; Rao, Venkatramana; Hussain, Munir

    2004-01-01

    Sodium lauryl sulfate (SLS) is a commonly used surfactant in dissolution media for poorly water soluble drugs. However, it has occasionally been observed that SLS negatively impacts the dissolution of drug products formulated in gelatin capsules. This study investigated the effect of SLS on the dissolution of hard gelatin capsule shells. The USP paddle method was used with online UV monitoring at 214 nm (peptide bond). Empty size #0 capsule shells were held to the bottom of the dissolution vessel by magnetic three-prong sinkers. SLS significantly slowed down the dissolution of gelatin shells at pH < 5. Visually, the gelatin shells transformed into some less-soluble precipitate under these conditions. This precipitate was found to contain a higher sulfur content than the gelatin control sample by elemental analysis, indicating that SLS is part of the precipitate. Additionally, the slowdown of capsule shell dissolution was shown to be dependent on the SLS concentration and the ionic strength of the media. SLS interacts with gelatin to form a less-soluble precipitate at pH < 5. The use of SLS in dissolution media at acidic pH should be carefully evaluated for gelatin capsule products.

  14. Vessel Operator System

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Operator cards are required for any operator of a charter/party boat and or a commercial vessel (including carrier and processor vessels) issued a vessel permit from...

  15. Problem Oriented Neutron-Gamma Cross Sections Libraries for WWER-440 and WWER-1000 Shielding and Reactor Vessel Dosimetry Application

    International Nuclear Information System (INIS)

    Belousov, S.; Antonov, S.; Ilieva, K.

    1997-01-01

    The 47 neutron and 20 gamma group libraries BGL-440 and BGL-1000 for the shielding and reactor vessel dosimetry application have been generated for WWER-440 and WWER-1000 by collapsing the VITAMIN-B6 library (199 neutron and 42 gamma groups on the base of ENDF/B-6). The first parts of the libraries for neutron-gamma transport calculation, BGL-440-1 (150 nuclides) and BGL-1000-1 (140 nuclides), have been generated by a modified version of SAS1X control module of the SCALE system. The appropriate zone-average neutron flux had been used for these sub-libraries collapsing. The BGL-440-2 and BGL-1000-2 sub-libraries consist of cross sections for all 120 nuclides of VITAMIN-B6, for calculation of the transport through non-reactor materials of dosimeters, capsules, specimens which may be placed in the cavity behind the reactor vessel. The neutron spectrum just beyond the RPV had been used for this collapsing. As the first test the comparative calculations of the neutron flux on/behind the WWER-1000 reactor vessel have been realised using the libraries BGL-1000 and BUGLE, intended for the American PWR reactors. The integral neutron flux values by BGL-1000 and BUGLE differ by 3% onto the vessel, and 5% behind the vessel. This result shows that the calculations of the neutron flux responses for the WWER vessel surveillance, especially in locations behind the WWER vessel have to be done by the appropriate BGL library. Key words: neutron transport, multigroup neutron cross section libraries

  16. Gammatography of thick lead vessels

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  17. Application of annealing for WWER vessels life extension

    International Nuclear Information System (INIS)

    Badanin, V.I.; Gorynin, I.V.; Nickolaev, V.A.; Dragunov, Y.G.; Fedorov, V.G.

    1989-01-01

    Safe operation of NPP is greatly dependent on the guarantee of reactor vessel brittle failure strength with account for the effect of radiation embrittlement of vessel material. Recovery of irradiated material properties is principally important way to extend radiation life of reactor vessel. The aim of this report is to demonstrate the efficiency of annealing for recovery of vessel material properties and extension of its service-life

  18. [Evaluation of nopal capsules in diabetes mellitus].

    Science.gov (United States)

    Frati Munari, A C; Vera Lastra, O; Ariza Andraca, C R

    1992-01-01

    To find out if commercial capsules with dried nopal (prickle-pear cactus, Opuntia ficus indica may have a role in the management of diabetes mellitus, three experiments were performed: 30 capsules where given in fasting condition to 10 diabetic subjects and serum glucose was measured through out 3 hours; a control test was performed with 30 placebo capsules. OGTT with previous intake of 30 nopal or placebo capsules was performed in ten healthy individuals. In a crossover and single blinded study 14 diabetic patients withdrew the oral hypoglycemic treatment and received 10 nopal or placebo capsules t.i.d. during one week; serum glucose, cholesterol and tryglycerides levels were measured before and after each one-week period. Five healthy subjects were also studied in the same fashion. Opuntia capsules did not show acute hypoglycemic effect and did not influence OGTT. In diabetic patients serum glucose, cholesterol and tryglycerides levels did not change with Opuntia, but they increased with placebo (P nopal, while cholesterol and triglycerides decreased (P < 0.01 vs. placebo). The intake of 30 Opuntia capsules daily in patients with diabetes mellitus had a discrete beneficial effect on glucose and cholesterol. However this dose is unpractical and at present it is not recommended in the management of diabetes mellitus.

  19. Pressure vessel lid

    International Nuclear Information System (INIS)

    Schoening, J.; Elter, C.; Becker, G.; Pertiller, S.

    1986-01-01

    The invention concerns a lid for closing openings in reactor pressure vessels containing helium, which is made as a circular casting with hollow spaces and a flat floor and is set on the opening and kept down. It consists of helium-tight metal cast material with sufficient temperature resistance. There are at least two concentric heat resistant seals let into the bottom of the lid. The bottom is in immediate contact with the container atmosphere and has hollow spaces in its inside in the area opposite to the opening. (orig./HP) [de

  20. Oxygen fugacity and piston cylinder capsule assemblies

    Science.gov (United States)

    Jakobsson, S.

    2011-12-01

    A double capsule assembly designed to control oxygen fugacity in piston cylinder experiments has been tested at 1200 °C and 10 kbar. The assembly consists of an outer Pt-capsule containing a solid buffer (Ni-NiO or Co-CoO plus H2O) and an inner AuPd-capsule containing the sample, H2O and a Pt-wire. To prevent direct contact with the buffer phases the AuPd-capsule is embedded in finely ground Al2O3 along with some coarser, fractured Al2O3 facilitating fluid inclusion formation. No water loss is observed in the sample even after 48 hrs but a slight increase in water content is observed in longer duration runs due to oxygen and hydrogen diffusion into the AuPd-capsule. Carbon from the furnace also diffuses through the outer Pt-capsule but reacts with H2O in the outer capsule to form CO2 and never reaches the inner capsule. Oxygen fugacity of runs in equilibrium with the Ni-NiO and Co-CoO buffers was measured by analyzing the Fe content of the Pt-wire in the sample1 and by analyzing Fe dissolved in the AuPd capsule2. The second method gives values that are in good agreement with established buffer whereas results from the first method are one half to one log units higher than the established values. References 1. E. Medard, C. A. McCammon, J. A. Barr, T. L. Grove, Am. Mineral. 93, 1838 (2008). 2. J. Barr, T. Grove, Contrib. Mineral. Petrol. 160, 631 (2010)

  1. A novel superconducting magnetic levitation method to support the laser fusion capsule by using permanent magnets

    Directory of Open Access Journals (Sweden)

    Xiaojia Li

    2018-05-01

    Full Text Available A novel magnetic levitation support method is proposed, which can relieve the perturbation caused by traditional support methods and provide more accurate position control of the capsule. This method can keep the perfect symmetry of the octahedral spherical hohlraum and has the characteristics in stability, tunability and simplicity. It is also favorable that all the results, such as supporting forces acting on the superconducting capsule, are calculated analytically, and numerical simulations are performed to verify these results. A typical realistic design is proposed and discussed in detail. The superconducting coating material is suggested, and the required superconducting properties are listed. Damped oscillation of the floating capsule in thin helium gas is discussed, and the restoring time is estimated. Keywords: ICF capsule support, Magnetic levitation, Symmetry, PACS Codes: 52.57.Fg, 74.70.Ad, 74.78.-W

  2. Design Improvements of a Fuel Capsule for Re-irradiation Tests

    International Nuclear Information System (INIS)

    Kang, Young-Hwan; Choi, Myung-Hwan; Kim, Jong Kiun; Youm, Ki Un; Yoon, Ki Byeong; Kim, Bong Goo

    2006-01-01

    The development of an advanced reactor system such as the next generation nuclear plant and other generation IV systems require new fuels, claddings, and structural materials. To characterize the performance of these new materials, it is necessary for us to have leading-edge technology to satisfy the specific test requirements of the recent R and D activities such as the high-fluence- and high burnup- related tests. Thus, new capsule assembling technology and re-instrumentation technology has been developed to meet the demands for the high burnup test at HANARO since 2003. In 2003, a mockup of the capsule assembly machine was designed and fabricated. The performance test which started in 2004 was undertaken to determine and present the main performance characteristics of the capsule assembly machine (CAM) including the special tools. In 2005, a series of analyses using a finite element analysis program, ANSYS and full scale tests in air were performed to improve the design of the capsule's components for an effective utilization of the CAM. The handling tools were fully qualified through the performance tests in 2006. KAERI is now reviewing the water flow area in the top region of a fuel capsule main body for re-irradiation tests and optimizing the design of the central region area of a capsule to be joined with special bolts

  3. Programmed temperature control of capsule in irradiation test with personal computer at JMTR

    International Nuclear Information System (INIS)

    Saito, H.; Uramoto, T.; Fukushima, M.; Obata, M.; Suzuki, S.; Nakazaki, C.; Tanaka, I.

    1992-01-01

    The capsule irradiation facility is one of various equipments employed at the Japan Materials Testing Reactor (JMTR). The capsule facility has been used in irradiation tests of both nuclear fuels and materials. The capsule to be irradiated consists of the specimen, the outer tube and inner tube with a annular space between them. The temperature of the specimen is controlled by varying the degree of pressure (below the atmospheric pressure) of He gas in the annular space (vacuum-controlled). Beside this, in another system the temperature of the specimen is controlled with electric heaters mounted around the specimen (heater-controlled). The use of personal computer in the capsule facility has led to the development of a versatile temperature control system at the JMTR. Features of this newly-developed temperature control system lie in the following: the temperature control mode for a operation period can be preset prior to the operation; and the vacuum-controlled irradiation facility can be used in cooperation with the heater-controlled. The introduction of personal computer has brought in automatic heat-up and cool-down operations of the capsule, setting aside the hand-operated jobs which had been conducted by the operators. As a result of this, the various requirements seeking a higher accuracy and efficiency in the irradiation can be met by fully exploiting the capabilities incorporated into the facility which allow the cyclic or delicate changes in the temperature. This paper deals with a capsule temperature control system with personal computer. (author)

  4. SATCAP: a program for thermal-hydraulic design of saturated temperature capsule

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Niimi, Motoji; Someya, Hiroyuki; Kobayashi, Toshiki.

    1988-02-01

    For material irradiation tests at JMTR, user's technical requirements are gradually becoming more rigid, permitting only a small temperature deviation from the desired during irradiation of test materials. As specimen temperature control equipment, several conception were proposed and some of them were translated into actual machines with the capsule having electrical seath heaters in it. This system is highly reliable unless the integrity of the heaters is threatened. However, in a test with the object of achieving a high exposure of specimen to neutrons, the break of a heater or deterioration of a heater caused by irradiation lowers the reliability of the system. To cope with this drawback, as a part of the irradiation technique improvement program, ''Satulated Temperature Capsule'' has been developing. This type capsule, in which the water suplied is boiled, bases on the conception of keeping the coolant at the saturated temperature facilitates the temperature control. Though there are various types of capsules employed at JMTR, the experience of the capsule into which the coolant is injected lacks. In designing, thermal performances have to fully understood. Therefore, a program was compiled to evaluate the thermal behavior in the capsule. The present report describes the calculation procedure and guides of input and output for the program. (author)

  5. Light-water-reactor pressure-vessel surveillance dosimetry using solid-state track recorders

    International Nuclear Information System (INIS)

    Ruddy, F.H.; Roberts, J.H.; Gold, R.; Preston, C.C.

    1983-07-01

    The accumulation of neutron dose by the pressure vessel of an operating nuclear power plant results in damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, dosimetric measurements must be made to evaluate the neutron dose to the pressure vessel and relate this dose to the cumulative radiation damage. Advanced dosimetry techniques are being evaluated for surveillance of operating reactors. Solid-state track recorder (SSTR) techniques are included among these advanced dosimetry techniques. Described herein are low neutron fluence calibration and standardization measurements that are being carried out in pressure vessel mockup benchmark neutron fields in the USA, Belgium, and England. In addition, high fluence SSTR dosimetry capsules have been irradiated with metallurgical specimens in a pressure vessel mockup facility. The design and deployment of advances SSTR dosimetry capsules in operating power reactors are also described

  6. Neutron Assay System for Confinement Vessel Disposition

    International Nuclear Information System (INIS)

    Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Valdez, Jose I.; Vigil, Georgiana M.

    2012-01-01

    Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the CVs. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of special nuclear material (SNM) in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of (le)100-g 239 Pu equivalent in a vessel for safeguards termination. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements.

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

    Science.gov (United States)

    Winn, Bradley Huegh

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

  8. Life extension of the BR2 aluminium vessel

    International Nuclear Information System (INIS)

    Koonen, E.; Fabry, A.; Chaouadi, R.; Verwerft, M.; Raedt, C. de; Winckel, S. van; Wacquier, W.; Dadoumont, J.; Verwimp, A.

    2000-01-01

    The BR2 reactor has recently undergone a major refurbishment comprising the replacement of all vessel internals. The vessel itself however was not replaced. An important requalification programme has been executed to prove that the vessel would remain fit during the contemplated life extension period of BR2. Representative material samples could be obtained from the shroud surrounding the vessel. A comprehensive in-service inspection was carried out and a vessel surveillance programme has been established. (author)

  9. Anterior joint capsule of the normal hip and in children with transient synovitis: US study with anatomic and histologic correlation

    NARCIS (Netherlands)

    S.G.F. Robben (Simon); M.H. Lequin (Maarten); A.F.M. Diepstraten (Ad); J.C. den Hollander (Jan); C.A. Entius; M. Meradji

    1999-01-01

    textabstractPURPOSE: To study the anatomic components of the anterior joint capsule of the normal hip and in children with transient synovitis. MATERIALS AND METHODS: Six cadaveric specimens were imaged with ultrasonography (US) with special attention to the anterior

  10. Capsule production by Pseudomonas aeruginosa

    Energy Technology Data Exchange (ETDEWEB)

    Lynn, A.R.

    1984-01-01

    Mucoid strains of Pseudomonas aeruginosa, associated almost exclusively with chronic respiratory infections in patients with cystic fibrosis, possess a capsule composed of alginic acid similar to one produced by Azotobacter vinelandii. Recent reports have provided evidence that the biosynthetic pathway for alginate in P. aeruginosa may differ from the pathway proposed for A. vinelandii in that synthesis in P. aeruginosa may occur by way of the Entner-Doudoroff pathway. Incorporation of isotope from (6-/sup 14/C)glucose into alginate by both P. aueroginosa and A. vinelandii was 10-fold greater than that from either (1-/sup 14/C)/sup -/ or (2-/sup 14/C)glucose, indicating preferential utilization of the bottom half of the glucose molecule for alginate biosynthesis. These data strongly suggest that the Entner-Doudoroff pathway plays a major role in alginate synthesis in both P. aeruginosa and A. vinelandii. The enzymes of carbohydrate metabolism in mucoid strains of P. aeruginosa appear to be unchanged whether alignate is actively produced or not and activities do not differ significantly from nonmucoid strain PAO.

  11. Vortex rings from Sphagnum moss capsules

    Science.gov (United States)

    Whitaker, Dwight; Strassman, Sam; Cha, Jung; Chang, Emily; Guo, Xinyi; Edwards, Joan

    2010-11-01

    The capsules of Sphagnum moss use vortex rings to disperse spores to suitable habitats many kilometers away. Vortex rings are created by the sudden release of pressurized air when the capsule ruptures, and are an efficient way to carry the small spores with low terminal velocities to heights where they can be carried by turbulent wind currents. We will present our computational model of these explosions, which are carried out using a 2-D large eddy simulation (LES) on FLUENT. Our simulations can reproduce the observed motion of the spore clouds observed from moss capsules with high-speed videos, and we will discuss the roles of bursting pressure, cap mass, and capsule morphology on the formation and quality of vortex rings created by this plant.

  12. Final report for the 1st ex-vessel neutron dosimetry installations and evaluations for Kori unit 2 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-11-15

    This report describes a neutron fluence assessment performed for the Kori unit 2 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 20 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 20.

  13. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Li, Nam Jin; Hong, Joon Wha

    2007-01-15

    This report describes a neutron fluence assessment performed for the Yonggwang Unit 1 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 16 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 1 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

  14. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

    2008-01-15

    This report describes a neutron fluence assessment performed for the Kori Unit 2 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 21 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 21.

  15. Final report for the 1st ex-vessel neutron dosimetry installation and evaluations for Kori unit 4 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-11-15

    This report describes a neutron fluence assessment performed for the Kori unit 4 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 16 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 4 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 16.

  16. Final report for the 2nd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Yonggwang Unit 2 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

    2008-01-15

    This report describes a neutron fluence assessment performed for the Yonggwang Unit 2 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During Cycle 16 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

  17. Vacuum vessel for thermonuclear device

    International Nuclear Information System (INIS)

    Kikuchi, Mitsuru; Kurita, Gen-ichi; Onozuka, Masaki; Suzuki, Masaru.

    1997-01-01

    Heat of inner walls of a vacuum vessel that receive radiation heat from plasmas by way of first walls is removed by a cooling medium flowing in channels for cooling the inner walls. Nuclear heat generation of constitutional materials of the vacuum vessel caused by fast neutrons and γ rays is removed by a cooling medium flowing in cooling channels disposed in the vacuum vessel. Since the heat from plasmas and the nuclear heat generation are removed separately, the amount of the cooling medium flowing in the channels for cooling inner walls is increased for cooling a great amount of heat from plasmas while the amount of the cooling medium flowing in the channels for cooling the inside of the vacuum vessel is reduced for cooling the small amount of nuclear heat generation. Since the amount of the cooling medium can thus be optimized, the capacity of the facilities for circulating the cooling medium can be reduced. In addition, since the channels for cooling the inner walls and the channels of cooling medium formed in the vacuum vessel are disposed to the inner walls of the vacuum vessel on the side opposite to plasmas, integrity of the channels relative to leakage of the cooling medium can be ensured. (N.H.)

  18. Vacuum vessel for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Mitsuru; Kurita, Gen-ichi [Japan Atomic Energy Research Inst., Tokyo (Japan); Onozuka, Masaki; Suzuki, Masaru

    1997-07-31

    Heat of inner walls of a vacuum vessel that receive radiation heat from plasmas by way of first walls is removed by a cooling medium flowing in channels for cooling the inner walls. Nuclear heat generation of constitutional materials of the vacuum vessel caused by fast neutrons and {gamma} rays is removed by a cooling medium flowing in cooling channels disposed in the vacuum vessel. Since the heat from plasmas and the nuclear heat generation are removed separately, the amount of the cooling medium flowing in the channels for cooling inner walls is increased for cooling a great amount of heat from plasmas while the amount of the cooling medium flowing in the channels for cooling the inside of the vacuum vessel is reduced for cooling the small amount of nuclear heat generation. Since the amount of the cooling medium can thus be optimized, the capacity of the facilities for circulating the cooling medium can be reduced. In addition, since the channels for cooling the inner walls and the channels of cooling medium formed in the vacuum vessel are disposed to the inner walls of the vacuum vessel on the side opposite to plasmas, integrity of the channels relative to leakage of the cooling medium can be ensured. (N.H.)

  19. 21 CFR 520.1660b - Oxytetracycline hydrochloride capsules.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Oxytetracycline hydrochloride capsules. 520.1660b... Oxytetracycline hydrochloride capsules. (a) Specifications. The drug is in capsule form with each capsule containing 125 or 250 milligrams of oxytetracycline hydrochloride. Oxytetracycline is the antibiotic...

  20. 60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  1. 60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-15

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

  2. 2nd RCM of the CRP on Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the Monju Reactor Vessel. Working Material

    International Nuclear Information System (INIS)

    2010-01-01

    The overall objective of the CRP is to improve the Member States’ analytical capabilities in the field of fast reactor in-vessel sodium thermal hydraulics. A necessary condition towards achieving this objective is a wide international validation effort of the data and codes currently employed for the simulation of the various physical effects involved in this field. Therefore, in providing the required wide international basis of interested Member States, each applying different methodologies, the CRP will contribute towards achieving the stated objective with the help of benchmark exercises focusing, in a first stage, on the numerical simulation of temperature stratification of sodium observed in the Monju reactor vessel at a turbine trip test conducted in December 1995 during the original start-up experiments, and with the help of a thorough assessment of the calculation versus measured data comparisons

  3. Adhesive capsulitis: review of imaging and treatment

    International Nuclear Information System (INIS)

    Harris, Guy; Bou-Haider, Pascal; Harris, Craig

    2013-01-01

    Adhesive capsulitis is one of the most common conditions affecting the shoulder; however, early clinical diagnosis can be challenging. Treatment is most effective when commenced prior to the onset of capsular thickening and contracture; consequently, the role of imaging is increasing. The aim of this review is to demonstrate the typical imaging appearances of adhesive capsulitis and to examine some of the evidence regarding each of these imaging modalities. An evaluation of the various management options available to the clinician is also presented.

  4. Radioactive gas-containing polymeric capsule

    International Nuclear Information System (INIS)

    Winchell, H.S.; Lewis, R.E.

    1975-01-01

    A disposable ventilation study system for dispensing a single patient dosage of gaseous radioisotopes to patients for pulmonary function studies is disclosed. A gas impermeable capsule encloses the gaseous radioisotope and is stored within a radioactivity shielding body of valve means which shears the capsule to dispense the radioisotope to the patient. A breathing bag receives the patient's exhalation of the radioisotope and permits rebreathing of the radioisotope by the patient. 18 claims, 7 drawing figures

  5. Benchmarking uranyl peroxide capsule chemistry in organic media

    Energy Technology Data Exchange (ETDEWEB)

    Neal, Harrison A.; Nyman, May [Department of Chemistry, Oregon State University, Corvallis, OR (United States); Szymanowski, Jennifer; Fein, Jeremy B.; Burns, Peter C. [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, Notre Dame, IN (United States)

    2017-01-03

    Uranyl peroxide capsules are a recent addition to polyoxometalate (POM) chemistry. Ten years of development has ensued only in water, while transition metal POMs are commonly exploited in aqueous and organic media, controlled by counterions or ligation to render the clusters hydrophilic or hydrophobic. Here, new uranyl POM behavior is recognized in organic media, including (1) stabilization and immobilization of encapsulated hydrophilic countercations, identified by Li nuclear magnetic resonance (NMR) spectroscopy, (2) formation of new cluster species upon phase transfer, (3) extraction of uranyl clusters from different starting materials including simulated spent nuclear fuel, (4) selective phase transfer of one cluster type from a mixture, and (5) phase transfer of clusters from both acidic and alkaline media. The capsule morphology of the uranyl POMs renders accurate characterization by X-ray scattering, including the distinction of geometrically similar clusters. Compositional analysis of the aqueous phase post-extraction provided a quantitative determination of the ion exchange process that enables transfer of the clusters into the organic phase. Preferential partitioning of uranyl POMs into organic media presents new frontiers in metal ion behavior and chemical reactions in the confined space of the cluster capsules in hydrophobic media, as well as the reactivity of clusters at the organic/aqueous interface. (copyright 2017 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  6. Benchmarking uranyl peroxide capsule chemistry in organic media

    International Nuclear Information System (INIS)

    Neal, Harrison A.; Nyman, May; Szymanowski, Jennifer; Fein, Jeremy B.; Burns, Peter C.

    2017-01-01

    Uranyl peroxide capsules are a recent addition to polyoxometalate (POM) chemistry. Ten years of development has ensued only in water, while transition metal POMs are commonly exploited in aqueous and organic media, controlled by counterions or ligation to render the clusters hydrophilic or hydrophobic. Here, new uranyl POM behavior is recognized in organic media, including (1) stabilization and immobilization of encapsulated hydrophilic countercations, identified by Li nuclear magnetic resonance (NMR) spectroscopy, (2) formation of new cluster species upon phase transfer, (3) extraction of uranyl clusters from different starting materials including simulated spent nuclear fuel, (4) selective phase transfer of one cluster type from a mixture, and (5) phase transfer of clusters from both acidic and alkaline media. The capsule morphology of the uranyl POMs renders accurate characterization by X-ray scattering, including the distinction of geometrically similar clusters. Compositional analysis of the aqueous phase post-extraction provided a quantitative determination of the ion exchange process that enables transfer of the clusters into the organic phase. Preferential partitioning of uranyl POMs into organic media presents new frontiers in metal ion behavior and chemical reactions in the confined space of the cluster capsules in hydrophobic media, as well as the reactivity of clusters at the organic/aqueous interface. (copyright 2017 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  7. Vacuum vessel for thermonuclear device

    International Nuclear Information System (INIS)

    Kikuchi, Mitsuru; Nagashima, Keisuke; Suzuki, Masaru; Onozuka, Masaki.

    1997-01-01

    A vacuum vessel main body and structural members at the inside and the outside of the vacuum vessel main body are constituted by structural materials activated by irradiation of neutrons from plasmas such as stainless steels. Shielding members comprising tungsten or molybdenum are disposed on the surface of the vacuum vessel main body and the structural members of the inside and the outside of the main body. The shielding members have a function also as first walls or a seat member for the first walls. Armor tiles may be disposed to the shielding members. The shielding members and the armor tiles are secured to a securing seat member disposed, for example, to an inner plate of the vacuum vessel main body by bolts. Since the shielding members are disposed, it is not necessary to constitute the vacuum vessel main body and the structural members at the inside and the outside thereof by using a low activation material which is less activated, such as a titanium alloy. (I.N.)

  8. Vacuum vessel for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Mitsuru; Nagashima, Keisuke [Japan Atomic Energy Research Inst., Tokyo (Japan); Suzuki, Masaru; Onozuka, Masaki

    1997-07-11

    A vacuum vessel main body and structural members at the inside and the outside of the vacuum vessel main body are constituted by structural materials activated by irradiation of neutrons from plasmas such as stainless steels. Shielding members comprising tungsten or molybdenum are disposed on the surface of the vacuum vessel main body and the structural members of the inside and the outside of the main body. The shielding members have a function also as first walls or a seat member for the first walls. Armor tiles may be disposed to the shielding members. The shielding members and the armor tiles are secured to a securing seat member disposed, for example, to an inner plate of the vacuum vessel main body by bolts. Since the shielding members are disposed, it is not necessary to constitute the vacuum vessel main body and the structural members at the inside and the outside thereof by using a low activation material which is less activated, such as a titanium alloy. (I.N.)

  9. Investigation of natural frequencies of laser inertial confinement fusion capsules using resonant ultrasound spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Xiaojun [Institute of Modern Physics, Fudan University, Shanghai 200433 (China); Research Center of Laser Fusion, CAEP, Mianyang 621900 (China); Tang, Xing; Wang, Zongwei [Research Center of Laser Fusion, CAEP, Mianyang 621900 (China); Chen, Qian; Qian, Menglu [Institute of Acoustic, Tongji University, Shanghai 200433 (China); Meng, Jie [Research Center of Laser Fusion, CAEP, Mianyang 621900 (China); Tang, Yongjian [Institute of Modern Physics, Fudan University, Shanghai 200433 (China); Research Center of Laser Fusion, CAEP, Mianyang 621900 (China); Zou, Yaming; Shen, Hao [Institute of Modern Physics, Fudan University, Shanghai 200433 (China); Gao, Dangzhong, E-mail: dgaocn@163.com [Research Center of Laser Fusion, CAEP, Mianyang 621900 (China)

    2017-01-15

    Highlights: • The frequency equation of isotropic multi-layer hollow spheres was derived using three-dimension (3D) elasticity theory and transfer matrix method. • The natural frequencies of the capsules with a millimeter-sized diameter are determined experimentally using resonant ultrasound spectrum (RUS) system. • The predicted natural frequencies of the frequency equation accord well with the observed results. • The theoretical and experimental investigation has proved the potential applicability of RUS to both metallic and non-metallic capsules. - Abstract: The natural frequency problem of laser inertial confinement fusion (ICF) capsules is one of the basic problems for determining non-destructively the elasticity modulus of each layer material using resonant ultrasound spectroscopy (RUS). In this paper, the frequency equation of isotropic one-layer hollow spheres was derived using three dimension (3D) elasticity theory and some simplified frequency equations were discussed under axisymmetric and spherical symmetry conditions. The corresponding equation of isotropic multi-layer hollow spheres was given employing transfer matrix method. To confirm the validity of the frequency equation and explore the feasibility of RUS for characterizing the ICF capsules, three representative capsules with a millimeter-sized diameter were determined by piezoelectric-based resonant ultrasound spectroscopy (PZT-RUS) and laser-based resonant ultrasound spectroscopy (LRUS) techniques. On the basis of both theoretical and experimental results, it is proved that the calculated and measured natural frequencies are accurate enough for determining the ICF capsules.

  10. Remote-welding technique for assembling in-pile IASCC capsule in hot cell

    International Nuclear Information System (INIS)

    Kawamata, Kazuo; Ishii, Toshimitsu; Kanazawa, Yoshiharu; Iwamatsu, Shigemi; Ohmi, Masao; Shimizu, Michio; Matsui, Yoshinori; Saito, Jun-ichi; Ugachi, Hirokazu; Kaji, Yoshiyuki; Tsukada, Takashi

    2006-01-01

    In order to investigate behavior of the irradiation assisted stress corrosion cracking (IASCC) caused by the simultaneous effects of neutron irradiation and high temperature water environment in such a light water reactor (LWR), it is necessary to perform crack growth tests in an in-pile IASCC capsule irradiated in the Japan Materials Testing Reactor (JMTR). The development of the remote-welding technique is essential for remotely assembling the in-pile IASCC capsule installing the pre-irradiated CT specimens. This report describes a new remote-welding machine developed for assembling the in-pile IASCC capsule. The remote-welding technique that the capsule tube is rotated light under the fixed torch was applied to the machine for the welding of thick and large-diameter tubes. The assembly work of four in-pile IASCC capsules having pre-irradiated CT specimens in the hot cell was succeeded for performing the crack growth test under the neutron irradiation in JMTR. The irradiation test of two capsules has been already finished in JMTR without problems. (author)

  11. Development of a design methodology for hydraulic pipelines carrying rectangular capsules

    International Nuclear Information System (INIS)

    Asim, Taimoor; Mishra, Rakesh; Abushaala, Sufyan; Jain, Anuj

    2016-01-01

    The scarcity of fossil fuels is affecting the efficiency of established modes of cargo transport within the transportation industry. Efforts have been made to develop innovative modes of transport that can be adopted for economic and environmental friendly operating systems. Solid material, for instance, can be packed in rectangular containers (commonly known as capsules), which can then be transported in different concentrations very effectively using the fluid energy in pipelines. For economical and efficient design of such systems, both the local flow characteristics and the global performance parameters need to be carefully investigated. Published literature is severely limited in establishing the effects of local flow features on system characteristics of Hydraulic Capsule Pipelines (HCPs). The present study focuses on using a well validated Computational Fluid Dynamics (CFD) tool to numerically simulate the solid-liquid mixture flow in both on-shore and off-shore HCPs applications including bends. Discrete Phase Modelling (DPM) has been employed to calculate the velocity of the rectangular capsules. Numerical predictions have been used to develop novel semi-empirical prediction models for pressure drop in HCPs, which have then been embedded into a robust and user-friendly pipeline optimisation methodology based on Least-Cost Principle. - Highlights: • Local flow characteristics in a pipeline transporting rectangular capsules. • Development of prediction models for the pressure drop contribution of capsules. • Methodology developed for sizing of Hydraulic Capsule Pipelines. • Implementation of the developed methodology to obtain optimal pipeline diameter.

  12. Analysis and radiation dose rate measurement of the Al-1050 capsule on the rabbit system facility

    International Nuclear Information System (INIS)

    Sarwani; Sutrisno; H, Saleh; Rohidi; M, Kawkab

    2000-01-01

    Aluminium is a kind of light metal with density of 2.7 gram /cm exp 3,regarding to the aluminium is characteristic such as easy to fabricated,has a good corrosion resistant and radiation heat resistant, therefore aluminum is selected to be used as a material for sample irradiation capsule with high neutron fluency. Analysis using neutron activation method and capsule irradiation by using high neutron fluency and dose radiation rate measurement was done. The analysis result show that impurities in the Al-1050 capsule are Fe, Cu, Mg, Sb, Zn, and Mn. The capsule irradiated at 15 MW during 6 Hours with neutron fluency of 2,8 x 10 exp 17 n/cm exp 2. The radiation doses rate after 24 hours decay is 220 mrad/h at 0-meter distance and 60 mrad/h at 1-meter distance. Respectively. From the analysis results and measurement show that the Al-1050 capsule has no high neutron absorption element and available to get continuing irradiation at 15 MW as far as 6 hours. Due to the personal safety, therefore the capsule handling could be carried out in the hot cell

  13. SATCAP-B: a program for thermal-hydraulic design of 'Saturated Temperature Capsule'

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Someya, Hiroyuki; Niimi, Motoji

    1989-11-01

    As an advanced irradiation technique, the JMTR (Japan Materials Testing Reactor) project is developing a 'Saturated Temperature Capsule' which water is injected in and boiled. When the water is kept at a constant pressure, the water temperature does not become higher than the saturated temperature. This type capsule is based on the conception of keeping the coolant to the saturated temperature and using the temperature control. In designing the capsule in which the inner coolant is injected, thermal performances have to be understood as exactly as possible. Then, a program (named SATCAP) was compiled to graps the thermal performance within the capsule. On the other hand, a 'Saturated Temperature Capsule' was made and irradiated in the JMTR core. It was indicated from supplied water temperatures recorded by thermo-couples attached in the capsule that heat transfer coefficients prefered models due to natural convection to models incorporated in the initial version of the program. Then, the program was revised by adding mainly heat transfer model based on natural convection. The present report describes the calculation procedure and guides of input and output for the revised program (SATCAP version-B). (author)

  14. Design verification test of instrumented capsule (02F-11K) for nuclear fuel irradiation in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Sohn, J. M.; Oh, J. M. [and others

    2004-01-01

    An instrumented capsule is being developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in HANARO. The instrumented capsule for measuring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. The instrumented capsule includes three test fuel rods installed thermocouple to measure fuel centerline temperature and three SPNDs (Self-Powered Neutron Detector) to monitor the neutron flux. Its stability was verified by out-of-pile performance test, and its safety evaluation was also shown that the safety requirements were satisfied. And then, to verify the design of the instrumented capsule in the test hole, it was successfully irradiated in the test hole of HANARO from March 14, 2003 to June 1, 2003 (53.8 full power days at 24 MWth). During irradiation, the centerline temperature of PWR UO{sub 2} fuel pellets fabricated by KEPCO Nuclear Fuel Company and the neutron flux were continuously measured and monitored. The test fuel rods were irradiated at less than 350 W/cm to 5.13 GWD/MTU with fuel centerline peak temperature below 1,375 .deg. C. The structural stability of the capsule was satisfied by the naked eye in service pool of HANARO. The capsule and test fuel rods were dismantled and test fuel rods were examined at the hot cell of IMEF (Irradiated Material Examination Facility)

  15. Design Improvements of the Capsule Components and the Handling Tools for an Effective Utilization of the Capsule Assembly Machine

    International Nuclear Information System (INIS)

    Kang, Young Hwan; Kim, Jong Kium; Youm, Ki Un; Yoon, Ki Byung; Choi, Myung Hwan; Kim, Bong Gu

    2006-01-01

    Various in-pile test programs for the development of new fuels and materials are planned for the HANARO reactor. To meet the demands for the required tests in the HANARO reactor, new capsule assembling technology is required in the HANARO reactor. For this reason, a series of analyses and experiments was performed in 2005. For the assembly workability tests of the capsule components, three different kinds of protection tubes and two different shapes of the locking bolt heads were proposed and tested. It was confirmed that the newly designed protection tube and bolts worked quite well without any problems. Since the new structure is quite similar to that of the currently used capsule, it was assumed that an additional vibration tests and seismic analysis would not be needed. Through the stress analysis of the three proposed structures by using ANSYS code, it showed that the maximum displacement and stress intensity for the tube reducer were 1.57mm and 21MPa, respectively. To improve the workability and handling capability of the bolting and clamping tools of stainless steel 304, Al6061/T6 was selected as one of the candidates and thus new tools were manufactured and tested. The assembly test results showed that the new tools were found to be useful for executing key tasks such as a bolting and a clamping and they were much faster than the old tools made of stainless steel, thereby increasing the workability rate and lowering the manufacturing costs

  16. Capsule endoscopy in neoplastic diseases

    Science.gov (United States)

    Pennazio, Marco; Rondonotti, Emanuele; de Franchis, Roberto

    2008-01-01

    Until recently, diagnosis and management of small-bowel tumors were delayed by the difficulty of access to the small bowel and the poor diagnostic capabilities of the available diagnostic techniques. An array of new methods has recently been developed, increasing the possibility of detecting these tumors at an earlier stage. Capsule endoscopy (CE) appears to be an ideal tool to recognize the presence of neoplastic lesions along this organ, since it is non-invasive and enables the entire small bowel to be visualized. High-quality images of the small-bowel mucosa may be captured and small and flat lesions recognized, without exposure to radiation. Recent studies on a large population of patients undergoing CE have reported small-bowel tumor frequency only slightly above that reported in previous surgical series (range, 1.6%-2.4%) and have also confirmed that the main clinical indication to CE in patients with small-bowel tumors is obscure gastrointestinal (GI) bleeding. The majority of tumors identified by CE are malignant; many were unsuspected and not found by other methods. However, it remains difficult to identify pathology and tumor type based on the lesion’s endoscopic appearance. Despite its limitations, CE provides crucial information leading in most cases to changes in subsequent patient management. Whether the use of CE in combination with other new diagnostic (MRI or multidetector CT enterography) and therapeutic (Push-and-pull enteroscopy) techniques will lead to earlier diagnosis and treatment of these neoplasms, ultimately resulting in a survival advantage and in cost savings, remains to be determined through carefully-designed studies. PMID:18785274

  17. How to replace a reactor pressure vessel

    International Nuclear Information System (INIS)

    Huber, R.

    1996-01-01

    A potential life extending procedure for a nuclear reactor after, say, 40 years of service life, might in some circumstances be the replacement of the reactor pressure vessel. Neutron induced degradation of the vessel might make replacement by one of a different material composition desirable, for example. Although the replacement of heavy components, such as steam generators, has been possible for many years, the pressure vessel presents a much more demanding task if only because it is highly irradiated. Some preliminary feasibility studies by Siemens are reported for the two removal strategies that might be considered. These are removal of the entire pressure vessel in one piece and dismantling it into sections. (UK)

  18. Reactor vessel decommissioning project. Final report

    International Nuclear Information System (INIS)

    Schoonen, D.H.

    1984-09-01

    This report describes a reactor vessel decommissioning project; it documents and explains the project objectives, scope, performance results, and sodium removal process. The project was successfully completed in FY-1983, within budget and without significant problems or adverse impact on the environment. Waste generated by the operation included the reactor vessel, drained sodium, and liquid, solid, and gaseous wastes which were significantly less than project estimates. Personnel radiation exposures were minimized, such that the project total was one-half the predicted exposure level. Except for the sodium removed, the material remaining in the reactor vessel is essentially the same as when the vessel arrived for processing

  19. Increase of cyclic durability of pressure vessels

    International Nuclear Information System (INIS)

    Vorona, V.A.; Zvezdin, Yu.I.

    1980-01-01

    The durability of multilayer pressure vessels under cyclic loading is compared with single-layer vessels. The relative conditional durability is calculated taking into account the assumption on the consequent destruction of layers and viewing a vessel wall as an indefinite plate. It is established that the durability is mainly determined by the number of layers and to a lesser degree depends on the relative size of the defect for the given layer thickness. The advantage of the multilayer vessels is the possibility of selecting layer materials so that to exclude the effect of agressive corrosion media on the strength [ru

  20. The TPX vacuum vessel and in-vessel components

    International Nuclear Information System (INIS)

    Heitzenroeder, P.; Bialek, J.; Ellis, R.; Kessel, C.; Liew, S.

    1994-01-01

    The Tokamak Physics Experiment (TPX) is a superconducting tokamak with double-null diverters. TPX is designed for 1,000-second discharges with the capability of being upgraded to steady state operation. High neutron yields resulting from the long duration discharges require that special consideration be given to materials and maintainability. A unique feature of the TPX is the use of a low activation, titanium alloy vacuum vessel. Double-wall vessel construction is used since it offers an efficient solution for shielding, bakeout and cooling. Contained within the vacuum vessel are the passive coil system, Plasma Facing Components (PFCs), magnetic diagnostics, and the internal control coils. All PFCs utilize carbon-carbon composites for exposed surfaces

  1. Safety vessels for explosive fusion reactor

    International Nuclear Information System (INIS)

    Mineev, V.

    1994-01-01

    The failure of several types of geometrically similar cylindrical and spherical steel and glass fibers vessels filled with water or air was investigated when an explosive charge of TNT was detonated in the center. Vessels had radius 50-1000 mm, thickness of walls 2-20%. The detonation on TNT imitated energy release. The parameter: K = M/mf is a measure of the strength of the vessel where M is the mass of the vessel, and mf is the mass of TNT for which the vessel fails. This demanded 2-4 destroyed and nondestroyed shots. It may be showed that: K=A/σ f where σ f is the fracture stress of the material vessel, and A = const = F(energy TNT, characteristic of elasticity of vessel material). The chief results are the following: (1) A similar increase in the geometrical dimensions of steel vessels by a factor of 10 leads to the increase of parameter K in about 5 times and to decrease of failure deformation in 7 times (scale effect). (2) For glass fibers, scale effect is absent. (3) This problem is solved in terms of theory energetic scale effect. (4) The concept of TNT equivalent explosive makes it possible to use these investigations to evaluate the response of safety vessels for explosive fusion reactor

  2. Pressurized wet digestion in open vessels (T11)

    International Nuclear Information System (INIS)

    Kettisch, P.; Maichin, P.; Zischka, M.; Knapp, G.

    2002-01-01

    Full text: Pressurized wet digestion in closed vessels, microwave assisted or with conventional conductive heating, is the most important sample preparation technique for digestion or leaching procedures in element analysis. In comparison to open vessel digestion closed vessel digestion methods have many advantages, but there is one disadvantage - complex and expensive vessel designs. A new technique - pressurized wet digestion in open vessels - combine the advantages of closed vessel sample digestion with the application of simple and cheap open vessels made of quartz or PFA. The vessels are placed in a high pressure Asher HPA, which is adapted with a Teflon liner and filled partly with water. The analytical results with 30 ml quartz vessels, 22 ml PFA vessels and 1.5 ml PIA auto sampler cups will be shown. In principle every dimensions of vessels can be used. The vessels are loaded with sample material (max. 1.5 g with quartz vessels, max. 0.5 g with PFA vessels and 50 mg with auto sampler cups) and digestion reagent. Afterwards the vessels are simply covered with PTFE stoppers and not sealed. The vessels are transferred into a special adapted HPA and digested at temperatures up to 270 o C. The digestion time is 90 min. and cooling down to room temperature 30 min. The analytical results of CRM's are within the certified values and no cross contamination and losses of volatile elements could be observed. (author)

  3. Development of Multiple Capsule Robots in Pipe

    Directory of Open Access Journals (Sweden)

    Shuxiang Guo

    2018-05-01

    Full Text Available Swallowable capsule robots which travel in body cavities to implement drug delivery, minimally invasive surgery, and diagnosis have provided great potential for medical applications. However, the space constraints of the internal environment and the size limitations of the robots are great challenges to practical application. To address the fundamental challenges of narrow body cavities, a different-frequency driven approach for multiple capsule robots with screw structure manipulated by external electromagnetic field is proposed in this paper. The multiple capsule robots are composed of driven permanent magnets, joint permanent magnets, and a screw body. The screw body generates a propulsive force in a fluidic environment. Moreover, robots can form new constructions via mutual docking and release. To provide manipulation guidelines for active locomotion, a dynamic model of axial propulsion and circumferential torque is established. The multiple start and step-out frequencies for multiple robots are defined theoretically. Moreover, the different-frequency driven approach based on geometrical parameters of screw structure and the overlap angles of magnetic polarities is proposed to drive multiple robots in an identical electromagnetic field. Finally, two capsule robots were prototyped and experiments in a narrow pipe were conducted to verify the different motions such as docking, release, and cooperative locomotion. The experimental results demonstrated the validity of the driven approach for multiple capsule robots in narrow body cavities.

  4. Cementing a wellbore using cementing material encapsulated in a shell

    Energy Technology Data Exchange (ETDEWEB)

    Aines, Roger D.; Bourcier, William L.; Duoss, Eric B.; Spadaccini, Christopher M.; Cowan, Kenneth Michael

    2016-08-16

    A system for cementing a wellbore penetrating an earth formation into which a pipe extends. A cement material is positioned in the space between the wellbore and the pipe by circulated capsules containing the cement material through the pipe into the space between the wellbore and the pipe. The capsules contain the cementing material encapsulated in a shell. The capsules are added to a fluid and the fluid with capsules is circulated through the pipe into the space between the wellbore and the pipe. The shell is breached once the capsules contain the cementing material are in position in the space between the wellbore and the pipe.

  5. Cementing a wellbore using cementing material encapsulated in a shell

    Energy Technology Data Exchange (ETDEWEB)

    Aines, Roger D.; Bourcier, William L.; Duoss, Eric B.; Floyd, III, William C.; Spadaccini, Christopher M.; Vericella, John J.; Cowan, Kenneth Michael

    2017-03-14

    A system for cementing a wellbore penetrating an earth formation into which a pipe extends. A cement material is positioned in the space between the wellbore and the pipe by circulated capsules containing the cement material through the pipe into the space between the wellbore and the pipe. The capsules contain the cementing material encapsulated in a shell. The capsules are added to a fluid and the fluid with capsules is circulated through the pipe into the space between the wellbore and the pipe. The shell is breached once the capsules contain the cementing material are in position in the space between the wellbore and the pipe.

  6. Probabilistic retinal vessel segmentation

    Science.gov (United States)

    Wu, Chang-Hua; Agam, Gady

    2007-03-01

    Optic fundus assessment is widely used for diagnosing vascular and non-vascular pathology. Inspection of the retinal vasculature may reveal hypertension, diabetes, arteriosclerosis, cardiovascular disease and stroke. Due to various imaging conditions retinal images may be degraded. Consequently, the enhancement of such images and vessels in them is an important task with direct clinical applications. We propose a novel technique for vessel enhancement in retinal images that is capable of enhancing vessel junctions in addition to linear vessel segments. This is an extension of vessel filters we have previously developed for vessel enhancement in thoracic CT scans. The proposed approach is based on probabilistic models which can discern vessels and junctions. Evaluation shows the proposed filter is better than several known techniques and is comparable to the state of the art when evaluated on a standard dataset. A ridge-based vessel tracking process is applied on the enhanced image to demonstrate the effectiveness of the enhancement filter.

  7. Modeling and preliminary thermal analysis of the capsule for a creep test in HANARO

    International Nuclear Information System (INIS)

    Choi, Myoung Hwan; Cho, Man Soon; Choo, Kee Nam; Kang, Young Hwan; Sohn, Jae Min; Shin, Yoon Taeg; Park, Sung Jae; Kim, Bong Goo; Kim, Young Jin

    2005-01-01

    A creep capsule is a device to investigate the creep characteristics of nuclear materials during inpile irradiation tests. To obtain the design data of the capsule through a preliminary thermal analysis, a 2-dimensional model for the cross section of the capsule including the specimens and components is generated, and an analysis using the ANSYS program is performed. The gamma-heating rates of the materials for the HANARO power of 30MW are considered, and the effect of the gap size and the control rod position on the temperature of the specimen is discussed. From the analysis it is found that the gap between the thermal media and the external tube has a significant effect on the temperature of the specimen. The temperature by increasing the position of the control rod is decreased

  8. Light-Responsive Polymer Micro- and Nano-Capsules

    Directory of Open Access Journals (Sweden)

    Valentina Marturano

    2016-12-01

    Full Text Available A significant amount of academic and industrial research efforts are devoted to the encapsulation of active substances within micro- or nanocarriers. The ultimate goal of core–shell systems is the protection of the encapsulated substance from the environment, and its controlled and targeted release. This can be accomplished by employing “stimuli-responsive” materials as constituents of the capsule shell. Among a wide range of factors that induce the release of the core material, we focus herein on the light stimulus. In polymers, this feature can be achieved introducing a photo-sensitive segment, whose activation leads to either rupture or modification of the diffusive properties of the capsule shell, allowing the delivery of the encapsulated material. Micro- and nano-encapsulation techniques are constantly spreading towards wider application fields, and many different active molecules have been encapsulated, such as additives for food-packaging, pesticides, dyes, pharmaceutics, fragrances and flavors or cosmetics. Herein, a review on the latest and most challenging polymer-based micro- and nano-sized hollow carriers exhibiting a light-responsive release behavior is presented. A special focus is put on systems activated by wavelengths less harmful for living organisms (mainly in the ultraviolet, visible and infrared range, as well as on different preparation techniques, namely liposomes, self-assembly, layer-by-layer, and interfacial polymerization.

  9. Improvement to reactor vessel

    International Nuclear Information System (INIS)

    1974-01-01

    The vessel described includes a prestressed concrete vessel containing a chamber and a removable cover closing this chamber. The cover is in concrete and is kept in its closed position by main and auxiliary retainers, comprising fittings integral with the concrete of the vessel. The auxiliary retainers pass through the concrete of the cover. This improvement may be applied to BWR, PWR and LMFBR type reactor vessel [fr

  10. The capsule biosynthesis locus of Haemophilus influenzae show conspicuous similarity to the corresponding locus in Haemophilus sputorum and may have been recruited from this species by horizontal gene transfer

    DEFF Research Database (Denmark)

    Nielsen, Signe Maria; de Gier, Camilla; Dimopoulou, Chrysoula

    2015-01-01

    in export and processing of the capsular material, show high similarity to the corresponding genes in capsulate lineages of the pathogenic species Haemophilus influenzae; indeed, standard bexA and bexB PCRs for detection of capsulated strains of H. influenzae give positive results with strains of H....... sputorum was only distantly related to H. influenzae. In contrast to H. influenzae, the capsule locus in H. sputorum is not associated with transposases or other transposable elements. Our data suggest that the capsule locus of capsulate lineages of H. influenzae may relatively recently have been recruited...

  11. Capsule Formulation of Ethanolic Extract of Pasak Bumi (Eurycoma Longifolia Jack., and its Effect on Human Health Vital Signs

    Directory of Open Access Journals (Sweden)

    Laela Hayu Nurani

    2017-08-01

    cover the bitter taste of E. longifolia. Clinical trials in this study use design pre-post treatment in healthy humans. Subjects used were male - healthy men and healthy women who met inclusion criteria and were subjected with formulated capsule for 14 days. The study resulted capsule formula comprising of the ethanolic extract of E. longifolia 300 mg, vivapur 101 300 mg, 58 mg maydis starch, aerosil 3%, talc 2%, and Mg stearate 1%. The results showed that the capsule of E. longifolia did not affect the value of heart rate, respiration rate, body temperature and weight (P > 0.05, based on paired t-test, but they causes a decrease in blood pressure of healthy human. The ethanol extract of E. longifolia caused vasodilation of blood vessels that can be used in antihypertensive therapy.

  12. ALICE HMPID Radiator Vessel

    CERN Document Server

    2003-01-01

    View of the radiator vessels of the ALICE/HMPID mounted on the support frame. Each HMPID module is equipped with 3 indipendent radiator vessels made out of neoceram and fused silica (quartz) windows glued together. The spacers inside the vessel are needed to stand the hydrostatic pressure. http://alice-hmpid.web.cern.ch/alice-hmpid

  13. The first capsule implosion experiments on Orion

    International Nuclear Information System (INIS)

    Garbett, W J; Horsfield, C J; Gales, S G; Leatherland, A E; Rubery, M S; Coltman, J E; Meadowcroft, A E; Rice, S J; Simons, A J; Woolhead, V E

    2016-01-01

    Direct drive capsule implosions are being developed on the Orion laser at AWE as a platform for ICF and HED physics experiments. The Orion facility combines both long pulse and short-pulse beams, making it well suited for studying the physics of alternative ignition approaches. Orion implosions also provide the opportunity to study aspects of polar direct drive. Limitations on drive symmetry from the relatively small number of laser beams makes predictive modelling of the implosions challenging, resulting in some uncertainty in the expected capsule performance. Initial experiments have been fielded to evaluate baseline capsule performance and inform future design optimization. Highly promising DD fusion neutron yields in excess of 10 9 have been recorded. Results from the experiments are presented alongside radiation-hydrocode modelling. (paper)

  14. Capsule endoscopy—A mechatronics perspective

    Science.gov (United States)

    Lin, Lin; Rasouli, Mahdi; Kencana, Andy Prima; Tan, Su Lim; Wong, Kai Juan; Ho, Khek Yu; Phee, Soo Jay

    2011-03-01

    The recent advances in integrated circuit technology, wireless communication, and sensor technology have opened the door for development of miniature medical devices that can be used for enhanced monitoring and treatment of medical conditions. Wireless capsule endoscopy is one of such medical devices that has gained significant attention during the past few years. It is envisaged that future wireless capsule endoscopies replace traditional endoscopy procedures by providing advanced functionalities such as active locomotion, body fluid/tissue sampling, and drug delivery. Development of energy-efficient miniaturized actuation mechanisms is a key step toward achieving this goal. Here, we review some of the actuators that could be integrated into future wireless capsules and discuss the existing challenges.

  15. Chord length distribution for a compound capsule

    International Nuclear Information System (INIS)

    Pitřík, Pavel

    2017-01-01

    Chord length distribution is a factor important in the calculation of ionisation chamber responses. This article describes Monte Carlo calculations of the chord length distribution for a non-convex compound capsule. A Monte Carlo code was set up for generation of random chords and calculation of their lengths based on the input number of generations and cavity dimensions. The code was written in JavaScript and can be executed in the majority of HTML viewers. The plot of occurrence of cords of different lengths has 3 peaks. It was found that the compound capsule cavity cannot be simply replaced with a spherical cavity of a triangular design. Furthermore, the compound capsule cavity is directionally dependent, which must be taken into account in calculations involving non-isotropic fields of primary particles in the beam, unless equilibrium of the secondary charged particles is attained. (orig.)

  16. Analysis of materials in connection with corium melt retention in WWER reactor vessels. Final report for the period 15 October 1995 - 14 October 1996

    International Nuclear Information System (INIS)

    Efanov, A.D.

    1997-02-01

    Analysis of the state of severe accident codes being developed in Russia describing processes of corium - reactor pressure vessel interaction during severe accidents showed that at present there is no reliable validated and verified code. This study considered some of the most advanced severe accident codes which include models of heat generating liquid convections and RPV tolerance capacity however the possibility of physico-chemical interactions is not being considered. The final report demonstrates one of the examples for the settlement modelling of processes of reactor core debris cooling and corium pool mirror cooling with the use of the KOSTER 2 Code developed at IPPE. It was shown that cooling by water spray within some definite region of drop dimensions (0.5 / 4 mm in diameter) could provide the opportunity of the WWER type RPV integrity preservation under the accepted conditions. Due to some uncertainties in calculation modules obtained results are recommended to be treated as preliminary. (author). 26 refs, figs, tabs

  17. Half-unit weighted bilinear algorithm for image contrast enhancement in capsule endoscopy

    Science.gov (United States)

    Rukundo, Olivier

    2018-04-01

    This paper proposes a novel enhancement method based exclusively on the bilinear interpolation algorithm for capsule endoscopy images. The proposed method does not convert the original RBG image components to HSV or any other color space or model; instead, it processes directly RGB components. In each component, a group of four adjacent pixels and half-unit weight in the bilinear weighting function are used to calculate the average pixel value, identical for each pixel in that particular group. After calculations, groups of identical pixels are overlapped successively in horizontal and vertical directions to achieve a preliminary-enhanced image. The final-enhanced image is achieved by halving the sum of the original and preliminary-enhanced image pixels. Quantitative and qualitative experiments were conducted focusing on pairwise comparisons between original and enhanced images. Final-enhanced images have generally the best diagnostic quality and gave more details about the visibility of vessels and structures in capsule endoscopy images.

  18. TMI-2 Vessel Investigation Project Metallurgical Program

    International Nuclear Information System (INIS)

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

    1990-01-01

    The TMI-2 [Three Mile Island unit 2] Vessel Investigation Project Metallurgical Program at Argonne National Laboratory is a part of the international TMI-2 Vessel Investigation Project being conducted jointly by the U.S. Nuclear Regulatory Commission and the Organization for Economic Co-operation and Development (OECD). The overall project consists of three phases, namely (1) recovery of material samples from the lower head of the TMI-2 reactor, (2) examination and analysis of the lower head samples and the preparation and testing of archive material subjected to a similar thermal history, and (3) procurement, examination, and analysis of companion core material located adjacent to or near the lower head material. The specific objectives of the ANL Metallurgical Program, which accounts for a major portion of Phase 2, are to prepare metallographic and mechanical test specimen blanks from the TMI-2 lower head material, prepare similar test specimen blanks from suitable archive material subjected to the appropriate thermal processing, determine the mechanical properties of the lower vessel head and archive materials under the conditions of the core-melt accident, and assess the lower head integrity and margin-to-failure during the accident. The ANL work consists of three tasks: (1) archive materials program, (2) fabrication of metallurgical and mechanical test specimens from the TMI-2 pressure vessel samples, and (3) mechanical property characterization of TMI-2 lower pressure vessel head and archive material

  19. The Cryogenic Impact Resistant Evaluation of Composite Materials for Use in Composite Pressure Vessels with an Additional Cryogenic Bonding Scope, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — The intent of the proposed effort is to investigate the detailed composite material performance characteristics after being subjected to cryogenic temperatures and...

  20. 46 CFR 115.812 - Pressure vessels and boilers.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels and boilers. 115.812 Section 115.812... CERTIFICATION Material Inspections § 115.812 Pressure vessels and boilers. (a) Pressure vessels must be tested... testing requirements for boilers are contained in § 61.05 in subchapter F of this chapter. [CGD 85-080, 61...

  1. Safety analysis report: packages. GPHS shipping package supplement 2 to the PISA shipping package (packaging of fissile and other radioactive materials). Final report

    International Nuclear Information System (INIS)

    Chalfant, G. G.

    1981-06-01

    Safety Analysis Report DPST-78-124-1 is amended to permit shipment of 6 General Purpose Heat Source (GPHS) capsules (max.). Each capsule contains an average of 2330 curies of 238 Pu, and each pair of capsules is contained in a welded stainless steel primary containment vessel, all of which are doubly contained in a flanged secondary containment vessel. This is in addition to the forms discussed in DPST-78-124-1 and Supplement 1

  2. A Study on the Manufacturing Properties of Crack Self-Healing Capsules Using Cement Powder for Addition to Cement Composites

    Directory of Open Access Journals (Sweden)

    Yun-Wang Choi

    2017-01-01

    Full Text Available We fabricated crack self-healing capsules using cement powder for mixing into cement composites and evaluated the properties of the capsule manufacturing process in this study. The manufacture of the self-healing capsules is divided into core production processing of granulating cement in powder form and a coating process for creating a wall on the surfaces of the granulated cement particles. The produced capsules contain unhardened cement and can be mixed directly with the cement composite materials because they are protected from moisture by the wall material. Therefore, the untreated cement is present in the form of a capsule within the cement composite, and hydration can be induced by moisture penetrating the crack surface in the event of cracking. In the process of granulating the cement, it is important to obtain a suitable consistency through the kneading agent and to maintain the moisture barrier performance of the wall material. We can utilize the results of this study as a basis for advanced self-healing capsule technology for cement composites.

  3. Design of Endoscopic Capsule With Multiple Cameras.

    Science.gov (United States)

    Gu, Yingke; Xie, Xiang; Li, Guolin; Sun, Tianjia; Wang, Dan; Yin, Zheng; Zhang, Pengfei; Wang, Zhihua

    2015-08-01

    In order to reduce the miss rate of the wireless capsule endoscopy, in this paper, we propose a new system of the endoscopic capsule with multiple cameras. A master-slave architecture, including an efficient bus architecture and a four level clock management architecture, is applied for the Multiple Cameras Endoscopic Capsule (MCEC). For covering more area of the gastrointestinal tract wall with low power, multiple cameras with a smart image capture strategy, including movement sensitive control and camera selection, are used in the MCEC. To reduce the data transfer bandwidth and power consumption to prolong the MCEC's working life, a low complexity image compressor with PSNR 40.7 dB and compression rate 86% is implemented. A chipset is designed and implemented for the MCEC and a six cameras endoscopic capsule prototype is implemented by using the chipset. With the smart image capture strategy, the coverage rate of the MCEC prototype can achieve 98% and its power consumption is only about 7.1 mW.

  4. Physics of ignition for ICF capsules

    International Nuclear Information System (INIS)

    Lindl, J.D.

    1989-01-01

    The implosion of an ICF capsule must accomplish both compression of the main fuel to several hundred grams per cubic centimeter and heating and compression of the central region of the fuel to ignition. This report discusses the physics of these conditions

  5. Extended School Year. Information Capsule. Volume 0910

    Science.gov (United States)

    Blazer, Christie

    2010-01-01

    Extended school years are being considered by districts around the country as educators search for new ways to raise student achievement. The addition of time to the school calendar is also supported by President Barack Obama, who recently stated that American students do not spend enough time in school. This Information Capsule addresses research…

  6. Safety analysis report for packaging: the ORNL in-pile capsule shipping cask

    International Nuclear Information System (INIS)

    Evans, J.H.; Chipley, K.K.; Haynie, C.B.; Crowley, W.K.; Just, R.A.

    1977-11-01

    The ORNL in-pile capsule shipping cask is used to transport irradiated experimental capsules and spent fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures were used to determine the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations

  7. Isolation of Capsulate Bacteria from Acute Dentoalveolar Abscesses

    OpenAIRE

    Lewis, M. A. O.; Milligan, S. G.; MacFarlane, T. W.; Carmichael, F. A.

    2011-01-01

    The presence of a capsule was determined for 198 bacterial strains (57 facultative anaerobes, 141 strict anaerobes) isobdted from pus samples aspirated from 40 acute dentoalveolar abscesses. A total of 133 (67 per cent) of the isolates (42 facultative anaerobes, 91 strict anaerobes) were found to have a capsule. Possession ofa capsule may in part explain the apparent pathogenicity of the bacterial species encountered in acute dentoalveolar abscess.Keywords - Bacterial capsule; Acute dentoalve...

  8. Prediction of the Effects of Radiation FOr Reactor pressure vessel and in-core Materials using multi-scale modeling - 60 years foreseen plant lifetime (PERFORM-60 project)

    International Nuclear Information System (INIS)

    Al Mazouzi, A.; Bugat, S.; Leclercq, S.; Massoud, J.-P.; Moinereau, D.; Lidbury, D.; Van Dyck, S.; Marini, B.; Alamo, Ana

    2010-01-01

    The PERFECT project of the EURATOM framework program (FP6) is a first step through the development of a simulation platform that contains several advanced numerical tools aiming at the prediction of irradiation damage in both the reactor pressure vessel (RPV) and its internals using state-of-the-art knowledge. These tools allow simulation of irradiation effects on the microstructure and the constitutive behavior of the RPV low alloy steels, as well as their fracture mechanics properties. For the reactor internals, the first partial models were established, describing radiation damage to the microstructure and providing a first description of the stress corrosion behaviour of austenitic steels in primary environment, without physical linking of the radiation and corrosion effects. Thus, relying on the existing PERFECT Roadmap, the FP7 Collaborative Project PERFORM 60 has mainly for objective to develop similar tools that would allow the simulation of the combined effects of irradiation and corrosion on internals, in addition to a further improvement of the existing ones on RPV made of bainitic steels. From the managerial view point, PERFORM 60 is based on two technical sub-projects, namely (i) RPV and (ii) Internals. In addition, a Users' Group and a training scheme have been adopted in order to allow representatives of constructors, utilities, research organizations... from Europe, USA and Japan to participate actively in the process of appraising the limits and potentialities of the developed tools as well as their validation against qualified experimental data

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

    International Nuclear Information System (INIS)

    Onozuka, Masanori; Nakahira, Masataka

    2006-01-01

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

  10. Renal capsule for augmentation cystoplasty in canine model: a favorable biomaterial?

    Directory of Open Access Journals (Sweden)

    Mehdi Salehipour

    2016-04-01

    Full Text Available ABSTRACT Purpose: To evaluate effectiveness of canine renal capsule for augmentation cystoplasty. Materials and Methods: Ten adult dogs participated in this study. After induction of anesthesia each animal underwent bed side urodynamic study, bladder capacity and bladder pressure was recorded. Then via mid line incision abdominal cavity was entered, right kidney was identified and its capsule was dissected. Bladder augmentation was done by anastomosing the renal capsule to the bladder. After 6 months bed side urodynamic study was performed again and changes in bladder volume and pressure were recorded. Then the animals were sacrificed and the augmented bladders were sent for histopathology evaluation. Results: Mean maximum anatomic bladder capacity before cystoplasty was 334.00±11.40cc which increased to 488.00±14.83cc post-operatively (p=0.039. Mean anatomic bladder pressure before cystoplasty was 19.00±1.58cmH2O which decreased to 12.60±1.14cmH2O post-operatively (p=0.039. Histopathology evaluation revealed epithelialization of the renal capsule with urothelium without evidence of fibrosis, collagen deposits or contracture. Conclusions: Our data shows that renal capsule is a favorable biomaterial for bladder augmentation in a canine model.

  11. Controllable fabrication and characterization of biocompatible core-shell particles and hollow capsules as drug carrier

    Science.gov (United States)

    Hao, Lingyun; Gong, Xinglong; Xuan, Shouhu; Zhang, Hong; Gong, Xiuqing; Jiang, Wanquan; Chen, Zuyao

    2006-10-01

    SiO 2@CdSe core-shell particles were fabricated by controllable deposition CdSe nanoparticles on silica colloidal spheres. Step-wise coating process was tracked by the TEM and XRD measurements. In addition, SiO 2@CdSe/polypyrrole(PPy) multi-composite particles were synthesized based on the as-prepared SiO 2@CdSe particles by cationic polymerization. The direct electrochemistry of myoglobin (Mb) could be performed by immobilizing Mb on the surface of SiO 2@CdSe particles. Immobilized with Mb, SiO 2@CdSe/PPy-Mb also displayed good bioelectrochemical activity. It confirmed the good biocompatible property of the materials with protein. CdSe hollow capsules were further obtained as the removal of the cores of SiO 2@CdSe spheres. Hollow and porous character of CdSe sub-meter size capsules made them becoming hopeful candidates as drug carriers. Doxorubicin, a typical an antineoplastic drug, was introduced into the capsules. A good sustained drug release behavior of the loading capsules was discovered via performing a release test in the PBS buffer (pH 7.4) solution at 310 k. Furthermore, SiO 2@CdSe/PPy could be converted to various smart hollow capsules via selectively removal of their relevant components.

  12. Design and testing of tubular polymeric capsules for self-healing of concrete

    Science.gov (United States)

    Araújo, M.; Van Tittelboom, K.; Feiteira, J.; Gruyaert, E.; Chatrabhuti, S.; Raquez, J.-M.; Šavija, B.; Alderete, N.; Schlangen, E.; De Belie, N.

    2017-10-01

    Polymeric healing agents have proven their efficiency to heal cracks in concrete in an autonomous way. However, the bottleneck for valorisation of self-healing concrete with polymeric healing agents is their encapsulation. In the present work, the suitability of polymeric materials such as poly(methyl methacrylate) (PMMA), polystyrene (PS) and poly(lactic acid) (PLA) as carriers for healing agents in self-healing concrete has been evaluated. The durability of the polymeric capsules in different environments (demineralized water, salt water and simulated concrete pore solution) and their compatibility with various healing agents have been assessed. Next, a numerical model was used to simulate capsule rupture when intersected by a crack in concrete and validated experimentally. Finally, two real-scale self-healing concrete beams were made, containing the selected polymeric capsules (with the best properties regarding resistance to concrete mixing and breakage upon crack formation) or glass capsules and a reference beam without capsules. The self-healing efficiency was determined after crack creation by 3-point-bending tests.

  13. Irradiation capsule design capable of continuously monitoring the creepdown of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Thoms, K.R.; Dodd, C.V.; van der Kaa, T.; Hobson, D.O.

    1978-01-01

    An irradiation capsule which permits continuous monitoring of the creepdown of Zircaloy tubing has been designed and given preliminary tests. This design effort is the major element of a cooperative research program between the United States Nuclear Regulatory Commission and the Netherlands Energy Research Foundation (ECN) and is a part of the NRC-sponsored Zircaloy creepdown program. The purpose of the Zircaloy creepdown program is to provide data on the deformation characteristics of Zircaloy tubes, typical of LWR fuel element cladding, under combined axial and tangential compressive stresses. These data will be used to verify and improve the material behavior codes that are used for the description of fuel pin behavior. The first capsule of this series contains a mockup test specimen which was machined with three different diameters, nominally 10.92-mm, 10.54-mm and 11.30-mm (.430-in., .415-in., and .445-in.). This test specimen can be moved axially thereby varying the lift-off and serving as a calibration device for the eddy-current deformation monitoring probes. Fabrication of this capsule has been completed and during out-or-reactor checkout we were able to obtain a resolution of better than 0.01-mm (0.0004-in.). The capsule is scheduled for installation in the HFR on February 8, 1978, for a 26 day irradiation test. The first pressurized capsule, and therefore the first one to monitor in-reactor cladding deformation, will be installed in the HFR on May 3, 1978

  14. Application of small specimens to fracture mechanics characterization of irradiated pressure vessel steels

    International Nuclear Information System (INIS)

    Sokolov, M.A.; Wallin, K.; McCabe, D.E.

    1996-01-01

    In this study, precracked Charpy V-notch (PCVN) specimens were used to characterize the fracture toughness of unirradiated and irradiated reactor pressure vessel steels in the transition region by means of three-point static bending. Fracture toughness at cleavage instability was calculated in terms of elastic-plastic K Jc values. A statistical size correction based upon weakest-link theory was performed. The concept of a master curve was applied to analyze fracture toughness properties. Initially, size-corrected PCVN data from A 533 grade B steel, designated HSST Plate O2, were used to position the master curve and a 5% tolerance bound for K Jc data. By converting PCVN data to IT compact specimen equivalent K Jc data, the same master curve and 5% tolerance bound curve were plotted against the Electric Power Research Institute valid linear-elastic K Jc database and the ASME lower bound K Ic curve. Comparison shows that the master curve positioned by testing several PCVN specimens describes very well the massive fracture toughness database of large specimens. These results give strong support to the validity of K Jc with respect to K Ic in general and to the applicability of PCVN specimens to measure fracture toughness of reactor vessel steels in particular. Finally, irradiated PCVN specimens of other materials were tested, and the results are compared to compact specimen data. The current results show that PCVNs demonstrate very good capacity for fracture toughness characterization of reactor pressure vessel steels. It provides an opportunity for direct measurement of fracture toughness of irradiated materials by means of precracking and testing Charpy specimens from surveillance capsules. However, size limits based on constraint theory restrict the operational test temperature range for K Jc data from PCVN specimens. 13 refs., 8 figs., 1 tab

  15. Ingestible capsule for remote controlled release of a substance

    DEFF Research Database (Denmark)

    2014-01-01

    The application relates to an ingestible capsule (102) for delivery of a substance e.g. a pharmaceutical drug, to a human or animal. The ingestible capsule comprises a capsule wall structure (202) forming a substantially sealed reservoir or lumen holding the substance (204). An electrical resonance...

  16. Metallurgy of steels for PWR pressure vessels

    International Nuclear Information System (INIS)

    Kepka, M.; Mocek, J.; Barackova, L.

    1980-01-01

    A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure. (B.S.)

  17. Metallurgy of steels for PWR pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kepka, M; Mocek, J; Barackova, L [Skoda, Plzen (Czechoslovakia)

    1980-09-01

    A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure.

  18. Method of burying vessel containing radioactive waste

    International Nuclear Information System (INIS)

    Koga, Yoshihito.

    1989-01-01

    A float having an inert gas sealed therein is attached to a tightly closed vessel containing radioactive wastes. The vessel is inserted and kept in a small hole for burying the tightly closed vessel in an excavated shaft in rocks such as of granite or rock salts, while filling bentonite as shielding material therearound. In this case, the float is so adjusted that the apparent specific gravity is made equal or nearer between the tightly closed vessel and the bentonite, so that the rightly closed vessel does not sink and cause direct contact with the rocks even if bentonite flows due to earthquakes, etc. This can prevent radioactivity contamination through water in the rocks. (S.K.)

  19. Pressure vessel design manual

    CERN Document Server

    Moss, Dennis R

    2013-01-01

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

  20. Development of a supplemental surveillance program for reactor pressure vessel thermal annealing

    International Nuclear Information System (INIS)

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

    1997-01-01

    The technical decision to thermally anneal a nuclear reactor pressure vessel (RPV) depends upon the level of embrittlement in the RPV steels, the amount of recovery of fracture toughness properties expected from the anneal, and the rate of re-embrittlement after the vessel is placed back into service. The recovery of Charpy impact toughness properties after annealing can be estimated initially by using a recovery model developed using experimental measurements of recovery (such as that developed by Eason et al. for U.S. vessel materials). However, actual validation measurements on plant-specific archived vessel materials (hopefully in the existing surveillance program) are needed; otherwise, irradiated surrogate materials, essentially the same as the RPV steels or bounding in expected behavior, must be utilized. The efficient use of any of these materials requires a supplemental surveillance program focused at both recovery and reirradiation embrittlement. Reconstituted Charpy specimens and new surveillance capsules will most likely be needed as part of this supplemental surveillance program. A new version of ASTM E 509 has recently been approved which provides guidance on thermal annealing in general and specifically for the development of an annealing supplemental surveillance program. The post-anneal re-embrittlement properties are crucial for continued plant operation, and the use of a re-embrittlement model, such as the lateral shift approach, may be overly conservative. This paper illustrates the new ASTM E 509 Standard Guide methodology for an annealing supplemental surveillance program. As an example, the proposed program for the Palisades RPV beltline steels is presented which covers the time from annealing to the end of operating license and beyond, if license renewal is pursued. The Palisades nuclear power plant RPV was planned to be annealed in 1998, but that plant is currently being re-evaluated. The proposed anneal was planned to be conducted at a

  1. Tumor Blood Vessel Dynamics

    Science.gov (United States)

    Munn, Lance

    2009-11-01

    ``Normalization'' of tumor blood vessels has shown promise to improve the efficacy of chemotherapeutics. In theory, anti-angiogenic drugs targeting endothelial VEGF signaling can improve vessel network structure and function, enhancing the transport of subsequent cytotoxic drugs to cancer cells. In practice, the effects are unpredictable, with varying levels of success. The predominant effects of anti-VEGF therapies are decreased vessel leakiness (hydraulic conductivity), decreased vessel diameters and pruning of the immature vessel network. It is thought that each of these can influence perfusion of the vessel network, inducing flow in regions that were previously sluggish or stagnant. Unfortunately, when anti-VEGF therapies affect vessel structure and function, the changes are dynamic and overlapping in time, and it has been difficult to identify a consistent and predictable normalization ``window'' during which perfusion and subsequent drug delivery is optimal. This is largely due to the non-linearity in the system, and the inability to distinguish the effects of decreased vessel leakiness from those due to network structural changes in clinical trials or animal studies. We have developed a mathematical model to calculate blood flow in complex tumor networks imaged by two-photon microscopy. The model incorporates the necessary and sufficient components for addressing the problem of normalization of tumor vasculature: i) lattice-Boltzmann calculations of the full flow field within the vasculature and within the tissue, ii) diffusion and convection of soluble species such as oxygen or drugs within vessels and the tissue domain, iii) distinct and spatially-resolved vessel hydraulic conductivities and permeabilities for each species, iv) erythrocyte particles advecting in the flow and delivering oxygen with real oxygen release kinetics, v) shear stress-mediated vascular remodeling. This model, guided by multi-parameter intravital imaging of tumor vessel structure

  2. Device for encapsulating radioactive materials

    International Nuclear Information System (INIS)

    Suthanthiran, K.

    1994-01-01

    A capsule for encapsulating radioactive material for radiation treatment comprising two or more interfitting sleeves, wherein each sleeve comprises a closed bottom portion having a circumferential wall extending therefrom, and an open end located opposite the bottom portion. The sleeves are constructed to fit over one another to thereby establish an effectively sealed capsule container. 3 figs

  3. Maury Journals - German Vessels

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — German vessels observations, after the 1853 Brussels Conference that set International Maritime Standards, modeled after Maury Marine Standard Observations.

  4. Sodium Iodide-131 (Na131I) AS Gelatin Capsules At TNRC-In Libya

    International Nuclear Information System (INIS)

    Sherief, M. F.; Abudeeb, F. N.; Abudaia, J. A.; Elghanoudy, Y. A.

    2004-01-01

    In this contribution, the production of a capsulated Na 131 I radiopharmaceutical, for treatment of variety of hyperthyroidism diseases, at Tajoura Nuclear Research Center in Tripoli-Libya is described. The process requires the application of a very small volume of iodine-131 (not more than 25μ l in some cases) with radioactivities reaching some 37 GBq per capsule. The application of such volume is necessary to prevent damage to gelatin material. Loading a volume of 100 μ l of radioactive Na 131 I solution containing 37 GBq. radioactivity within a capsule filled with anhydrous sodium hydrogen phosphate as an adsorption material for Na 131 I solution brings such solution into a direct interaction with the gelatin material. This is assumed to have an inadequate effect in therapy. To overcome this problem, the work team has introduced some substantial alterations on the irradiation procedure and the process of the pre-irradiation treatment of the target. As a consequence, that has successfully culminated in production of Na 131 I capsules with proper perspective (e.g. radioactive yield of 74 GBq from 37 GBq previously and radioactive concentration of 37 GBq/ml). (Authors)

  5. Fabrication data package for HEDL dosimetry in the ORNL Poolside Facility: LWR Pressure Vessel Mock-up irradiation

    International Nuclear Information System (INIS)

    Lippincott, E.P.; McElroy, W.N.; Kellogg, L.S.; Gold, R.; Guthrie, G.L.; Ruddy, F.H.; Ulseth, J.A.

    1981-09-01

    This document provides a complete description of the HEDL dosimetry inserted in the metallurgical specimen irradiation in the LWR Pressure Vessel Mock-up at the Oak Ridge Reactor Poolside Facility (PSF). This experiment is being conducted under the Nuclear Regulatory Commission sponsored program on Surveillance Dosimetry Improvement. The irradiation started April 1980 with recovery of the 2 x 10 19 (nominal fluence with E > 1 MeV) capsule in September 1980, the 4 x 10 19 surveillance capsule in November 1981 and the pressure vessel and void box capaules about August 1982

  6. Process-integrated online monitoring of safety-relevant aluminum airbag pressure vessel components for a combined defect detection and material property determination by using contactless NDT (EMUS and EC)

    International Nuclear Information System (INIS)

    Becker, R.; Dobmann, G.; Salzburger, H.-J.

    1999-01-01

    Airbag pressure vessels for the north-American market mainly are made by forging and by the use of steel alloys. In Europe aluminum alloys are common and the manufacturing process is extrusion of circular blanks - made from cold rolled plates - in a form applying a 100 t press at room temperature. Then by heat treatment the strength/hardness of the material is properly adjusted and after that the pressure vessel parts have to be continuously inspected with an inspection and handling cycle time of 3 s. Inspection of the axis-symmetric parts is asked for surface breaking extrusion defects as well as for surface parallel delaminations in the bulk volume. Furthermore, the material strength is a quality characteristic that has to be nondestructively registered and documented. The inspection is performed by eddy current probes and an EMAT, of which the eddy current impedance measurements are used for surface breaking extrusion defect detection and sizing (single frequency technique with digital locus curve filtering) and strength characterization (3-frequency technique with digital filtering for signal-to-noise enhancement). The bulk delaminations are detected by an EMAT-resonance technique using a spiral eddy current coil and permanent magnets for the EMA-energy transformation. The inspections are performed by singling the parts on a conveying belt, rotating two of them parallel on turntables scanning with the transducers in specially selected circular scan paths. The performance of the system is characterized by a number of 6000 parts per shift in the two time-parallel inspection lines with 3 shifts in 24 hours. The registered quality characteristics are documented by laser writing onto the surface of each part. The emphasis of the contribution is on the presentation and discussion of the safety and economical benefits by process-integrated NDT. (author)

  7. Design of the ITER vacuum vessel

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  8. Design, fabrication and irradiation test report on HANARO instrumented capsule (03M-06U) for researches of universities in 2003

    International Nuclear Information System (INIS)

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

    2005-03-01

    As a part of 2003 project for active utilization of HANARO, an instrumented capsule (03M-06U) was designed, fabricated and irradiated for the irradiation test of various nuclear materials under irradiation conditions requested by external researchers from universities. The basic structure of 03M-06U capsule was based on the 00M-01U, 01M-05U and 02M-05U capsules successfully irradiated in HANARO as 2000, 2001 and 2002 projects. However, because of the limited number of specimens and budget of 4 universities, the remained space of the capsule was charged with KAERI specimens for the development of the precise temperature control technology under irradiation. The material of the specimens is mainly Fe-based alloys partially mixed with Zr, Al and Cu-Ag alloys. The capsule is composed of 5 stages having many kinds of specimens and independent electric heater in each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. Various types of specimens such as tensile, Charpy, TEM, toughness, electrical resistance specimens were inserted in the capsule. The capsule was firstly irradiated in the CT test hole of HANARO of 30MW thermal output at 275∼500±10 .deg. C up to a fast neutron fluence of 5.4 x 10 20 (n/cm 2 ) (E>1.0MeV). The obtained results will be very valuable for the related researches of the users

  9. BWRVIP-123, Revision 1NP: BWR Vessel and Internals Project Removal and Analysis of Material Samples from Core Shroud and Top Guide at Susquehanna Unit 2

    International Nuclear Information System (INIS)

    Howell, D.; Haertel, T.; Lindberg, J.; Oliver, B.; Greenwood, L.

    2005-01-01

    Fast and thermal fluence were determined by a laboratory analysis of the samples. Fluence in the upper regions of the shroud (between the H1 and H2 welds) was substantially lower than that in the belt line region (near the H4 weld). Fluence in the top guide was significantly higher than fluence on the core shroud. As expected, helium concentrations were highest in regions where fluence was highest. Estimates of the initial boron concentration were similar to measurements made on materials removed from other reactors. A technical justification evaluated the acceptability of the sampling process with respect to structural consequences of material removal and to increased cracking susceptibility due to the as-left condition. It was determined that the sampling process was acceptable on both counts

  10. Revisiting the reactor pressure vessel for long-time operation; Revisitando la vasija a presion del reactor para largos tiempos de operacion

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  11. Automatic blood detection in capsule endoscopy video

    Czech Academy of Sciences Publication Activity Database

    Novozámský, Adam; Flusser, Jan; Tachecí, I.; Sulík, L.; Bureš, J.; Krejcar, O.

    2016-01-01

    Roč. 21, č. 12 (2016), s. 1-8, č. článku 126007. ISSN 1083-3668 R&D Projects: GA ČR GA15-16928S Institutional support: RVO:67985556 Keywords : Automatic blood detection * capsule endoscopy video Subject RIV: JD - Computer Applications, Robotics Impact factor: 2.530, year: 2016 http://library.utia.cas.cz/separaty/2016/ZOI/flusser-0466936.pdf

  12. Small bowel endoluminal imaging (capsule and enteroscopy).

    Science.gov (United States)

    Murino, Alberto; Despott, Edward J

    2017-04-01

    Over the last 16 years, the disruptive technologies of small bowel capsule endoscopy and device-assisted enteroscopy have revolutionised endoluminal imaging and minimally invasive therapy of the small bowel. Further technological developments continue to expand their indications and use. This brief review highlights the state-of-the-art in this arena and aims to summarise the current and potential future role of these technologies in clinical practice.

  13. Confinement Vessel Assay System: Calibration and Certification Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-17

    Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1 to 2 inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of SNM in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of {le} 100-g {sup 239}Pu equivalent in a vessel for safeguards termination. The system was calibrated in three different mass regions (low, medium, and high) to cover the entire plutonium mass range that will be assayed. The low mass calibration and medium mass calibration were verified for material positioned in the center of an empty vessel. The systematic uncertainty due to position bias was estimated using an MCNPX model to simulate the response of the system to material localized at various points along the inner surface of the vessel. The background component due to cosmic ray spallation was determined by performing measurements of an empty vessel and comparing to measurements in the same location with no vessel present. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements of CVs before and after cleanout.

  14. Confinement Vessel Assay System: Calibration and Certification Report

    International Nuclear Information System (INIS)

    Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Gomez, Cipriano; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Vigil, Georgiana M.

    2012-01-01

    Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1 to 2 inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of SNM in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of (le) 100-g 239 Pu equivalent in a vessel for safeguards termination. The system was calibrated in three different mass regions (low, medium, and high) to cover the entire plutonium mass range that will be assayed. The low mass calibration and medium mass calibration were verified for material positioned in the center of an empty vessel. The systematic uncertainty due to position bias was estimated using an MCNPX model to simulate the response of the system to material localized at various points along the inner surface of the vessel. The background component due to cosmic ray spallation was determined by performing measurements of an empty vessel and comparing to measurements in the same location with no vessel present. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements of CVs before and after cleanout.

  15. Problems in Pressure Vessel Design and Manufacture

    Energy Technology Data Exchange (ETDEWEB)

    Hellstroem, O [Uddeholms AB, Degerfors (Sweden); Nilson, Ragnar [AB Atomenergi, Nykoeping (Sweden)

    1963-05-15

    The general desire by the power reactor process makers to increase power rating and their efforts to involve more advanced thermal behaviour and fuel handling facilities within the reactor vessels are accompanied by an increase in both pressure vessel dimensions and various difficulties in giving practical solutions of design materials and fabrication problems. In any section of this report it is emphasized that difficulties and problems already met with will meet again in the future vessels but then in modified forms and in many cases more pertinent than before. As for the increase in geometrical size it can be postulated that with use of better materials and adjusted fabrication methods the size problems can be taken proper care of. It seems likely that vessels of sufficient large diameter and height for the largest power output, which is judged as interesting in the next ten year period, can be built without developing totally new site fabrication technique. It is, however, supposed that such a fabrication technique will be feasible though at higher specific costs for the same quality requirements as obtained in shop fabrication. By the postulated use of more efficient vessel material with principally the same good features of easy fabrication in different stages such as preparation, welding, heat treatment etc as ordinary or slightly modified carbon steels the increase in wall thickness might be kept low. There exists, however, a development work to be done for low-alloy steels to prove their justified use in large reactor pressure vessels.

  16. Problems in Pressure Vessel Design and Manufacture

    International Nuclear Information System (INIS)

    Hellstroem, O.; Nilson, Ragnar

    1963-05-01

    The general desire by the power reactor process makers to increase power rating and their efforts to involve more advanced thermal behaviour and fuel handling facilities within the reactor vessels are accompanied by an increase in both pressure vessel dimensions and various difficulties in giving practical solutions of design materials and fabrication problems. In any section of this report it is emphasized that difficulties and problems already met with will meet again in the future vessels but then in modified forms and in many cases more pertinent than before. As for the increase in geometrical size it can be postulated that with use of better materials and adjusted fabrication methods the size problems can be taken proper care of. It seems likely that vessels of sufficient large diameter and height for the largest power output, which is judged as interesting in the next ten year period, can be built without developing totally new site fabrication technique. It is, however, supposed that such a fabrication technique will be feasible though at higher specific costs for the same quality requirements as obtained in shop fabrication. By the postulated use of more efficient vessel material with principally the same good features of easy fabrication in different stages such as preparation, welding, heat treatment etc as ordinary or slightly modified carbon steels the increase in wall thickness might be kept low. There exists, however, a development work to be done for low-alloy steels to prove their justified use in large reactor pressure vessels

  17. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    Dooley, R.A.

    1986-01-01

    This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

  18. Containment vessel drain system

    Science.gov (United States)

    Harris, Scott G.

    2018-01-30

    A system for draining a containment vessel may include a drain inlet located in a lower portion of the containment vessel. The containment vessel may be at least partially filled with a liquid, and the drain inlet may be located below a surface of the liquid. The system may further comprise an inlet located in an upper portion of the containment vessel. The inlet may be configured to insert pressurized gas into the containment vessel to form a pressurized region above the surface of the liquid, and the pressurized region may operate to apply a surface pressure that forces the liquid into the drain inlet. Additionally, a fluid separation device may be operatively connected to the drain inlet. The fluid separation device may be configured to separate the liquid from the pressurized gas that enters the drain inlet after the surface of the liquid falls below the drain inlet.

  19. Posterior capsule opacification and neovascularization treated with intravitreal bevacizumab and Nd:YAG capsulotomy

    Directory of Open Access Journals (Sweden)

    Grimelda Yuriana Sánchez-Castro

    2008-10-01

    Full Text Available Grimelda Yuriana Sánchez-Castro1, Alejandra Hitos-Fájer1, Erick Mendoza-Schuster1, Raul Velez-Montoya2, Cecilio Francisco Velasco-Barona11Asociación para Evitar la Ceguera en México. Hospital “Dr. Luis Sánchez Bulnes”, México, D.F. Ophthalmology Department – Anterior Segment; 2Asociación para Evitar la Ceguera en México. Hospital “Dr. Luis Sánchez Bulnes”, México, D.F. Ophthalmology Department – Retina departmentAbstract: We reported a 75-year-old diabetic man, who developed opacification and neovascularization of the posterior capsule after extracapsular cataract extraction and posterior chamber intraocular lens implantation. The patient was treated with two injections of 2.5 mg of intravitreal bevacizumab. The treatment produced an important regression of the posterior capsular new vessels, allowing us to perform a successful Nd:YAG capsulotomy, clearing the visual axis and improving the visualization of the posterior pole. Even though, best corrected visual acuity was 20/200 due to diabetic macular edema.Keywords: posterior capsule opacification, posterior capsule neovascularization, cataract surgery, postoperative complications, intravitreal bevacizumab

  20. Radiation embrittlement of PWR vessel supports

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Robinson, G.C.; Pennell, W.E.; Nanstad, R.K.

    1989-01-01

    Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

  1. Reaction-in-flight neutrons as a signature for shell mixing in National Ignition Facility capsules

    International Nuclear Information System (INIS)

    Hayes, A. C.; Bradley, P. A.; Grim, G. P.; Jungman, Gerard; Wilhelmy, J. B.

    2010-01-01

    Analytic calculations and results from computational simulations are presented that suggest that reaction-in-flight (RIF) neutrons can be used to diagnose mixing of the ablator shell material into the fuel in deuterium-tritium (DT) capsules designed for the National Ignition Facility (NIF) [J. A. Paisner, J. D. Boyes, S. A. Kumpan, W. H. Lowdermilk, and M. S. Sorem, Laser Focus World 30, 75 (1994)]. Such mixing processes in NIF capsules are of fundamental physical interest and can have important effects on capsule performance, quenching the total thermonuclear yield. The sensitivity of RIF neutrons to hydrodynamical mixing arises through the dependence of RIF production on charged-particle stopping lengths in the mixture of DT fuel and ablator material. Since the stopping power in the plasma is a sensitive function of the electron temperature and density, it is also sensitive to mix. RIF production scales approximately inversely with the degree of mixing taking place, and the ratio of RIF to down-scattered neutrons provides a measure of the mix fraction and/or the mixing length. For sufficiently high-yield capsules, where spatially resolved RIF images may be possible, neutron imaging could be used to map RIF images into detailed mix images.

  2. Chitosan Derivatives/Calcium Carbonate Composite Capsules Prepared by the Layer-by-Layer Deposition Method

    Directory of Open Access Journals (Sweden)

    Takashi Sasaki

    2008-01-01

    Full Text Available Core/shell capsules composed of calcium carbonate whisker core (rod-like shape and chitosan/chitosansulfate shell were prepared by the layer-by-layer deposition technique. Two chitosan samples of different molecular weights (Mw=9.7×104 and 1.09×106g·mol-1 were used as original materials. Hollow capsules were also obtained by dissolution of the core in hydrochloric acid. Electron microscopy revealed that the surface of the shell is rather ragged associated with some agglomerates. The shell thickness l obeys a linear relation with respect to the number of deposited layers m as l=md+a(a>0. The values of d (thickness per layer were 4.0 and 1.0 nm for the higher and lower Mw chitosan materials, respectively, both of which are greater than the thickness of the monolayer. The results suggest that the feature of the deposition does not obey an ideal homogeneous monolayer-by-monolayer deposition mechanism. Shell crosslinked capsules were also prepared via photodimerization reaction of cinnamoyl groups after a deposition of cinnamoyl chitosan to the calcium carbonate whisker core. The degree of crosslink was not enough to stabilize the shell structure, and hollow capsule was not obtained.

  3. Effects of food on a gastrically degraded drug: azithromycin fast-dissolving gelatin capsules and HPMC capsules.

    Science.gov (United States)

    Curatolo, William; Liu, Ping; Johnson, Barbara A; Hausberger, Angela; Quan, Ernest; Vendola, Thomas; Vatsaraj, Neha; Foulds, George; Vincent, John; Chandra, Richa

    2011-07-01

    Commercial azithromycin gelatin capsules (Zithromax®) are known to be bioequivalent to commercial azithromycin tablets (Zithromax®) when dosed in the fasted state. These capsules exhibit a reduced bioavailability when dosed in the fed state, while tablets do not. This gelatin capsule negative food effect was previously proposed to be due to slow and/or delayed capsule disintegration in the fed stomach, resulting in extended exposure of the drug to gastric acid, leading to degradation to des-cladinose-azithromycin (DCA). Azithromycin gelatin capsules were formulated with "superdisintegrants" to provide fast-dissolving capsules, and HPMC capsule shells were substituted for gelatin capsule shells, in an effort to eliminate the food effect. Healthy volunteers were dosed with these dosage forms under fasted and fed conditions; pharmacokinetics were evaluated. DCA pharmacokinetics were also evaluated for the HPMC capsule subjects. In vitro disintegration of azithromycin HPMC capsules in media containing food was evaluated and compared with commercial tablets and commercial gelatin capsules. When the two fast-dissolving capsule formulations were dosed to fed subjects, the azithromycin AUC was 38.9% and 52.1% lower than after fasted-state dosing. When HPMC capsules were dosed to fed subjects, the azithromycin AUC was 65.5% lower than after fasted-state dosing. For HPMC capsules, the absolute fasting-state to fed-state decrease in azithromycin AUC (on a molar basis) was similar to the increase in DCA AUC. In vitro capsule disintegration studies revealed extended disintegration times for commercial azithromycin gelatin capsules and HPMC capsules in media containing the liquid foods milk and Ensure®. Interaction of azithromycin gelatin and HPMC capsules with food results in slowed disintegration in vitro and decreased bioavailability in vivo. Concurrent measurement of serum azithromycin and the acid-degradation product DCA demonstrates that the loss of azithromycin

  4. Effect of microstructure on the impact toughness and temper embrittlement of SA508Gr.4N steel for advanced pressure vessel materials.

    Science.gov (United States)

    Yang, Zhiqiang; Liu, Zhengdong; He, Xikou; Qiao, Shibin; Xie, Changsheng

    2018-01-09

    The effect of microstructure on the impact toughness and the temper embrittlement of a SA508Gr.4N steel was investigated. Martensitic and bainitic structures formed in this material were examined via scanning electron microscopy, electron backscatter diffraction, transmission electron microscopy, and Auger electron spectroscopy (AES) analysis. The martensitic structure had a positive effect on both the strength and toughness. Compared with the bainitic structure, this structure consisted of smaller blocks and more high-angle grain boundaries (HAGBs). Changes in the ultimate tensile strength and toughness of the martensitic structure were attributed to an increase in the crack propagation path. This increase resulted from an increased number of HAGBs and refinement of the sub-structure (block). The AES results revealed that sulfur segregation is higher in the martensitic structure than in the bainitic structure. Therefore, the martensitic structure is more susceptible to temper embrittlement than the bainitic structure.

  5. NCSX Vacuum Vessel Fabrication

    International Nuclear Information System (INIS)

    Viola ME; Brown T; Heitzenroeder P; Malinowski F; Reiersen W; Sutton L; Goranson P; Nelson B; Cole M; Manuel M; McCorkle D.

    2005-01-01

    The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in conjunction with the Oak Ridge National Laboratory (ORNL). The goal of this experiment is to develop a device which has the steady state properties of a traditional stellarator along with the high performance characteristics of a tokamak. A key element of this device is its highly shaped Inconel 625 vacuum vessel. This paper describes the manufacturing of the vessel. The vessel is being fabricated by Major Tool and Machine, Inc. (MTM) in three identical 120 o vessel segments, corresponding to the three NCSX field periods, in order to accommodate assembly of the device. The port extensions are welded on, leak checked, cut off within 1-inch of the vessel surface at MTM and then reattached at PPPL, to accommodate assembly of the close-fitting modular coils that surround the vessel. The 120 o vessel segments are formed by welding two 60 o segments together. Each 60 o segment is fabricated by welding ten press-formed panels together over a collapsible welding fixture which is needed to precisely position the panels. The vessel is joined at assembly by welding via custom machined 8-inch (20.3 cm) wide spacer ''spool pieces''. The vessel must have a total leak rate less than 5 X 10 -6 t-l/s, magnetic permeability less than 1.02(micro), and its contours must be within 0.188-inch (4.76 mm). It is scheduled for completion in January 2006

  6. Method of detecting water leakage in radioactive waste containing vessel

    International Nuclear Information System (INIS)

    Ishioka, Hitoshi; Takao, Yoshiaki; Hayakawa, Kiyoshige.

    1989-01-01

    Lower level radioactive wastes formed upon operation of nuclear facilities are processed by underground storage. In this case, a plurality of drum cans packed with radioactive wastes are contained in a vessel and a water soluble dye material is placed at the inside of the vessel. The method of placing the water soluble dye material at the inside of the vessel includes a method of coating the material on the inner surface of the vessel and a method of mixing the material in sands to be filled between each of the drum cans. Then, leakage of water soluble dye material is detected when water intruding from the outside into the vessel is again leached out of the vessel, to detect the water leakage from the inside of the vessel. In this way, it is possible to find a water-invaded vessel before corrosion of the drum can by water intruded into the vessel and leakage of nuclides in the drum can. Accordingly, it is possible to apply treatment such as repair before occurrence of accident and can maintain the safety of radioactive water processing facilities. (I.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  9. Motion of an elastic capsule in a square microfluidic channel.

    Science.gov (United States)

    Kuriakose, S; Dimitrakopoulos, P

    2011-07-01

    In the present study we investigate computationally the steady-state motion of an elastic capsule along the centerline of a square microfluidic channel and compare it with that in a cylindrical tube. In particular, we consider a slightly over-inflated elastic capsule made of a strain-hardening membrane with comparable shearing and area-dilatation resistance. Under the conditions studied in this paper (i.e., small, moderate, and large capsules at low and moderate flow rates), the capsule motion in a square channel is similar to and thus governed by the same scaling laws with the capsule motion in a cylindrical tube, even though in the channel the cross section in the upstream portion of large capsules is nonaxisymmetric (i.e., square-like with rounded corners). When the hydrodynamic forces on the membrane increase, the capsule develops a pointed downstream edge and a flattened rear (possibly with a negative curvature) so that the restoring tension forces are increased as also happens with droplets. Membrane tensions increase significantly with the capsule size while the area near the downstream tip is the most probable to rupture when a capsule flows in a microchannel. Because the membrane tensions increase with the interfacial deformation, a suitable Landau-Levich-Derjaguin-Bretherton analysis reveals that the lubrication film thickness h for large capsules depends on both the capillary number Ca and the capsule size a; our computations determine the latter dependence to be (in dimensionless form) h ~ a(-2) for the large capsules studied in this work. For small and moderate capsule sizes a, the capsule velocity Ux and additional pressure drop ΔP+ are governed by the same scaling laws as for high-viscosity droplets. The velocity and additional pressure drop of large thick capsules also follow the dynamics of high-viscosity droplets, and are affected by the lubrication film thickness. The motion of our large thick capsules is characterized by a Ux-U ~ h ~ a(-2

  10. Final processing vessel for radioactive waste

    International Nuclear Information System (INIS)

    Tejima, Takaya; Hiraki, Akimitsu.

    1989-01-01

    An inorganic inner layer comprising dense inorganic material such as organic polymer-impregnated concretes is formed to about 10 - 50 mm in average thickness at the inside of a metal vessel. Further, the surface of the vessel is formed as a flat surface with no or only small reinforcing protrusions. Thus, if the final processing vessel should be dropped during transportation or handling by mistake, since impact shocks do not concentrate to protrusions as usual, no local stress concentration occurs to the inorganic inner liner layer. Accordingly, the risk of rapture can be reduced greatly. Further, since impact shock resistance layer put between the metal vessel and the inorganic inner liner layer absorbs shocks, a further sufficient strength can be obtained against dropping accident. (T.M.)

  11. Polar tent for reduced perturbation of NIF ignition capsules

    Science.gov (United States)

    Hammel, B. A.; Pickworth, L.; Stadermann, M.; Field, J.; Robey, H.; Scott, H. A.; Smalyuk, V.

    2016-10-01

    In simulations, a tent that contacts the capsule near the poles and departs tangential to the capsule surface greatly reduces the capsule perturbation, and the resulting mass injected into the hot-spot, compared to current capsule support methods. Target fabrication appears feasible with a layered tent (43-nm polyimide + 8-nm C) for increased stiffness. We are planning quantitative measurements of the resulting shell- ρR perturbation near peak implosion velocity (PV) using enhanced self-emission backlighting, achieved by adding 1% Ar to the capsule fill in Symcaps (4He + H). Layered DT implosions are also planned for an integrated test of capsule performance. We will describe the design and simulation predictions. Prepared by LLNL under Contract DE-AC52-07NA27344.

  12. Capsule-odometer: a concept to improve accurate lesion localisation.

    Science.gov (United States)

    Karargyris, Alexandros; Koulaouzidis, Anastasios

    2013-09-21

    In order to improve lesion localisation in small-bowel capsule endoscopy, a modified capsule design has been proposed incorporating localisation and - in theory - stabilization capabilities. The proposed design consists of a capsule fitted with protruding wheels attached to a spring-mechanism. This would act as a miniature odometer, leading to more accurate lesion localization information in relation to the onset of the investigation (spring expansion e.g., pyloric opening). Furthermore, this capsule could allow stabilization of the recorded video as any erratic, non-forward movement through the gut is minimised. Three-dimensional (3-D) printing technology was used to build a capsule prototype. Thereafter, miniature wheels were also 3-D printed and mounted on a spring which was attached to conventional capsule endoscopes for the purpose of this proof-of-concept experiment. In vitro and ex vivo experiments with porcine small-bowel are presented herein. Further experiments have been scheduled.

  13. Cheboygan Vessel Base

    Data.gov (United States)

    Federal Laboratory Consortium — Cheboygan Vessel Base (CVB), located in Cheboygan, Michigan, is a field station of the USGS Great Lakes Science Center (GLSC). CVB was established by congressional...

  14. High Performance Marine Vessels

    CERN Document Server

    Yun, Liang

    2012-01-01

    High Performance Marine Vessels (HPMVs) range from the Fast Ferries to the latest high speed Navy Craft, including competition power boats and hydroplanes, hydrofoils, hovercraft, catamarans and other multi-hull craft. High Performance Marine Vessels covers the main concepts of HPMVs and discusses historical background, design features, services that have been successful and not so successful, and some sample data of the range of HPMVs to date. Included is a comparison of all HPMVs craft and the differences between them and descriptions of performance (hydrodynamics and aerodynamics). Readers will find a comprehensive overview of the design, development and building of HPMVs. In summary, this book: Focuses on technology at the aero-marine interface Covers the full range of high performance marine vessel concepts Explains the historical development of various HPMVs Discusses ferries, racing and pleasure craft, as well as utility and military missions High Performance Marine Vessels is an ideal book for student...

  15. 2011 Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  16. 2011 Fishing Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  17. 2013 Tanker Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  18. Maury Journals - US Vessels

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — U.S. vessels observations, after the 1853 Brussels Conference that set International Maritime Standards, modeled after Maury Marine Standard Observations.

  19. Coastal Logbook Survey (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains catch (landed catch) and effort for fishing trips made by vessels that have been issued a Federal permit for the Gulf of Mexico reef fish,...

  20. Reactor pressure vessel support

    International Nuclear Information System (INIS)

    Butti, J.P.

    1977-01-01

    A link and pin support system provides the primary vertical and lateral support for a nuclear reactor pressure vessel without restricting thermally induced radial and vertical expansion and contraction. (Auth.)

  1. 2013 Cargo Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  2. 2013 Fishing Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  3. 2013 Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  4. Ocean Station Vessel

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Ocean Station Vessels (OSV) or Weather Ships captured atmospheric conditions while being stationed continuously in a single location. While While most of the...

  5. Vessel Sewage Discharges

    Science.gov (United States)

    Vessel sewage discharges are regulated under Section 312 of the Clean Water Act, which is jointly implemented by the EPA and Coast Guard. This homepage links to information on marine sanitation devices and no discharge zones.

  6. Nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

  7. Starting procedure for internal combustion vessels

    Science.gov (United States)

    Harris, Harry A.

    1978-09-26

    A vertical vessel, having a low bed of broken material, having included combustible material, is initially ignited by a plurality of ignitors spaced over the surface of the bed, by adding fresh, broken material onto the bed to buildup the bed to its operating depth and then passing a combustible mixture of gas upwardly through the material, at a rate to prevent back-firing of the gas, while air and recycled gas is passed through the bed to thereby heat the material and commence the desired laterally uniform combustion in the bed. The procedure permits precise control of the air and gaseous fuel mixtures and material rates, and permits the use of the process equipment designed for continuous operation of the vessel.

  8. Reactor pressure vessel design

    International Nuclear Information System (INIS)

    Foehl, J.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 2, the general principles of reactor pressure vessel design are elaborated. Crack and fracture initiation and propagation are treated in some detail

  9. Graywater Discharges from Vessels

    Science.gov (United States)

    2011-11-01

    metals (e.g., cadmium, chromium, lead, copper , zinc, silver, nickel, and mercury), solids, and nutrients (USEPA, 2008b; USEPA 2010). Wastewater from... flotation ), and disinfection (using ultraviolet light) as compared to traditional Type II MSDs that use either simple maceration and chlorination, or...Coliform Naval Vessels Oceanographic Vessels Small Cruise Ships 25a Vendor 2 Hamann AG Biological Treatment with Dissolved Air Flotation and

  10. Non-destructive tests of capsules for JMTR irradiation examination

    International Nuclear Information System (INIS)

    Tanaka, Hidetaka; Nagao, Yoshiharu; Sato, Masashi; Osawa, Kenji

    2007-03-01

    Irradiation examination are increasing in advanced irradiation research for accurate prediction control and evaluation of irradiation parameter such as neutron fluence, etc. by using JMTR. Irradiation capsule internals are therefore structurally complicated recently. This report described the procedure of non destructive tests such as radiographic test, penetrant test, ultrasonic test, etc. for inspection of irradiation capsules in JMTR, and the result of Test-case of confirmation procedure for internal parts of irradiation capsules. (author)

  11. Regional Variation Is Present in Elbow Capsules after Injury

    OpenAIRE

    Germscheid, Niccole M.; Hildebrand, Kevin A.

    2006-01-01

    Myofibroblast numbers and α-smooth muscle actin expression are increased in anterior joint capsules of patients with posttraumatic elbow contractures. The purpose of our study was to determine whether these changes occur regionally or throughout the entire joint capsule. We hypothesized that the α-smooth muscle actin mRNA expression and the myofibroblast numbers in posterior joint capsules would be elevated in elbows obtained from patients with posttraumatic joint contractures compared with j...

  12. Myofibroblast Numbers are Elevated in Human Elbow Capsules After Trauma

    OpenAIRE

    Hildebrand, Kevin A.; Zhang, Mei; van Snellenberg, Wistara; King, Graham J. W.; Hart, David A.

    2004-01-01

    Elbow contractures, a frequent problem after injury, can be treated by excision of the joint capsule. However, the underlying changes in the joint capsule are poorly understood. Based on skin healing work, we examined the hypotheses that myofibroblast numbers and expression of a myofibroblast marker α-smooth muscle actin, are elevated in patients with posttraumatic joint contractures. Anterior capsules were obtained from six patients who had operative release of posttraumatic contractures gre...

  13. Capsule Shields the Function of Short Bacterial Adhesins

    OpenAIRE

    Schembri, Mark A.; Dalsgaard, Dorte; Klemm, Per

    2004-01-01

    Bacterial surface structures such as capsules and adhesins are generally regarded as important virulence factors. Here we demonstrate that capsules block the function of the self-recognizing protein antigen 43 through physical shielding. The phenomenon is not restricted to Escherichia coli but can occur in other gram-negative bacteria. Likewise, we show that other short adhesins exemplified by the AIDA-I protein are blocked by the presence of a capsule. The results support the notion that cap...

  14. Introduction to the physics of ICF capsules

    International Nuclear Information System (INIS)

    Lindl, J.D.

    1989-01-01

    Inertial Confinement Fusion is an approach to fusion which relies on the inertia of the fuel mass to provide confinement. To achieve conditions under which this confinement is sufficient for efficient thermonuclear burn, high gain ICF targets designed to be imploded directly by laser light. These capsules are generally a spherical shell which is filled with low density DT gas. The shell is composed of an outer region which forms the ablator and an inner region of frozen or liquid DT which forms the main fuel. Energy from the driver is delivered to the ablator which heats up and expands. As the ablator expands and blows outward, the rest of the shell is forced inward to conserve momentum. In this implosion process, several features are important. We define the in-flight-aspect-ratio (IFAR) as the ratio of the shell radius R as it implodes to its thickness ΔR. Hydrodynamic instabilities during the implosion impose limits on this ratio which results in a minimum pressure requirement of about 100 Mbar. The convergence ratio is defined as the ratio of the initial outer radius of the ablator to the final compressed radius of the hot spot. This hot spot is the central region of the compressed fuel which is required to ignite the main fuel in high gain designs. Typical convergence ratios are 30--40. To maintain a nearly spherical shape during the implosion, when convergence ratios are this large, the flux delivered to the capsule must be uniform to a few percent. The remainder of this paper discusses the conditions necessary to achieve thermonuclear ignition in these ICF capsules

  15. Characterization of an aged WESF capsule

    International Nuclear Information System (INIS)

    Kenna, B.T.; Schultz, F.J.

    1983-07-01

    A joint effort by SNLA and ORNL was initiated for a detailed characterization of an 18-year-old WESF 137 Cs source which has been used in the Sandia Irradiator for Dried Sewage Solids. The study included evaluation of the inner and outer stainless steel capsules by optical metallography, electron microprobe, and physical testing. Analysis of the residual atmospheres within the two containers was also done. The CsCl was analyzed for isotopic content and impurities. No potential problem areas, including corrosion, were found

  16. Postirradiation examination of capsule GF-4

    International Nuclear Information System (INIS)

    Kovacs, W.J.; Sedlak, B.J.

    1980-10-01

    The GF-4 capsule test was irradiated in the SILOE reactor at Grenoble, France between April 8, 1975 and July 26, 1976. High-enriched uranium (HEU) UC 2 and weak acid resin (WAR) UC/sub x/O/sub y/ fissile and ThO 2 fertile particles were tested. Postirradiation examination of cured-in-place fuel rods showed no fuel rod/graphite element interaction. In addition, all rods exhibited adequate structural integrity. Irradiation-induced dimensional changes for rods containing all TRISO-coated fuel were consistent with model predictions; however, rods containing BISO-coated fuel exhibited greater volumetric contractions than predicted

  17. Stability analysis of directly driven Nif capsules

    Energy Technology Data Exchange (ETDEWEB)

    Goncharov, V.N.; Skupsky, S.; McKenty, P.W.; Delettrez, J.A.; Town, R.P.J. [Rochester Univ., NY (United States). Lab. for Laser Energetics; Cherfils-Clerouin, C. [CEA/DAM-Ile de France, DIF, 91 - Bruyeres Le Chatel (France)

    2000-07-01

    An analytical model is presented to study perturbation evolution at the ablation and inner surfaces of the imploding shell. The model describes the ablative Rayleigh-Taylor and Bell-Plesset instabilities. The initial conditions for the model are determined by using existing theories of laser imprint, ablative Richtmyer-Meshkov instability, 'feed-out' and by performing a series of 2-D ORCHID simulations. The model and simulations show that the direct-drive cryogenic {alpha} = 3 NIF capsules remain intact during the implosion if laser nonuniformities are smoothed by 2-D SSD used in the current direct-drive target designs. (authors)

  18. FFTF and CRBRP reactor vessels

    International Nuclear Information System (INIS)

    Morgan, R.E.

    1977-01-01

    The Fast Flux Test Facility (FFTF) reactor vessel and the Clinch River Breeder Reactor Plant (CRBRP) reactor vessel each serve to enclose a fast spectrum reactor core, contain the sodium coolant, and provide support and positioning for the closure head and internal structure. Each vessel is located in its reactor cavity and is protected by a guard vessel which would ensure continued decay heat removal capability should a major system leak develop. Although the two plants have significantly different thermal power ratings, 400 megawatts for FFTF and 975 megawatts for CRBRP, the two reactor vessels are comparable in size, the CRBRP vessel being approximately 28% longer than the FFTF vessel. The FFTF vessel diameter was controlled by the space required for the three individual In-Vessel Handling Machines and Instrument Trees. Utilization of the triple rotating plug scheme for CRBRP refueling enables packaging of the larger CRBRP core in a vessel the same diameter as the FFTF vessel

  19. Building a polysaccharide hydrogel capsule delivery system for control release of ibuprofen.

    Science.gov (United States)

    Chen, Zhi; Wang, Ting; Yan, Qing

    2018-02-01

    Development of a delivery system which can effectively carry hydrophobic drugs and have pH response is becoming necessary. Here we demonstrate that through preparation of β-cyclodextrin polymer (β-CDP), a hydrophobic drug molecule of ibuprofen (IBU) was incorporated into our prepared β-CDP inner cavities, aiming to improve the poor water solubility of IBU. A core-shell capsule structure has been designed for achieving the drug pH targeted and sustained release. This delivery system was built with polysaccharide polymer of Sodium alginate (SA), sodium carboxymethylcellulose (CMC) and hydroxyethyl cellulose (HEC) by physical cross-linking. The drug pH-response control release is this hydrogel system's chief merit, which has potential value for synthesizing enteric capsule. Besides, due to our simple preparing strategy, optimal conditions can be readily determined and the synthesis process can be accurately controlled, leading to consistent and reproducible hydrogel capsules. In addition, phase-solubility method was used to investigate the solubilization effect of IBU by β-CDP. SEM was used to prove the forming of core and shell structure. FT-IR and 1 H-NMR were also used to perform structural characteristics. By the technique of UV determination, the pH targeted and sustained release study were also performed. The results have proved that our prepared polysaccharide hydrogel capsule delivery system has potential applications as oral drugs delivery in the field of biomedical materials.

  20. Design procedure of capsule with multistage heater control (named MUSTAC)

    International Nuclear Information System (INIS)

    Someya, Hiroyuki; Endoh, Yasuichi; Hoshiya, Taiji; Niimi, Motoji; Harayama, Yasuo

    1990-11-01

    A capsule with electric heaters at multistage (named MUSTAC) is a type of capsule used in JMTR. The heaters are assembled in the capsule. Supply electric current to the heaters can be independently adjusted with a control systems that keeps irradiation specimens to constant temperature. The capsule being used, the irradiation specimen are inserted into specimen holders. Gas-gap size, between outer surface of specimen holders and inner surface of capsule casing, is calculated and determined to be flatten temperature of loaded specimens over the region. The rise or drop of specimen temperature in accordance with reactor power fluctuations is corrected within the target temperature of specimen by using the heaters filled into groove at specimen holder surface. The present report attempts to propose a reasonable design procedure of the capsules by means of compiling experience for designs, works and irradiation data of the capsules and to prepare for useful informations against onward capsule design. The key point of the capsule lies on thermal design. Now design thermal calculations are complicated in case of specimen holder with multihole. Resolving these issues, it is considered from new on that an emphasis have to placed on settling a thermal calculation device, for an example, a computer program on calculation specimen temperature. (author)

  1. Safety evaluation for packaging (onsite) singly encapsulated cesium chloride capsules

    International Nuclear Information System (INIS)

    Smyth, W.W.

    1997-01-01

    Three nonstandard Waste Encapsulation and Storage Facility (WESF) cesium chloride capsules are being shipped from WESF (225B building) to the 324 building. They would normally be shipped in the Beneficial Uses Shipping System (BUSS) cask under its US Department of Energy (DOE) license (DOE 1996), but these capsules are nonstandard: one has a damaged or defective weld in the outer layer of encapsulation, and two have the outer encapsulation removed. The 3 capsules, along with 13 other capsules, will be overpacked in the 324 building to meet the requirements for storage in WESF's pool

  2. Increasing Z-pinch vacuum hohlraum capsule coupling efficiency

    International Nuclear Information System (INIS)

    Callahan, Debbie; Vesey, Roger Alan; Cochrane, Kyle Robert; Nikroo, A.; Bennett, Guy R.; Schroen, Diana Grace; Ruggles, Laurence E.; Porter, John L.; Streit, Jon; Mehlhorn, Thomas Alan; Cuneo, Michael Edward

    2004-01-01

    Symmetric capsule implosions in the double-ended vacuum hohlraum (DEH) on Z have demonstrated convergence ratios of 14-21 for 2.15-mm plastic ablator capsules absorbing 5-7 kJ of x-rays, based on backlit images of the compressed ablator remaining at peak convergence (1). Experiments with DD-filled 3.3-mm diameter capsules designed to absorb 14 kJ of x-rays have begun as an integrated test of drive temperature and symmetry, complementary to thin-shell symmetry diagnostic capsules. These capsule implosions are characterized by excellent control of symmetry (< 3% time-integrated), but low hohlraum efficiency (< 2%). Possible methods to increase the capsule absorbed energy in the DEH include mixed-component hohlraums, large diameter foam ablator capsules, transmissive shine shields between the z-pinch and capsule, higher spoke electrode x-ray transmission, a double-sided power feed, and smaller initial radius z-pinch wire arrays. Simulations will explore the potential for each of these modifications to increase the capsule coupling efficiency for near-term experiments on Z and ZR

  3. 21 CFR 520.1920 - Prochlorperazine, isopropamide sustained release capsules.

    Science.gov (United States)

    2010-04-01

    ... HUMAN SERVICES (CONTINUED) ANIMAL DRUGS, FEEDS, AND RELATED PRODUCTS ORAL DOSAGE FORM NEW ANIMAL DRUGS... dogs in which gastrointestinal disturbances are associated with emotional stress. (2)(i) Capsules...

  4. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels (Final Report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, J. H.; Lee, B. S.; Chi, S. H.; Kim, J. H.; Oh, Y. J.; Yoon, J. H.; Kwon, S. C.; Park, D. G.; Kang, Y. H.; Choo, K. N.; Oh, J. M.; Park, S. J.; Kim, B. K.; Shin, Y. T.; Cho, M. S.; Sohn, J. M.; Kim, D. S.; Choo, Y. S.; Ahn, S. B.; Oh, W. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-05-01

    Reactor pressure vessel materials, which were produced by a domestic company, Doosan Heavy Industries and construction Co., Ltd., have been evaluated using the neutron irradiation facility HANARO. For this evaluation, instrumented capsules were used for neutron irradiation of various kinds of specimens made of different heats of steels, which are VCD(Y4), VCD+Al(U4), Si+Al(Y5), U4 weld metal, and U4 HAZ, respectively. The fast neutron fluence levels ranged 1 to 5 (x10{sup 19} n/cm{sup 2}, E>1MeV) depending on the specimens and the irradiation temperature was controlled within 290{+-}10 deg C. The test results showed that, in the ranking of the material properties of the base metals, both before and after neutron irradiation, Y5 is the best, U4 the next and Y4 the last. Y4 showed a substantial change by neutron irradiation as well as the properties was worse than others in the unirradiated state. However, Y5, which showed the best properties in unirradiated state, was also the best in the resistance for irradiation embrittlement and one can hardly detect the property change after irradiation. The weldment showed a reasonably good resistance to irradiation embrittlement while the unirradiated properties were worse than base metals. The RPV steels are all expected to meet the screening criteria of the USNRC codes and regulations during the end of plant life. 39 refs., 42 figs., 27 tabs. (Author)

  5. Radiation embrittlement of WWER-1000 reactor vessel steels

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  6. Analyses of copper from a prototype capsule 5 and 6; Analyser av koppar fraan prototypkapsel 5 och 6

    Energy Technology Data Exchange (ETDEWEB)

    Taxen, Claes; Lundholm, Martin; Persson, Dan; Jakobsson, Dan; Sedlakova, Miroslava; Randelius, Mats; Karlsson, Oskar; Rydgren, Pontus; Kimab, Swerea

    2012-12-15

    'Prototype' is a series of experiments where SKB, the Swedish Nuclear Fuel and Waste Management Co, expose the full scale copper canisters under conditions intended to be representative of a repository for spent nuclear fuel, however, without radioactivity (SKB 2012) . Copper from one of these installations, deposition 5, has been studied for corrosion . Samples were also taken from the capsule that had been exposed in deposition 6. Drill cores across the capsule wall has been documented regarding microstructure. All samples have been exposed for about seven years in the prototype repository. Studies carried out leads to the following conclusions: Regarding the ring on top of the capsule from the deposition 5; There are local corrosion with a depth of 3-5 microns. The general or uniform corrosion that has occurred can not be quantified. The relatively sharp traces of processing of the material before exposure indicates that the general corrosion was minor. Small amounts of corrosion product has been detected in surface analysis. The surface profile on the copper surface, aside from the grooves after processing and areas of local corrosion, are relatively even. Metallographic examination of cross section shows no tendency to pitting or intergranular corrosion. Analysis for hydrogen by melting a quantity of metal does not show any increased hydrogen content. Regarding the material of the capsule from the deposition 6: The capsule has not been specifically tested for corrosion. Cross sections of drill cores through the copper canister has been documented and metallo graphically exhibits nothing remarkable.

  7. Generation of colloidal granules and capsules from double emulsion drops

    Science.gov (United States)

    Hess, Kathryn S.

    Assemblies of colloidal particles are extensively used in ceramic processing, pharmaceuticals, inks and coatings. In this project, the aim was to develop a new technique to fabricate monodispersed colloidal assemblies. The use of microfluidic devices and emulsion processing allows for the fabrication of complex materials that can be used in a variety of applications. A microfluidic device is used to create monodispersed water/oil/water (w/o/w) double emulsions with interior droplets of colloidal silica suspension ranging in size from tens to hundreds of microns. By tailoring the osmotic pressure using glycerol as a solute in the continuous and inner phases of the emulsion, we can control the final volume size of the monodispersed silica colloidal crystals that form in the inner droplets of the double emulsion. Modifying the ionic strength in the colloidal dispersion can be used to affect the particle-particle interactions and crystal formation of the final colloidal particle. This w/o/w technique has been used with other systems of metal oxide colloids and cellulose nanocrystals. Encapsulation of the colloidal suspension in a polymer shell for the generation of ceramic-polymer core-shell particles has also been developed. These core-shell particles have spawned new research in the field of locally resonant acoustic metamaterials. Systems and chemistries for creating cellulose hydrogels within the double emulsions have also been researched. Water in oil single emulsions and double emulsions have been used to create cellulose hydrogel spheres in the sub-100 micron diameter range. Oil/water/oil double emulsions allow us to create stable cellulose capsules. The addition of a second hydrogel polymer, such as acrylate or alginate, further strengthens the cellulose gel network and can also be processed into capsules and particles using the microfluidic device. This work could have promising applications in acoustic metamaterials, personal care products, pharmaceuticals

  8. Potassium iodide capsule treatment of feline sporotrichosis.

    Science.gov (United States)

    Reis, Erica G; Gremião, Isabella D F; Kitada, Amanda A B; Rocha, Raphael F D B; Castro, Verônica S P; Barros, Mônica B L; Menezes, Rodrigo C; Pereira, Sandro A; Schubach, Tânia M P

    2012-06-01

    Sporotrichosis is a mycosis caused by Sporothrix schenckii. The most affected animal is the cat; it has played an important role in the zoonotic transmission of this disease, especially in Rio de Janeiro, Brazil, since 1998. In order to evaluate the treatment of feline sporotrichosis with potassium iodide, an observational cohort was conducted in 48 cats with sporotrichosis at Instituto de Pesquisa Clínica Evandro Chagas, Fiocruz. All cats received potassium iodide capsules, 2.5 mg/kg to 20 mg/kg q24h. The cure rate was 47.9%, treatment failure was 37.5%, treatment abandonment was 10.4% and death was 4.2%. Clinical adverse effects were observed in 52.1% of the cases. Thirteen cats had a mild increase in hepatic transaminase levels during the treatment, six of them presented clinical signs suggestive of hepatotoxicity. Compared to previous studies with itraconazole and iodide in saturated solution, potassium iodide capsules are an alternative for feline sporotrichosis treatment.

  9. Petrous bone fractures violating otic capsule.

    Science.gov (United States)

    Magliulo, Giuseppe; Ciniglio Appiani, Mario; Iannella, Giannicola; Artico, Marco

    2012-12-01

    This study presents our experience with a series of patients suffering from petrous bone fractures violating the otic capsule who underwent subtotal petrosectomy combined with eustachian tube, middle ear, and mastoid obliteration, with the goal of preventing cerebrospinal fluid (CSF) leak and meningitis. This study enrolled 26 patients between 1997 and 2011. The clinical symptoms, otoscopy, and preoperative and postoperative audiometry and facial function, as well as CSF leak or meningitis, were evaluated in each patient. The entire group underwent a subtotal petrosectomy using the technique described in detail by Fisch. In addition, each patient was interviewed using a questionnaire to evaluate the impact on quality of life. Intraoperatively, we found significant CSF leaks in 14 patients (42.5%). No patient reported other episodes of CSF leak or meningitis after the surgery. The patients' responses of facial nerve function were slightly worse than the House-Brackmann evaluation (50% versus 42.3%; p < 0.05). The vast majority (88.5%) of the patients experienced no social impact. Our findings suggest the importance of not underestimating the risk for CSF leak in the petrous bone fractures violating the otic capsule. Preoperative counseling regarding the various troublesome complications must adequately motivate candidates to undergo surgery by pointing out the positive impact of the proposed treatment.

  10. Potential value of Cs-137 capsules

    Energy Technology Data Exchange (ETDEWEB)

    Bloomster, C.H.; Brown, D.R.; Bruno, G.A.; Hazelton, R.F.; Hendrickson, P.L.; Lezberg, A.J.; Tingey, G.L.; Wilfert, G.L.

    1985-04-01

    We determined the value of Cs-137 compared to Co-60 as a source for the irradiation of fruit (apples and cherries), pork and medical supplies. Cs-137, in the WESF capsule form, had a value of approximately $0.40/Ci as a substitute for Co-60 priced at approximately $1.00/Ci. The comparison was based on the available curies emitted from the surface of each capsule. We developed preliminary designs for fourteen irradiation facilities; seven were based on Co-60 and seven were based on Cs-137. These designs provided the basis for estimating capital and operating costs which, in turn, provided the basis for determining the value of Cs-137 relative to Co-60 in these applications. We evaluated the effect of the size of the irradiation facility on the value of Cs-137. The cost of irradiation is low compared to the value of the product. Irradiation of apples for disinfestation costs $.01 to .02 per pound. Irradiation for trichina-safe pork costs $.02 per pound. Irradiation of medical supplies for sterilization costs $.07 to .12 per pound. The cost of the irradiation source, either Co-60 or Cs-137, contributed only a minor amount to the total cost of irradiation, about 5% for the fruit and hog cases and about 20% for the medical supply cases. We analyzed the sensitivity of the irradiation costs and Cs-137 value to several key assumptions.

  11. Potential value of Cs-137 capsules

    International Nuclear Information System (INIS)

    Bloomster, C.H.; Brown, D.R.; Bruno, G.A.; Hazelton, R.F.; Hendrickson, P.L.; Lezberg, A.J.; Tingey, G.L.; Wilfert, G.L.

    1985-04-01

    We determined the value of Cs-137 compared to Co-60 as a source for the irradiation of fruit (apples and cherries), pork and medical supplies. Cs-137, in the WESF capsule form, had a value of approximately $0.40/Ci as a substitute for Co-60 priced at approximately $1.00/Ci. The comparison was based on the available curies emitted from the surface of each capsule. We developed preliminary designs for fourteen irradiation facilities; seven were based on Co-60 and seven were based on Cs-137. These designs provided the basis for estimating capital and operating costs which, in turn, provided the basis for determining the value of Cs-137 relative to Co-60 in these applications. We evaluated the effect of the size of the irradiation facility on the value of Cs-137. The cost of irradiation is low compared to the value of the product. Irradiation of apples for disinfestation costs $.01 to .02 per pound. Irradiation for trichina-safe pork costs $.02 per pound. Irradiation of medical supplies for sterilization costs $.07 to .12 per pound. The cost of the irradiation source, either Co-60 or Cs-137, contributed only a minor amount to the total cost of irradiation, about 5% for the fruit and hog cases and about 20% for the medical supply cases. We analyzed the sensitivity of the irradiation costs and Cs-137 value to several key assumptions

  12. Materialism.

    Science.gov (United States)

    Melnyk, Andrew

    2012-05-01

    Materialism is nearly universally assumed by cognitive scientists. Intuitively, materialism says that a person's mental states are nothing over and above his or her material states, while dualism denies this. Philosophers have introduced concepts (e.g., realization and supervenience) to assist in formulating the theses of materialism and dualism with more precision, and distinguished among importantly different versions of each view (e.g., eliminative materialism, substance dualism, and emergentism). They have also clarified the logic of arguments that use empirical findings to support materialism. Finally, they have devised various objections to materialism, objections that therefore serve also as arguments for dualism. These objections typically center around two features of mental states that materialism has had trouble in accommodating. The first feature is intentionality, the property of representing, or being about, objects, properties, and states of affairs external to the mental states. The second feature is phenomenal consciousness, the property possessed by many mental states of there being something it is like for the subject of the mental state to be in that mental state. WIREs Cogn Sci 2012, 3:281-292. doi: 10.1002/wcs.1174 For further resources related to this article, please visit the WIREs website. Copyright © 2012 John Wiley & Sons, Ltd.

  13. Procurement of replacement pressure vessels for MURR

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  14. Reinforcing effects of caffeine in coffee and capsules.

    Science.gov (United States)

    Griffiths, R R; Bigelow, G E; Liebson, I A

    1989-09-01

    In a residential research ward the reinforcing and subjective effects of caffeine were studied under double-blind conditions in volunteer subjects with histories of heavy coffee drinking. In Experiment 1, 6 subjects had 13 opportunities each day to self-administer either a caffeine (100 mg) or a placebo capsule for periods of 14 to 61 days. All subjects developed a clear preference for caffeine, with intake of caffeine becoming relatively stable after preference had been attained. Preference for caffeine was demonstrated whether or not preference testing was preceded by a period of 10 to 37 days of caffeine abstinence, suggesting that a recent history of heavy caffeine intake (tolerance/dependence) was not a necessary condition for caffeine to function as a reinforcer. In Experiment 2, 6 subjects had 10 opportunities each day to self-administer a cup of coffee or (on different days) a capsule, dependent upon completing a work requirement that progressively increased and then decreased over days. Each day, one of four conditions was studied: caffeinated coffee (100 mg/cup), decaffeinated coffee, caffeine capsules (100 mg/capsule), or placebo capsules. Caffeinated coffee maintained the most self-administration, significantly higher than decaffeinated coffee and placebo capsules but not different from caffeine capsules. Both decaffeinated coffee and caffeine capsules were significantly higher than placebo capsules but not different from each other. In both experiments, subject ratings of "linking" of coffee or capsules covaried with the self-administration measures. These experiments provide the clearest demonstrations to date of the reinforcing effects of caffeine in capsules and in coffee.

  15. Acrylic vessel cleaning tests

    International Nuclear Information System (INIS)

    Earle, D.; Hahn, R.L.; Boger, J.; Bonvin, E.

    1997-01-01

    The acrylic vessel as constructed is dirty. The dirt includes blue tape, Al tape, grease pencil, gemak, the glue or residue form these tapes, finger prints and dust of an unknown composition but probably mostly acrylic dust. This dirt has to be removed and once removed, the vessel has to be kept clean or at least to be easily cleanable at some future stage when access becomes much more difficult. The authors report on the results of a series of tests designed: (a) to prepare typical dirty samples of acrylic; (b) to remove dirt stuck to the acrylic surface; and (c) to measure the optical quality and Th concentration after cleaning. Specifications of the vessel call for very low levels of Th which could come from tape residues, the grease pencil, or other sources of dirt. This report does not address the concerns of how to keep the vessel clean after an initial cleaning and during the removal of the scaffolding. Alconox is recommended as the cleaner of choice. This acrylic vessel will be used in the Sudbury Neutrino Observatory

  16. Uncertainty study of the PWR pressure vessel fluence. Adjustment of the nuclear data base; Etudes des incertitudes sur la fluence dans les cuves des REP. Ajustement des donnees de base

    Energy Technology Data Exchange (ETDEWEB)

    Kodeli, I A

    1994-01-01

    The code system devoted to the calculation of the sensitivity and uncertainty of of the neutron flux and reaction rates calculated by the transport codes, has been developed. Adjustment of the basic data to experimental results can be performed as well. Various sources of uncertainties can be taken into account, such as those due to the uncertainties in the cross-sections, response functions, fission spectrum and space distribution of neutron source, geometry and material composition uncertainties... One -As well as two- dimensional analysis can be performed. Linear perturbation theory is applied. The code system is sufficiently general to be used for various analysis in the fields of fission and fusion. The principal objective of our studies concerns the capsule dosimetry study realized in the framework of the 900 MWe PWR pressure vessel surveillance program. The analysis indicates that the present calculations, performed by the code TRIPOLI-2, using the ENDF/B-IV based, non-perturbed neutron cross-section library in 315 energy groups, allows to estimate the neutron flux and the reaction rates in the surveillance capsules and in the most calculated and measured reaction rates permits to reduce these uncertainties. The results obtained with the adjusted iron cross-sections, response functions and fission spectrum show that the agreement between the calculation and the experiment was improved to become within 10% approximately. The neutron flux deduced from the experiment is then extrapolated from the capsule to the most exposed pressure vessel location using the calculated lead factor. The uncertainty in this factor was estimated to be about 7%. (author). 39 refs., 52 figs., 30 tabs.

  17. Comparison of hydrophobic and hydrophilic intraocular lens in preventing posterior capsule opacification after cataract surgery

    OpenAIRE

    Zhao, Yang; Yang, Ke; Li, Jiaxin; Huang, Yang; Zhu, Siquan

    2017-01-01

    Abstract Background: Posterior capsular opacification (PCO) is a common long-term complication of cataract surgery. Intraocular lens design and material have been implicated in influencing the development of PCO. This study evaluated the association of hydrophobic and hydrophilic intraocular lenses on preventing PCO. Methods: Medline, Cochrane, EMBASE, and Google Scholar databases were searched until August 3, 2016, using the following search terms: cataract, posterior capsule opacification, ...

  18. Improved nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

  19. The pressure vessel for the NSF tandem

    International Nuclear Information System (INIS)

    Jones, C.W.

    1979-04-01

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

  20. Crack propagation on spherical pressure vessels

    International Nuclear Information System (INIS)

    Lebey, J.; Roche, R.

    1975-01-01

    The risk presented by a crack on a pressure vessel built with a ductile steel cannot be well evaluated by simple application of the rules of Linear Elastic Fracture Mechanics, which only apply to brittle materials. Tests were carried out on spherical vessels of three different scales built with the same steel. Cracks of different length were machined through the vessel wall. From the results obtained, crack initiation stress (beginning of stable propagation) and instable propagation stress may be plotted against the lengths of these cracks. For small and medium size, subject to ductile fracture, the resulting curves are identical, and may be used for ductile fracture prediction. Brittle rupture was observed on larger vessels and crack propagation occurred at lower stress level. Preceedings curves are not usable for fracture analysis. Ultimate pressure can be computed with a good accuracy by using equivalent energy toughness, Ksub(1cd), characteristic of the metal plates. Satisfactory measurements have been obtained on thin samples. The risks of brittle fracture may then judged by comparing Ksub(1cd) with the calculated K 1 value, in which corrections for vessel shape are taken into account. It is thus possible to establish the bursting pressure of cracked spherical vessels, with the help of two rules, one for brittle fracture, the other for ductile instability. A practical method is proposed on the basis of the work reported here

  1. Design, fabrication and irradiation test report on HANARO instrumented capsule (05M-07U) for the researches of universities in 2005

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-09-15

    As a part of the 2005 project for an active utilization of HANARO, an instrumented capsule (05M-07U) was designed, fabricated and irradiated for an irradiation test of various unclear materials under irradiation conditions which was requested by external researchers from universities. The basic structure of the 05M-07U capsule was based on the 00M-01U, 01M-05U, 02M-05U, 03M-06U and 04M-07U capsules which had been successfully irradiated in HANARO as part of the 2000, 2001, 2002, 2003 and 2004 projects. However, because of a limited number of specimens and the budget of one university, the remaining space in the capsule was filled with various KAERI specimens for researches on a nuclear core and SMART materials, and parts of a nuclear fuel assembly of KNFC. Various types of specimens such as tensile, Charpy, TEM, hardness, compression and growth specimens made of Zr 702, Ti and Ni alloys, Zirlo, Inconel, STS 316L and Cr-Mo alloys were placed in the capsule. Especially, this capsule was designed to evaluate the nuclear characteristics of the parts of a nuclear fuel assembly and the Ti tubes in HANARO. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. The capsule was irradiated in the CT test hole of HANARO of a 30MW thermal output at 270 ∼ 400 .deg. C up to a fast neutron fluence of 5.7 x 10{sup 20} (n/cm{sup 2}) (E >1.0MeV). The obtained results will be very valuable for the related research of the users.

  2. Out-pile test of the capsule with cone shape bottom structures

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-01-01

    The design modification of bottom guide structures for the instrumented capsule which is used for the irradiation test in the research reactor, HANARO is done because of the cutting trouble of the bottom guide arm's pin. The previous structure of the 3-pin arm shape is changed into one body of the cone shape. The specimens of the bottom end cap ring with three different sizes ({phi}68mm, {phi}70mm, {phi}72mm) are designed and manufactured. The out-pile test for the capsule with previous 3-pin arm and new three bottom structures of the cone shape is performed using the one-channel flow test facilities. In order to estimate the compatibility with HANARO, the structural stability and integrity of the capsule, the out-pile test such as a loading/unloading test, a pressure drop test, a thermal performance test, a displacement measurement due to a vibration and an endurance test etc. is conducted, and the outer diameter of the bottom end cap ring to meet the HANARO requirements is selected. From out-pile test results the capsule with cone shape bottom structures is evaluated as to have the structural stability and the benefit from the fluid's flow respect. Also the size satisfied various requirements among three kinds of bottom end cap rings is 70mm in diameter. It is expected that the new bottom structures of the cone shape with 70mm in diameter will be applicable to all material and special capsules which will be designed and manufactured for the purpose of irradiation tests in the future.

  3. Thermal Behavior of the Reactor Vessel Penetration Under External Vessel Cooling During a Severe Accident

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Park, Rae-Joon; Kim, Jong-Tae; Min, Byung-Tae; Lee, Ki-Young; Kim, Sang-Baik

    2004-01-01

    Experimental and analytical studies on the thermal behavior of reactor vessel penetration have been performed under external vessel cooling during a severe accident in the Korean next-generation reactor APR1400. Two types of tests, SUS-EXT and SUS-DRY with and without external vessel cooling, respectively, have been performed using sustained heating by an induction heater. Three tests have been carried out varying the cooling conditions at the vessel outer surface in the SUS-EXT tests. The experimental results have been thermally estimated using the LILAC computer code. The experimental results indicate that the inner surface of the vessel was ablated by the 45-mm thickness in the SUS-DRY test. Despite the total ablation of the welding material, the penetration was not ejected outside the vessel, which could be attributed to the thermal expansion of the penetration. Unlike the SUS-DRY test, the thickness of the ablation was ∼15 to 20 mm at most, so the welding was preserved in the SUS-EXT tests. It is concluded from the experimental results that the external vessel cooling highly affected the ablation configuration and the thermal behaviors of the vessel and the penetration. An increase in coolant mass flow rate from 0.047 to 0.152 kg/s had effects on the thermal behavior of the lower head vessel and penetration in the SUS-EXT tests. The LILAC analytical results on temperature distribution and ablation depth in the lower head vessel and penetration were very similar to the experimental results

  4. Colloidal micro- and nano-particles as templates for polyelectrolyte multilayer capsules.

    Science.gov (United States)

    Parakhonskiy, Bogdan V; Yashchenok, Alexey M; Konrad, Manfred; Skirtach, Andre G

    2014-05-01

    Colloidal particles play an important role in various areas of material and pharmaceutical sciences, biotechnology, and biomedicine. In this overview we describe micro- and nano-particles used for the preparation of polyelectrolyte multilayer capsules and as drug delivery vehicles. An essential feature of polyelectrolyte multilayer capsule preparations is the ability to adsorb polymeric layers onto colloidal particles or templates followed by dissolution of these templates. The choice of the template is determined by various physico-chemical conditions: solvent needed for dissolution, porosity, aggregation tendency, as well as release of materials from capsules. Historically, the first templates were based on melamine formaldehyde, later evolving towards more elaborate materials such as silica and calcium carbonate. Their advantages and disadvantages are discussed here in comparison to non-particulate templates such as red blood cells. Further steps in this area include development of anisotropic particles, which themselves can serve as delivery carriers. We provide insights into application of particles as drug delivery carriers in comparison to microcapsules templated on them. Copyright © 2014 Elsevier B.V. All rights reserved.

  5. Radioactive liquid containing vessel

    International Nuclear Information System (INIS)

    Sakurada, Tetsuo; Kawamura, Hironobu.

    1993-01-01

    Cooling jackets are coiled around the outer circumference of a container vessel, and the outer circumference thereof is covered with a surrounding plate. A liquid of good conductivity (for example, water) is filled between the cooling jackets and the surrounding plate. A radioactive liquid is supplied to the container vessel passing through a supply pipe and discharged passing through a discharge pipe. Cooling water at high pressure is passed through the cooling water jackets in order to remove the heat generated from the radioactive liquid. Since cooling water at high pressure is thus passed through the coiled pipes, the wall thickness of the container vessel and the cooling water jackets can be reduced, thereby enabling to reduce the cost. Further, even if the radioactive liquid is leaked, there is no worry of contaminating cooling water, to prevent contamination. (I.N.)

  6. Test and evaluation of pressure vessel materials

    International Nuclear Information System (INIS)

    Choi, Sun Pil; Hong, Jun Hwa; Nho, Kye Hoe; Han, Dae June; Chi, Se Hwan

    1985-01-01

    We have prepared a method for analyzing the Charpy impact test data, which is deduced from ''the standard anelastic solid equation''. The theoretical expression for the absorbed energy is in a form of W=Wsub(U)+(Wsub(R)-Wsub(U))/ [1+(ωtau) 2 ] showing the Debye characteristics and where tau is given by the Arrhenius equation; tau=tau 0 exp(ΔH/ksub(B)T). Four measurable parameters, at the present stage, can characterize the dynamic hehavior of cracking (Charpy impact result). They are the upper shelf energy(Wsub(R), the lower shelf energy (Wsub(U)), the activation energy of crack (ΔH, and wtau(0) where w tau(0) are the resonance frequency of the specimen and the jumping pre-exponential factor of propagating crack respectively. However the states of R (relaxed) and U (un-relaxed) should be defined from reasonable physical conditions in the future and it is possible that Wsub(U) is small enough to be taken as zero. The effects of irradiation, alloying elements, and heat treatment on the impact results should be interpreted as changes in the above characteristic parameters. The present method has been applied for weld metal of SA 508-2 irradiated up to a fluence of 4x10 18 n/cm 2 , E>1.0Mev, resulting in about 29% decrease in Wsub(R), negligible change in Wsub(U), 5.6 times increase in ωtau 0 , and no change in ΔH. This seems to indicate that irradiation degrades an average value of YOUNG's modulus so that cracks propagate more easily and it does not effect on breaking the lattice bond. However much more systematic analyses should be necessary for correct judgment. It is concluded that the present method is quite adequate for analyzing the Charpy impact data even though plastic deformation in the specimen was not considered separately so that the method should be applied for various cases in order to evaluate the proper trend of effects of irradiation, alloying elements, and heat treatment on the Charpy impact results. (Author)

  7. Seismic analysis for shroud facility in-pile tube and saturated temperature capsules

    International Nuclear Information System (INIS)

    Iimura, Koichi; Yamaura, Takayuki; Ogawa, Mitsuhiro

    2009-07-01

    At Oarai Research and Development Center, Japan Atomic Energy Agency (JAEA), the plan of repairing and refurbishing Japan Materials Testing Reactor (JMTR) has progressed in order to restart JMTR operation in the fiscal 2011. As a part of effective use of JMTR, the neutron irradiation tests of LWR fuels and materials has been planned in order to study their soundness. By using Oarai Shroud Facility (OSF-1) and Fuel Irradiation Facility with the He-3 gas control system for power lamping test using Boiling Water Capsules (BOCA Irradiation Facility), the irradiation tests with power ramping will be carried out to study the soundness of fuel under LWR Transient condition. OSF-1 is the irradiation facility of shroud type that can insert and eject the capsule under reactor operation, and is composed of 'In-pile Tube', 'Cooling system' and 'Capsule exchange system'. BOCA Irradiation Facility is the facility which simulates irradiation environment of LWR, and is composed of 'Boiling water Capsule', 'Capsule control system' and 'Power control system by He-3'. By using Saturated temperature Capsules and the water environment control system, the material irradiation tests under the water chemistry condition of LWR will be carried out to clarify the mechanism of IASCC. In JMTR, these facilities are in service at the present. However, the detailed design for renewal or remodeling was carried out based on the new design condition in order to be correspondent to the irradiation test plan after restart JMTR operation. In this seismic analysis of the detailed design, each equipment classification and operating state were arranged with 'Japanese technical standards of the structure on nuclear facility for test research' and 'Technical guidelines for seismic design of nuclear power plants on current, and then, stress calculation and evaluation were carried out by FEM piping analysis code 'SAP' and structure analysis code 'ABAQUS'. About the stress of the seismic force, it was proven

  8. Capsule shields the function of short bacterial adhesins

    DEFF Research Database (Denmark)

    Schembri, Mark; Dalsgaard, D.; Klemm, Per

    2004-01-01

    Bacterial surface structures such as capsules and adhesins are generally regarded as important virulence factors. Here we demonstrate that capsules block the function of the self-recognizing protein antigen 43 through physical shielding. The phenomenon is not restricted to Escherichia coli but can...

  9. Iodometric determination of ampicillin in proprietary capsules | Ejele ...

    African Journals Online (AJOL)

    The concentration of ampicillin in ampicillin capsule preparations purchased in Owerri main market, Imo State of Nigeria, was determined using the iodometric titration method. The results showed that the ampicillin concentrations in the capsules contained between 250 and 260 mg/cap of ampicillin trihydrate. Statistical ...

  10. The development of the nasal capsule of the silver carp ...

    African Journals Online (AJOL)

    The development of the nasal capsule, including the procartilaginous as well as some mesenchymatous developmental stages, is described in the chinese silver carp,. Hypophthalmichthys molitrix (VaL). In the nasal capsule a primitive feature, namely the presence of a complete paraphysial bridge was observed.

  11. Temporary presence of myofibroblasts in human elbow capsule after trauma

    NARCIS (Netherlands)

    Doornberg, Job N.; Bosse, Tjalling; Cohen, Mark S.; Jupiter, Jesse B.; Ring, David; Kloen, Peter

    2014-01-01

    Elbow stiffness is a common complication after elbow trauma. The elbow capsule is often thickened, fibrotic, and contracted at the time of surgical release. The limited studies available suggest that the capsule is contracted because of fibroblast-to-myofibroblast differentiation. We hypothesize

  12. Evidence for an intact polysaccharide capsule in Bordetella pertussis.

    Science.gov (United States)

    Neo, YiLin; Li, Rui; Howe, Josephine; Hoo, Regina; Pant, Aakanksha; Ho, SiYing; Alonso, Sylvie

    2010-03-01

    Polysaccharide capsules contribute to the pathogenesis of many bacteria species by providing resistance against various defense mechanisms. The production of a capsule in Bordetella pertussis, the etiologic agent of whooping cough, has remained controversial; earlier studies reported this pathogen as a capsulated microorganism whereas the recent B. pertussis genome analysis revealed the presence of a truncated capsule locus. In this work, using transmission electron microscopy and immunostaining approaches, we provide a formal evidence for the presence of an intact microcapsule produced at the surface of both laboratory strain and clinical isolates of B. pertussis. In agreement with previous studies, we found that the capsule is optimally produced in avirulent phase. Unexpectedly, the presence of the capsule was also detected at the surface of virulent B. pertussis bacteria. Consistently, a substantial transcriptional activity of the capsule operon was detected in virulent phase, suggesting that the capsular polysaccharide may play a role during pertussis pathogenesis. In vitro assays indicated that the presence of the capsule does not affect B. pertussis adherence to mammalian cells and does not further protect the bacterium from phagocytosis, complement-mediated killing or antimicrobial peptide attack. Copyright 2009. Published by Elsevier SAS.

  13. Clinical effect of Resina Draconis capsules on primary dysmenorrhoea

    African Journals Online (AJOL)

    Clinical effect of Resina Draconis capsules on primary dysmenorrhoea. Li Sun, Jia Wang. Abstract. Purpose: To examine the effectiveness of Resina Draconis capsules in the treatment of primary dysmenorrhoea. Methods: In total, 324 patients with primary dysmenorrhoea were randomly allocated to three groups based on ...

  14. Reactor pressure vessel thermal annealing

    International Nuclear Information System (INIS)

    Lee, A.D.

    1997-01-01

    The steel plates and/or forgings and welds in the beltline region of a reactor pressure vessel (RPV) are subject to embrittlement from neutron irradiation. This embrittlement causes the fracture toughness of the beltline materials to be less than the fracture toughness of the unirradiated material. Material properties of RPVs that have been irradiated and embrittled are recoverable through thermal annealing of the vessel. The amount of recovery primarily depends on the level of the irradiation embrittlement, the chemical composition of the steel, and the annealing temperature and time. Since annealing is an option for extending the service lives of RPVs or establishing less restrictive pressure-temperature (P-T) limits; the industry, the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC) have assisted in efforts to determine the viability of thermal annealing for embrittlement recovery. General guidance for in-service annealing is provided in American Society for Testing and Materials (ASTM) Standard E 509-86. In addition, the American Society of Mechanical Engineers (ASME) Code Case N-557 addresses annealing conditions (temperature and duration), temperature monitoring, evaluation of loadings, and non-destructive examination techniques. The NRC thermal annealing rule (10 CFR 50.66) was approved by the Commission and published in the Federal Register on December 19, 1995. The Regulatory Guide on thermal annealing (RG 1.162) was processed in parallel with the rule package and was published on February 15, 1996. RG 1.162 contains a listing of issues that need to be addressed for thermal annealing of an RPV. The RG also provides alternatives for predicting re-embrittlement trends after the thermal anneal has been completed. This paper gives an overview of methodology and recent technical references that are associated with thermal annealing. Results from the DOE annealing prototype demonstration project, as well as NRC activities related to the

  15. On the hydrostatic test for nuclear vessels

    International Nuclear Information System (INIS)

    Palmero, A.

    1979-01-01

    A comparison of the pressure test requirements, namely specified values of pressure and temperature, for nuclear vessels designed and constructed according to the ASME Code and Spanish Rules is presented. Also the relationship of the design criteria and the pressure test requirements is indicated with a particular emphasis on the test temperature in order to avoid brittle behaviour of the materials. (author)

  16. Adsorption and desorption of phosphorus in ceramic capsules

    International Nuclear Information System (INIS)

    Almeida, J.R.F. de.

    1983-01-01

    Experiments were carried out in order to analyse the capacity of adsorving P from water using ceramic capsules with 32P, in the presence and absence of water flow through the capsule. Also studied was the desorption of 32 P from the capsule in water, with and without water flow. The desorption of residual 32 P was analysed by isotopic exchange with 31 P, also with and without water flow. It was observed that, in the presence of a flow, the capsule retained 32 P from the solution, which was weakly desorbed by water but was isotopically exchanged with 31 P. In the absence of a flow, the capsule was not an efficient P adsorber. (Author) [pt

  17. Three-dimensional simulations of Nova capsule implosion experiments

    International Nuclear Information System (INIS)

    Marinak, M.M.; Tipton, R.E.; Landen, O.L.

    1995-01-01

    Capsule implosion experiments carried out on the Nova laser are simulated with the three-dimensional HYDRA radiation hydrodynamics code. Simulations of ordered near single mode perturbations indicate that structures which evolve into round spikes can penetrate farthest into the hot spot. Bubble-shaped perturbations can burn through the capsule shell fastest, however, causing even more damage. Simulations of a capsule with multimode perturbations shows spike amplitudes evolving in good agreement with a saturation model during the deceleration phase. The presence of sizable low mode asymmetry, caused either by drive asymmetry or perturbations in the capsule shell, can dramatically affect the manner in which spikes approach the center of the hot spot. Three-dimensional coupling between the low mode shell perturbations intrinsic to Nova capsules and the drive asymmetry brings the simulated yields into closer agreement with the experimental values

  18. No stabilizing effect of the elbow joint capsule. A kinematic study

    DEFF Research Database (Denmark)

    Nielsen, K K; Olsen, Bo Sanderhoff

    1999-01-01

    We dissected 7 cadaveric elbow specimens, leaving the collateral ligaments and the joint capsule intact. The anterior and the posterior capsule were sequentially transected, followed by kinematic testings. We found no change in joint laxity after total transection of the capsule.......We dissected 7 cadaveric elbow specimens, leaving the collateral ligaments and the joint capsule intact. The anterior and the posterior capsule were sequentially transected, followed by kinematic testings. We found no change in joint laxity after total transection of the capsule....

  19. Investigation of impulsively loaded pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N.; Cornwell, R.; Hanner, D.; Leichter, H.; Mohr, P.

    1963-10-15

    Explosion containment vessels for containing from 2,000 to 3,000 five ton nuclear explosions are considered. Analysis methods appear adequate and lowest weights using the most advanced materials available in the next five years are projected.None of these materials can be fabricated today and all require extensive development. Present material technology limits the choice of materials and defines the weight. The addition of safety factors and fixtures (nozzles, etc.) will add to this weight considerably, and may well radically alter the vessel response. Improvements in the strength weight ratios of metals and glasses over those considered in this report do not appear reasonable at this time. Winding schemes to utilize the high strength of steel wires and somehow maintain a reasonable thickness appear to offer the most promise. A `ductile` beryllium would of course offer vast improvement, but no indications that this is being developed have appeared and all presently known beryllium is much too brittle.

  20. High-speed precision weighing of pharmaceutical capsules

    International Nuclear Information System (INIS)

    Bürmen, Miran; Pernuš, Franjo; Likar, Boštjan

    2009-01-01

    In this paper, we present a cost-effective method for fast and accurate in-line weighing of hard gelatin capsules based on the optimized capacitance sensor and real-time processing of the capsule capacitance profile resulting from 5000 capacitance measurements per second. First, the effect of the shape and size of the capacitive sensor on the sensitivity and stability of the measurements was investigated in order to optimize the performance of the system. The method was tested on two types of hard gelatin capsules weighing from 50 mg to 650 mg. The results showed that the capacitance profile was exceptionally well correlated with the capsule weight with the correlation coefficient exceeding 0.999. The mean precision of the measurements was in the range from 1 mg to 3 mg, depending on the size of the capsule and was significantly lower than the 5% weight tolerances usually used by the pharmaceutical industry. Therefore, the method was found feasible for weighing pharmaceutical hard gelatin capsules as long as certain conditions are met regarding the capsule fill properties and environment stability. The proposed measurement system can be calibrated by using only two or three sets of capsules with known weight. However, for most applications it is sufficient to use only empty and nominally filled capsules for calibration. Finally, a practical application of the proposed method showed that a single system is capable of weighing around 75 000 capsules per hour, while using multiple systems could easily increase the inspection rate to meet almost any requirements