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Sample records for ceramic breeder blanket

  1. Proceedings of the fifteenth international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    This report is the Proceedings of 'the Fifteenth International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors. This workshop was held in Sapporo, Japan on 3-4, Sept. 2009. Twenty six participants from EU, Japan, India, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket development. By this workshop, advance of key technologies for solid breeder blanket development was shared among the participants. Also, desired direction of further investigation and development was recognized. The 20 of the presented papers are indexed individually. (J.P.N.)

  2. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    This report is the Proceedings of ''the Sixth International Workshop on Ceramic Breeder Blanket Interactions'' which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: 1) fabrication and characterization of ceramic breeders, 2) properties data for ceramic breeders, 3) tritium release characteristics, 4) modeling of tritium behavior, 5) irradiation effects on performance behavior, 6) blanket design and R and D requirements, 7) hydrogen behavior in materials, and 8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li2TiO3, tritium release behavior of Li2TiO3 and Li2ZrO3 including tritium diffusion, modeling of tritium release from Li2ZrO3 in ITER condition, helium release behavior from Li2O, results of tritium release irradiation tests of Li4SiO4 pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  3. Proceedings of the eleventh international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    This report is the Proceedings of 'the Eleventh International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. This workshop was held in Tokyo, Japan on December 15-17, 2003. About thirty experts from China, EU, Japan, Korea, Latvia, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of Li2TiO3 and so on, fabrication technology developments and characterization of the Li2TiO3 and Li4SiO4 pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li2TiO3 and Li4SiO4 pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system. The 26 of the presented papers are indexed individually. (J.P.N.)

  4. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji [ed.

    1998-03-01

    This report is the Proceedings of `the Sixth International Workshop on Ceramic Breeder Blanket Interactions` which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: (1) fabrication and characterization of ceramic breeders, (2) properties data for ceramic breeders, (3) tritium release characteristics, (4) modeling of tritium behavior, (5) irradiation effects on performance behavior, (6) blanket design and R and D requirements, (7) hydrogen behavior in materials, and (8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li{sub 2}TiO{sub 3}, tritium release behavior of Li{sub 2}TiO{sub 3} and Li{sub 2}ZrO{sub 3} including tritium diffusion, modeling of tritium release from Li{sub 2}ZrO{sub 3} in ITER condition, helium release behavior from Li{sub 2}O, results of tritium release irradiation tests of Li{sub 4}SiO{sub 4} pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  5. Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Zhu, Qingjun; Li, Jia; Liu, Songlin

    2016-07-01

    In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  6. Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio, E-mail: enoeda.mikio@jaea.go.jp [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Suzuki, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Ezato, Koichiro; Seki, Yohji; Yoshikawa, Akira; Tsuru, Daigo; Akiba, Masato [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan)

    2012-08-15

    The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.

  7. R and D status on Water Cooled Ceramic Breeder Blanket Technology

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio, E-mail: enoeda.mikio@jaea.go.jp; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Yokoyama, Kenji

    2014-10-15

    Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li{sub 2}TiO{sub 3} pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.

  8. Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Gao, Fangfang; Zhang, Xiaokang; Pu, Yong; Zhu, Qingjun; Liu, Songlin

    2016-08-01

    Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor (CFETR) operating on a Deuterium-Tritium (D-T) fuel cycle. It is necessary to study the tritium breeding ratio (TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder (WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket, the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code (MCNP) and the fusion activation code FISPACT-2007. The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation. In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015GB108002, and 2014GB119000), and by National Natural Science Foundation of China (No. 11175207)

  9. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    International Nuclear Information System (INIS)

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits. (fusion engineering)

  10. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    Science.gov (United States)

    Ma, Xuebin; Liu, Songlin; Li, Jia; Pu, Yong; Chen, Xiangcun

    2014-04-01

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.

  11. Tauro: a ceramic composite structural material self-cooled Pb-17Li breeder blanket concept

    International Nuclear Information System (INIS)

    The use of a low-activation (LA) ceramic composite (CC) as structural material appears essential to demonstrate the potential of fusion power reactors for being inherently or, at least, passively safe. Tauro is a self-cooled Pb-17Li breeder blanket with a SiC/SiC composite as structure. This study determines the required improvements for existing industrial LA composites (mainly SiC/SiC) in order to render them acceptable for blanket operating conditions. 3D SiC/SiC CC, recently launched on the market, is a promising candidate. A preliminary evaluation of a possible joining technique for SiC/SiC is also described. (orig.)

  12. Considerations on techniques for improving tritium confinement in helium-cooled ceramic breeder blankets

    International Nuclear Information System (INIS)

    Tritium control issues such as the development of permeation barriers and the choice of the coolant and purge-gas chemistry are of crucial importance for solid breeder blankets. In order to quantify these problems for the helium-cooled ceramic BIT blanket concept, the tritium leakage into the coolant was evaluated and the consequent tritium losses into the steam circuit were determined. Our results indicate that under certain specified conditions the total tritium release from the coolant can be limited to approximately 10 Ci/d, but only on the assumption that experimental data for tritium permeation barriers can be attained under realistic operating conditions. An experimental study on the impact of the gas chemistry on tritium losses is proposed. (orig.)

  13. A ceramic breeder in a poloidal tube blanket for a tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Amici, A.; Anzidei, L.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.; Zampaglione, V.; Petrizzi, L. (Associazione Euratom-CNEN sulla Fusione, Centro di Frascati (Italy))

    1989-04-01

    A conceptual study of a helium-cooled solid breeder blanket for a tokamak reactor is presented. Tritium breeding capability together with system reliability are taken as the main design criteria. The blanket consists of tubular poloidal modules made of a central bundle of ceramic rods ({gamma}LiAlO/sub 2/) with a coaxial distribution of the inlet/outlet coolant flow (He) surrounded by a multiplier material (Be) in the form of bored bricks. The Be to {gamma}LiAlO/sub 2/ volume ratio is 4/1. The He inlet and outlet branches are cooling Be and {gamma}LiAlO/sub 2/, respectively. A purge He flow running through small central holes of the ceramic rods is derived from the main flow. Under the typical conditions of a tokamak reactor (neutron wall load=2 MW/m/sup 2/), a full coverage tritium breeding ratio of 1.47 is achieved for the following design and operating parameters: outlet He temperature=570/sup 0/C; inlet He temperature=250/sup 0/; total extracted power=2700 MW; He pumping power percentage=2%; minimum/maximum {gamma}LiAlO/sub 2/ temperature=400/900/sup 0/C; maximum structural temperature=475/sup 0/C; and maximum Be temperature=525/sup 0/C. (orig.).

  14. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    The BOT (Breeder Outside Tube) Solid Breeder Blanket for a fusion DEMO reactor is presented. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. In the paper the reference blanket design and external loops are described as well as the results of the theoretical and experimental work in the fields of neutronics, thermohydraulics, mechanical stresses, tritium control and extraction, development and irradiation of the ceramic breeder material, beryllium development, ferromagnetic forces caused by disruptions, safety and reliability. An outlook is given on the remaining open questions and on the required R and D program. (orig.)

  15. Electrical behaviour of ceramic breeder blankets in pebble form after γ-radiation

    Directory of Open Access Journals (Sweden)

    E. Carella

    2015-07-01

    Full Text Available Lithium orthosilicate (Li4SiO4 ceramics in from of pebble bed is the European candidate for ITER testing HCPB (Helium Cooled Pebble Bed breeding modules. The breeder function and the shielding role of this material, represent the areas upon which attention is focused. Electrical measurements are proposed for monitoring the modification created by ionizing radiation and at the same time provide information on lithium movement in this ceramic structure. The electrical tests are performed on pebbles fabricated by Spray-dryer method before and after gamma-irradiation through a 60Co source to a fluence of 4.8 Gy/s till a total dose of 5 ∗ 105 Gy. The introduction of thermal annealing treatments during the electrical impedance spectroscopy (EIS measurements points out the recombination effect of the temperature on the γ-induced defects.

  16. RF test blanket sub-module with ceramic breeder and helium cooling for test in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kovalenko, V. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation)]. E-mail: koval@nikiet.ru; Kapyshev, V. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Leshukov, A. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Poliksha, V. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Shatalov, G. [Russian Research Center ' Kurchatov Institute' , Kurchatov Square 1, 123182 Moscow (Russian Federation); Strebkov, Yu. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Strizhov, A. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Sviridenko, M. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation)

    2006-02-15

    International thermonuclear experimental reactor (ITER) is anticipated as the only one step to DEMO fusion reactor. One of its main objectives is to demonstrate the availability and integration of technologies essential for a fusion reactor by testing of components for a future reactor including the test blanket modules (TBM) with different types of breeding materials. RF proposed to divide the TBM on two parts and to use two independent test blanket sub-modules (TBSM) which fixed on the frame in ITER horizontal experimental port for testing. CHC TBSM design description, its mechanical attachment on the frame, and principle schemes of helium cooling system and tritium cycle system are presented in this paper.

  17. ITER solid breeder blanket materials database

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States); Dienst, W. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Material- und Festkoerperforschung; Flament, T. [CEA Centre d`Etudes de Fontenay-aux-Roses (France). Commissariat A L`Energie Atomique; Lorenzetto, P. [NET Team, Garching (Germany); Noda, K. [Japan Atomic Energy Research Inst., Takai, Ibaraki, (Japan); Roux, N. [CEA Centre d`Etudes et de Recherches Les Materiaux (France). Commissariat a L`Energie Atomique

    1993-11-01

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

  18. Assessment of the activation, decay heat, and waste disposal of the US helium-cooled ceramic breeder test blanket module in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Youssef, Mahmoud Z. [University of California-Los Angeles, Los Angeles, CA 90025 (United States); University of Wisconsin-Madison (United States)], E-mail: youssef@fusion.ucla.edu; Ying, Alice [University of California-Los Angeles, Los Angeles, CA 90025 (United States)

    2008-12-15

    The radioactivity inventory and decay heat in the US helium-cooled ceramic breeder (HCCB) test blanket module (TBM) have been assessed at shutdown and for several times thereafter. The sub-module will have its own FW and structural container box that houses the breeder and beryllium pebble bed units, arranged in an edge-on-configuration. Low activation ferritic steel (F82H) is used as the structure and helium is used as a coolant. The breeder beds are made of Li{sub 2}TiO{sub 3} pebbles in which lithium has been enriched up to 75% in Li-6. Pulsed operation mode is assumed. During operation in the D-T phase, the total heating rate in the TBM is {approx}263 kW. The total amount of tritium generated in the breeder and the beryllium multiplier is {approx}9 g and 0.07 g, respectively, after reaching the 0.3 MWa/m{sup 2} fluence limit. At shutdown, the total radioactivity and decay heat levels are {approx}0.89 MCi and {approx}0.002 MW, respectively. These values drop sharply after 1 min to {approx}0.098 MCi and {approx}0.0006 MW. The contribution from the F82H structure is the dominant one up to {approx}10 years following shutdown. After {approx}10 years, the contribution to the total activation and decay heat from the breeder material is the dominant one due to the generated tritium. The WDR of various components are far below unity and thus are well within ITER regulatory guidelines.

  19. Development of Solid Breeder Blanket at JAERI

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been performing blanket development based on the long-term research program of fusion blankets in Japan, which was approved by the Fusion Council of Japan in 1999. The blanket development consists of out-pile R and D, In-pile R and D, TBM Neutronics and TPR Tests and Tritium Recovery System R and D. Based on the achievements of element technology development, the R and D program is now stepping to the engineering testing phase, in which scalable mockup tests will be performed for obtaining engineering data unique to the specific structure of the components, with the objective to define the fabrication specification of test blanket modules for ITER. This paper presents the major achievements of the element technology development of solid breeder blanket in JAERI

  20. A review of fusion breeder blanket technology, part 1

    International Nuclear Information System (INIS)

    This report presents the results of a study of fusion breeder blanket technology. It reviews the role of the breeder blanket, the current understanding of the scientific and engineering bases of liquid metal and solid breeder blankets and the programs now underway internationally to resolve the uncertainities in current knowledge. In view of existing national expertise and experience, a solid breeder R and D program for Canada is recommended

  1. Conceptual design of a water cooled breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Pu, Yong; Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Li, Jia; Peng, ChangHong [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Ma, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Chen, Lei [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2014-10-15

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by {sup 6}Li(n,α)T reaction. Li{sub 2}TiO{sub 3} pebbles and Be{sub 12}Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li{sub 2}TiO{sub 3} and Be{sub 12}Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be{sub 12}Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option

  2. Conceptual design of a water cooled breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by 6Li(n,α)T reaction. Li2TiO3 pebbles and Be12Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li2TiO3 and Be12Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be12Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into

  3. Assessment of the activation, decay heat, and waste disposal of the US helium-cooled ceramic breeder test blanket module in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Youssef, M.; Ying, A. [California Univ., Los Angeles, CA (United States)

    2007-07-01

    The radioactivity inventory and after heat in the U.S. helium-cooled ceramic breeder (HCCB) test blanket module (TBM) have been accessed at shut down and for several times thereafter. Also assessed is the waste disposal rating (WDR) of its various components. The objectives are: (1) to provide the information needed for further safety assessment of the generated radionuclides and their volatility, as well as after heat on the safety operation of ITER, and (2) to aid in determining the waiting cooling period prior to removing and transporting the TBM for further treatment outside ITER site. The TBM is proposed to be placed in one of the three dedicated test ports of ITER. The current proposal is that it will occupy 1/3 of the horizontal upper half of a port next to Japan and Korea sub-modules. The sub-module will have its own FW and structural container box that houses the breeder and beryllium pebble bed units, arranged in an edge-on-configuration. Helium is used to cool the FW, sides of the box, and the internal plates. Conventional ferritic steel (F82H) is used as the structure. The sub-module has 71 cm height, 38.9 cm wide and 60 cm depth in the radial direction. The breeder beds are made of Li{sub 2}TiO{sub 3} pebbles with 94% theoretical density and 62% packing factor (as the beryllium pebbles). Lithium-6 is enriched to 75%. A 2 mm thick beryllium layer is used as a plasma facing material on the FW area subjected to 0.78 MW/m{sup 2} neutron wall load. Pulsed operation mode is assumed. Each pulse is assumed to be 400 s full flat top followed by 1800 s dwell time, during which the decay of the generated radionuclides are accounted for. The 500 MW pulses are assumed to be generated one after another until a fluence limit of 0.3 MWa/m{sup 2} is reached without replacing the TBM. This gives upper conservative estimates for the radioactive inventory and decay heat. During operation in the D-T phase, the total heating rate in the TBM is {proportional_to}263 KW. The

  4. Research and development status of ceramic breeder materials

    International Nuclear Information System (INIS)

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was also recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option breeder material. Blanket design studies have indicated areas in the properties data base that need further investigation. Current studies are focusing on issues such as tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests are underway, some as part of an international collaboration for development of ceramic breeder materials. 36 refs

  5. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture

  6. Fast-Breeder-Blanket Project: FBBF. Final report

    International Nuclear Information System (INIS)

    This report is the final report for DOE contract DE-AC02-76ET37237 with the Purdue Fast Breeder Blanket Project. The Project was initiated to investigate the uncertainties in Fast Breeder Reactor blanket calculations. Absolute measurements of key neutron reaction rates, neutron spectra, and gamma-ray energy depositions were made in simulated FBF blankets in the Fast Breeder Blanket Facility (FBBF), a Cf-252 driven subcritical facility. Calculation of the spectra and integral reaction rates were made using methods, computer codes, and cross section data typical of those currently used in the design of FBR's. Comparisons of calculated to experimental integral neutron reaction rates give good agreement at the inner portions of the blanket by diverge to C/E ratios of about 0.65 at the outer edge of the blanket for reactions sensitive to the neutron density

  7. Progress in tritium retention and release modeling for ceramic breeders

    International Nuclear Information System (INIS)

    Tritium behavior in ceramic breeder blankets is a key design issue for this class of blanket because of its impact on safety and fuel self-sufficiency. Over the past 10-15 years, substantial theoretical and experimental efforts have been dedicated world-wide to develop a better understanding of tritium transport in ceramic breeders. Models that are available today seem to cover reasonably well all the key physical transport and trapping mechanisms. They have allowed for reasonable interpretation and reproduction of experimental data and have helped in pointing out deficiencies in material property data base, in providing guidance for future experiments, and in analyzing blanket tritium behavior. This paper highlights the progress in tritium modeling over the last decade. Key tritium transport mechanisms are briefly described along with the more recent and sophisticated models developed to help understand them. Recent experimental data are highlighted and model calibration and validation discussed. Finally, example applications to blanket cases are shown as illustration of progress in the prediction of ceramic breeder blanket tritium inventory

  8. Progress in tritium retention and release modeling for ceramic breeders

    International Nuclear Information System (INIS)

    Tritium behavior in ceramic breeder blankets is a key design issue for this class of blanket because of its impact on safety and fuel self-sufficiency. Over the past 10-15 years, substantial theoretical and experimental effort has been dedicated worldwide to the development of a better understanding of tritium transport in ceramic breeders. The models available today seem to cover reasonably well all of the key physical transport and trapping mechanisms. They allow for reasonable interpretation and reproduction of experimental data, help to point out deficiencies in the material property database, provide guidance for future experiments and aid in the analysis of blanket tritium behavior.This paper highlights the progress in tritium modeling over the last decade. Key tritium transport mechanisms are briefly described, together with the more recent, sophisticated models which have been developed to help understand them. Recent experimental data are highlighted and model calibration and validation are discussed. Finally, example applications to blanket cases are shown as an illustration of the progress in the prediction of ceramic breeder blanket tritium inventory. (orig.)

  9. Degrading the Plutonium Produced in Fast Breeder Reactor Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jor-Shan; Kuno, Yusuke [Tokyo University, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 113-8656 (Japan)

    2009-06-15

    Plutonium quality, defined as the plutonium isotopic composition, is an important measure for proliferation-resistance (PR) of a nuclear energy system. The quality of the plutonium produced in the blanket assemblies of a fast breeder reactor could be as good as or better than the weapons-grade (WG). The presence of such good quality plutonium is a proliferation concern. There are various options to degrade the plutonium produced in the breeder blanket. The obvious one is to blend the blanket plutonium with those produced from the reactor core during reprocessing. Other options try to prevent the generation of good quality plutonium (Pu). The Protected Plutonium Production (P{sup 3}) Project proposed by Tokyo Institute of Technology (TIT)1,2,3 advocates the doping of certain amount of neptunium (Np), or americium (Am) in fresh blanket fuel for irradiation. The increased production of {sup 238}Pu, {sup 240}Pu and {sup 242}Pu by neutron capture in {sup 237}Np and Am would degrade the blanket plutonium. However, as {sup 237}Np is a controlled material according to IAEA, its use as doping material in fresh blanket fuel presents a concern for nuclear proliferation. In addition, the fabrication of fresh blanket fuel with inclusion of americium would be complicated due to the emission of intense low-energy gamma radiation from {sup 241}Am. Am is normally accompanied by Cm since the separation of those 2 elements is very difficult. Fuel containing both Am and Cm may make Safeguards measurement difficult. A variation would be doping the fresh blanket fuel with minor actinide (e.g., a group of neptunium, americium, and curium), or with separated reactor-grade (RG) plutonium. The drawback of such schemes would be the need for glove boxes in fresh blanket fuel fabrication. It is possible to fuel the breeder blankets with recycled (reprocessed) uranium oxide. The recycled uranium, recovered from reprocessing, contains {sup 236}U, which when irradiated in the blanket would

  10. Ceramic helium-cooled blanket test module

    Energy Technology Data Exchange (ETDEWEB)

    Leshukov, A. E-mail: leshu@entek.ru; Kovalenko, V.; Shatalov, G.; Goroshkin, G.; Obukhov, A

    2000-11-01

    The design of RF DEMO-relevant ceramic helium cooled blanket test module (CHC BTM) for testing in international thermonuclear experimental reactor (ITER) is under consideration. The RF concept of DEMO BTM is based upon the breeder inside tube (BIT)-concept. This concept suggests the use of solid breeding ceramic material, helium as coolant and tritium purge-gas, ferrite-martensite steel as structural material, and beryllium as neutron multiplier. The parameters of the primary circuit coolant are the following, pressure -8 MPa, inlet/outlet temperature -300/550 deg. C, respectively. Helium (0.1 MPa pressure) is used for tritium removal from ceramic breeder. The ITER water coolant is the secondary circuit coolant of DEMO BTM cooling system. Lithium orthosilicate (Li{sub 4}SiO{sub 4}) is used as tritium breeding material (pebbles-bed of diameter 0.5-1 mm spheres). It is planned to use the beryllium as neutron multiplier (spheres diameter 1 mm pebbles-bed or the porous beryllium). The 3-D neutronic calculations on Monte Carlo method, in accordance with FENDL-1 library of the nuclear data, have been performed for CHC BTM. To validate the CHC BTM concept, the thermal hydraulic analysis has been performed for the design elements and cooling system equipment. The preliminary stress analysis for BTM design elements has been carried out on the ASME-code and RF strength regulations. The four types of LOFA and LOCA accidents have been investigated. The parameters of cooling, coolant purification and tritium extraction systems have been determined.

  11. Fast breeder reactor blanket management: comparison of LMFBR and GCFR blankets

    International Nuclear Information System (INIS)

    The economic performance of the fast breeder reactor blanket, considering different fuel management schemes was studied. To perform this, the investigation started with a standard reactor physics calculation. Then, two economic models for evaluation of the economic performance of the radial blanket were developed. These models formed the basis of a computer code, ECOBLAN, which computes the net economic gain and the levelized fuel cost due to the radial blanket. The net gain in terms of dollars and $/kgHM-y and the levelized fuel cost in mills/kWhe were obtained as a function of blanket thickness and a residence time of the fuel in the blanket. A LMFBR and a GCFR were the reactor models considered in this study. The optimum radial blanket of a GCFR consists of two rows, that of a LMFBR consists of three rows. Regarding the different fuel management schemes, the fixed blanket was found to be more favorable than reshuffled blanket. Out-in and in-out reshuffled blanket offer almost the same net gain. In all the cases, the burnup calculated for the fuel was found to be less than the acceptable limit. There is an optimum residence time for the fuel in the blanket which depends on the position of the fuel in the blanket and the fuel management scheme studied. As expected, except for very short residence times (less than 2.5 years), the radial blanket is a net income producer. There is no significant difference between the economic performance of the blanket of a LMFBR and a GCFR

  12. First adaptation of the European ceramic B. I. T. blanket design to the updated DEMO specifications

    Energy Technology Data Exchange (ETDEWEB)

    Anzidei, L.; Cecchi, P.; Cevolani, S.; Gallina, M.; Petrizzi, L.; Rado, V.; Talarico, C.; Violante, V.; Vettraino, V.; Zampaglione, V. (Associazione Euratom-ENEA sulla Fusione, Frascati (Italy)); Proust, E.; Giancarli, L.; Raepsaet, X.; Szczepanski, J.; Vallette, F.; Baraer, L.; Bielak, B.; Mercier, J. (Commissariat a l' Energie Atomique, DRN/DMT/SERMA, C.E.N. Saclay, 91 - Gif-sur-Yvette (France))

    1991-12-01

    The DEMO specifications defined so as to ensure the consistency of the various blanket conceptual design studies performed within the framework of the European Test Blanket Programme have been recently updated. A very first attempt has been made to adapt the European Ceramic Breeder Inside-Tube DEMO blanket to these new specifications. Two solutions have been investigated. The first would ensure tritium self-sufficiency of the plant with a large safety margin. The other one, which fully preserves the design simplicity and reliability of the initial design, appears to be somewhat marginal from the tritium breeding capability point of view, but to offer good improvement prospects. (orig.).

  13. New progress on design and R and D for solid breeder test blanket module in China

    Energy Technology Data Exchange (ETDEWEB)

    Feng, K.M., E-mail: fengkm@swip.ac.cn; Zhang, G.S.; Hu, G.; Chen, Y.J.; Feng, Y.J.; Li, Z.X.; Wang, P.H.; Zhao, Z.; Ye, X.F.; Xiang, B.; Zhang, L.; Wang, Q.J.; Cao, Q.X.; Zhao, F.C.; Wang, F.; Liu, Y.; Zhang, M.C.

    2014-10-15

    Highlights: • The new progress on design and R and D of Chinese solid breeder TBM are introduced. • The mock-up fabrication and component tests for Chinese HCCB TBM have being developed. • The neutron multiplier Be pebbles, tritium breeder Li{sub 4}SiO{sub 4} pebbles, and structure material CFL-1 are being prepared. • The fabrication of 1/3 sized mock-up is being carried-out. • The key technology development is proceeding to the large-scale mock-up fabrication. - Abstract: ITER will be used to test tritium breeding module concepts, which will lead to the design of DEMO fusion reactor demonstrating tritium self-sufficiency and the extraction of high grade heat for electricity production. China plans to test the HCCB TBM modules during different operation phases. Related design and R and D activities for each TBM module with the auxiliary system are introduced. The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. The preliminary conceptual design of CN HCCB TBM has been completed. A modified design to reduce the RAFM material mass to 1.3 ton has been carried out based on the ITER technical requirement. Basic characteristics and main design parameters of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being developed. Recent status of the components of CN HCCB TBM and fabrication technology development are also reported. The neutron multiplier Be pebbles, tritium breeder Li{sub 4}SiO{sub 4} pebbles, and structure material CLF-1 of ton-class are being prepared in laboratory scale. The fabrication of pebble bed container and experiment of tritium breeder pebble bed will be started soon. The fabrication technology development is proceeding as the large-scale mock-up fabrication enters into the R and D stage and demonstration tests toward TBM testing on ITER test port are being done as scheduled.

  14. Design and safety analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shuai; Zhou, Guangming; Lv, Zhongliang; Jin, Cheng; Chen, Hongli [University of Science and Technology of China, Anhui (China). School of Nuclear Science and Technology

    2016-05-15

    This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

  15. Design and safety analysis of the helium cooled solid breeder blanket for CFETR

    International Nuclear Information System (INIS)

    This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

  16. Influence of start up and pulsed operation on tritium release and inventory of NET ceramic blanket

    International Nuclear Information System (INIS)

    A first estimate for the tritium release behaviour of a ceramic breeder blanket in pulsed operation is obtained by assuming a linear steady state temperature distribution and taking into account the time constant of the thermal behaviour. The release behaviour of the breeder exposed to consecutive periods of tritium generation is described with an analytical solution of the diffusion equation. The results are compared with a simple exponential approach valid for surfacte desorption controlled release. The exponential model is used to simulate a blanket with aluminate as breeder material, which takes longest to reach steady state. The simulation demonstrates that a significant fraction (>67%) of steady state can be achieved after a testing time of about one day. (author). 7 refs.; 8 figs.; 3 tabs

  17. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 2: BOT helium cooled solid breeder blanket. Vol. 2

    International Nuclear Information System (INIS)

    The BOT (Breeder Outside Tube) Helium Cooled Solid Breeder Blanket for a fusion Demo reactor and the status of the R and D program is presented. This is the KfK contribution to the European Program for the Demo relevant test blankets to be irradiated in NET/ITER. Volume 1 (KfK 4928) contains the summary, volume 2 (KfK 4929) a more detailed version of the report. In both volumes are described the reasons for the selected design, the reference blanket design for the Demo reactor, the design of the test blanket including the ancillary systems together with the present status of the relative R and D program in the fields of neutronic and thermohydraulic calculations, of the electromagnetic forces caused by disruptions, of the development and irradiation of the ceramic breeder material, of the tritium release and recovery, and of the technological investigations. An outlook is given on the required R and D program for the BOT Helium Cooled Solid Breeder Blanket prior to tests in NET/ITER and the proposed test program in NET/ITER. (orig.)

  18. A Feasible DEMO Blanket Concept Based on Water Cooled Solid Breeder

    International Nuclear Information System (INIS)

    Full text: JAEA has conducted the conceptual design study of blanket for a fusion DEMO reactor SlimCS. Considering DEMO specific requirements, we place emphasis on a blanket concept with durability to severe irradiation, ease of fabrication for mass production, operation temperature of blanket materials, and maintainability using remote handling equipment. This paper present a promising concept satisfying these requirements, which is characterized by minimized welding lines near the front, a simplified blanket interior consisting of cooling tubes and a mixed pebble bed of breeder and neutron multiplier, and approximately the same outlet temperature for all blanket modules. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in tritium breeding ratio even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production, which will facilitate the access of remote handling equipment for replacement of the blanket modules and improve the access of diagnostics. (author)

  19. Preliminary neutronics design and analysis of helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Zhongliang; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Chen, Chong; Li, Min; Zhou, Guangming

    2015-06-15

    Highlights: • Neutronics design of a helium cooled solid breeder blanket for CFETR was presented. • The breeding zones parallel to FW and perpendicular to FW were optimized. • A series of neutronics analyses for the proposed blanket were shown. - Abstract: Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. Tritium self-sufficiency is one of the most important issues for CFETR and the tritium breeding ratio (TBR) is recommended not less than 1.2. As one of the candidates, a helium cooled solid breeder blanket for CFETR superconducting tokamak option was proposed. In the concept, radial arranged U-shaped breeding zones are adopted for higher TBR and simpler structure. In this work, three-dimensional neutronics design and analysis of the blanket were performed using the Monte Carlo N-Particle transport code MCNP with IAEA data library FENDL-2.1. Tritium breeding capability of the proposed blanket was assessed and the breeding zones parallel to first wall (FW) and perpendicular to FW were optimized. Meanwhile, the nuclear heating analysis and shielding performance were also presented for later thermal and structural analysis. The results showed that the blanket could well meet the tritium self-sufficiency target and the neutron shield could satisfy the design requirements.

  20. The gas-cooled Li2O moderator/breeder canister blanket for fusion-synfuels

    International Nuclear Information System (INIS)

    A new integrated power and breeding blanket is described. The blanket incorporates features that make it suitable for synthetic fuel production. It is matched to the thermal and electrical requirements of the General Atomic water-splitting process for producing hydrogen. The fusion reaction is the Tandem Mirror Reactor (TMR) using Mirror Advanced Reactor Study (MARS) physics. The canister blanket is a high temperature, pressure balanced, crossflow heat exchanger contained within a low activity, independently cooled, moderate temperature, first wall structural envelope. The canister uses Li2O as the moderator/breeder and helium as the coolant. ''In situ'' tritium control, combined with slip stream processing and self-healing permeation barriers, assures a hydrogen product essentially free of tritium. The blanket is particularly adapted to synfuels production but is equally useful for electricity production or co-generation

  1. Fast Breeder Blanket Facility FBBF. Annual report, January 1, 1981-December 31, 1981

    International Nuclear Information System (INIS)

    This annual report contains a summmary of fission rate, spectra, and gamma-ray heating rate measurements made in the first blanket of the Purdue Fast Breeder Blanket Facility. The first blanket consisted of aluminum clad, natural UO2 fuel rods with a secondary cladding of stainless steel or aluminum. The blanket was arranged in two concentric regions around the neutron source and converter regions. A neutron diffusion code, 2DB, and a Monte Carlo code, VIM, both using homogeneous cross section groups have been used to calculate the reaction rates. Calculated to experimental values for a number of important reactions are presented. A modified method of applying Bondarenko self-shielding factors to correct for the self shielding of resonance energy neutrons in aluminum, stainless steel and UO2 has improved the agreement between the calculations and experiment, but does not account for all of the differences

  2. R and D activities of the liquid breeder blanket in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won, E-mail: dwlee@kaeri.re.kr [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Eo Hwak; Kim, Suk Kwon; Yoon, Jae Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer MARS and GAMMA were developed for He coolant and liquid breeder analysis. Black-Right-Pointing-Pointer FMS/FMS and Be/FMS joining methods were developed and verified with high heat flux test. Black-Right-Pointing-Pointer High temperature and pressure nitrogen and He loops were constructed for heat transfer experiment for developed codes validation. Black-Right-Pointing-Pointer A PbLi breeder loop was constructed for components, MHD, and corrosion tests. Black-Right-Pointing-Pointer A chamber for tritium extraction with a gas-liquid contact method was constructed. - Abstract: A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the common fields of a solid TBM such as design tools, structural material, fabrication methods, and He cooling technology to support this concept for the ITER. Also, other fields such as a liquid breeder technology and tritium extraction have been developed from the designed liquid TBM. For design tools, system codes for safety analysis such as Multi-dimensional Analysis of Reactor Safety (MARS) and GAs Multi-component Mixture Analysis (GAMMA) were developed for He coolant and liquid breeder. For the fabrication methods, Ferritic Martensitic Steel (FMS) to FMS and Be to FMS joinings with a Hot Isostatic Pressing (HIP) were developed and verified with a high heat flux test of up to 0.5-1.0 MW/m{sup 2}. Moreover, three mockups were successfully fabricated and a 10-channel prototype is being fabricated to make a rectangular channel FW. For the integrity of the joining, two high heat flux test facilities were constructed, and one using an electron beam has been constructed. With the 6 MPa nitrogen loop, a basic heat transfer experiment for code validation was performed. From the verification of the components such as preheater and

  3. Extraction of tritium from ceramic breeder material

    International Nuclear Information System (INIS)

    The first generation of fusion reactors will use deuterium and tritium as fuel since this reaction takes place at relatively low temperature. Since tritium is not available in nature, it must be produced in the fusion reactor blanket which surrounds the plasma zone. The lithium bearing compound is available in plenty in earths crust and by absorbing neutron, lithium produces tritium by the reactions 6Li (n, α) T and 7Li (n, n'α) T. Natural lithium consists of 93% 7Li and the remaining 7% as 6Li. Since the inelastic scattering of 7Li with fast neutrons produces one tritium and one neutron, more than one tritium atom can be produced per neutron. Hence by suitably designing the lithium blanket, more than one tritium atom per fusion reaction can be produced. In the absence of thermonuclear reactions, the (D,T) neutrons which are energetic 14-MeV neutrons, are produced in the accelerator based neutron generators. In order to ensure that sufficient amount of tritium would be produced in the future fusion reactor blankets, experiments are carried out to irradiate the lithium assembly using the available neutron source and measurements are done to estimate the tritium breeding. Also, it is required to extract the tritium produced in the lithium blanket. This work consists of tritium breeding measurement technique and a design of tritium extraction system. (author)

  4. Isotope exchange reactions on ceramic breeder materials and their effect on tritium inventory

    Energy Technology Data Exchange (ETDEWEB)

    Nishikawa, M.; Baba, A. [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering; Kawamura, Y.; Nishi, M.

    1998-03-01

    Though lithium ceramic materials such as Li{sub 2}O, LiAlO{sub 2}, Li{sub 2}ZrO{sub 3}, Li{sub 2}TiO{sub 3} and Li{sub 4}SiO{sub 4} are considered as breeding materials in the blanket of a D-T fusion reactor, the release behavior of the bred tritium in these solid breeder materials has not been fully understood. The isotope exchange reaction rate between hydrogen isotopes in the purge gas and tritium on the surface of breeding materials have not been quantified yet, although helium gas with hydrogen or deuterium is planned to be used as the blanket purge gas in the recent blanket designs. The mass transfer coefficient representing the isotope exchange reaction between H{sub 2} and D{sub 2}O or that between D{sub 2} and H{sub 2}O in the ceramic breeding materials bed is experimentally obtained in this study. Effects of isotope exchange reactions on the tritium inventory in the bleeding blanket is discussed based on data obtained in this study where effects of diffusion of tritium in the grain, absorption of water in the bulk of grain, and adsorption of water on the surface of grain, together with two types of isotope exchange reactions are considered. The way to estimate the tritium inventory in a Li{sub 2}ZrO{sub 3} blanket used in this study shows a good agreement with data obtained in such in-situ experiments as MOZART, EXOTIC-5, 6 and TRINE experiments. (author)

  5. Characterization of the effects of continuous salt processing on the performance of molten salt fusion breeder blankets

    International Nuclear Information System (INIS)

    Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs. The effects of processing on blanket performance have been assessed for three levels of processing and various equilibrium uranium concentrations in the salt. A one-dimensional model of the blanket was used in the neutronics analysis which incorporated transport calculations with time-dependent isotope generation and depletion calculations. The level of salt processing was found to have little effect on the behavior of the blanket during reactor operation; however, significant effects were observed during the decay period after reactor shutdown

  6. Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

    Science.gov (United States)

    Li, Jia; Zhang, Xiaokang; Gao, Fangfang; Pu, Yong

    2016-02-01

    China Fusion Engineering Test Reactor (CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2 to ensure tritium self-sufficiency. A concept design for a water cooled ceramics breeding blanket (WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR. Based on this concept, a one-dimensional (1D) radial built breeding blanket was first designed, and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build. A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models, addressing neutron wall loading (NWL), tritium breeding ratio (TBR), fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components. The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  7. Thermohydraulics design and thermomechanics analysis of two European breeder blanket concepts for DEMO. Pt. 1 and Pt. 2. Pt. 1: BOT helium cooled solid breeding blanket. Pt. 2: Dual coolant self-cooled liquid metal blanket

    International Nuclear Information System (INIS)

    Two different breeding blanket concepts are being elaborated at Forschungszentrum Karlsruhe within the framework of the DEMO breeding blanket development, the concept of a helium cooled solid breeding blanket and the concept of a self-cooled liquid metal blanket. The breeder material used in the first concept is Li4SiO4 as a pebble bed arranged separate from the beryllium pebble bed, which serves as multiplier. The breeder material zone is cooled by several toroidally-radially configurated helium cooling plates which, at the same time, act as reinforcements of the blanket structures. In the liquid metal blanket concept lead-lithium is used both as the breeder material and the coolant. It flows at low velocity in poloidal direction downwards and back in the blanket front zone. In both concepts the First Wall is cooled by helium gas. This report deals with the thermohydraulics design and thermomechanics analysis of the two blanket concepts. The performance data derived from the Monte-Carlo computations serve as a basis for the design calculations. The coolant inlet and outlet temperatures are chosen with the design criteria and the economics aspects taken into account. Uniform temperature distribution in the blanket structures can be achieved by suitable branching and routing of the coolant flows which contributes to reducing decisively the thermal stress. The computations were made using the ABAQUS computer code. The results obtained of the stresses have been evaluated using the ASME code. It can be demonstrated that all maximum values of temperature and stress are below the admissible limit. (orig.)

  8. Modeling of tritium behavior in ceramic breeder materials

    International Nuclear Information System (INIS)

    Computer models are being developed to predict tritium release from candidate ceramic breeder materials for fusion reactors. Early models regarded the complex process of tritium release as being rate limited by a single slow step, usually taken to be tritium diffusion. These models were unable to explain much of the experimental data. We have developed a more comprehensive model which considers diffusion and desorption from the grain surface. In developing this model we found that it was necessary to include the details of the surface phenomena in order to explain the results from recent tritium release experiments. A diffusion-desorption model with a desorption activation energy which is dependent on the surface coverage was developed. This model provided excellent agreement with the results from the CRITIC tritium release experiment. Since evidence suggests that other ceramic breeder materials have desorption activation energies which are dependent on surface coverage, it is important that these variations in activation energy be included in a model for tritium release. 17 refs., 12 figs

  9. Preliminary Study on Melting and Reaction with Liquid Metal Breeders for Developing the Korean Test Blanket Module in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D. W.; Yoon, J. S.; Kim, S. K.; Lee, E. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, H. G. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the liquid TBM. In the Korean liquid TBM and breeder blanket, liquid lithium (Li) and lead-lithium (PbLi) are considered as breeders. Related research has been performed: an Experimental Loop for a Liquid breeder (ELLI) constructed to develop an electromagnetic (EM) pump for circulating the liquid breeder, a magnetohydrodynamic (MHD) experiment, and a flow corrosion test. In the ELLI, Pb-15.7Li, where Li is 15.7 at % (called PbLi hereafter), is used as the breeding material. It was purchased from Stachow Metall Company, Germany, and its impurities are shown in Table 1. An EM pump circulates the material in the loop with a maximum flow rate of 60 lpm. The operating pressure and temperature in the loop are 0.4 MPa and 300 .deg. C, respectively, and the maximum operating pressure and temperature are 0.5 MPa and 550 .deg. C Before the loop operation, the melting and solidifying temperatures of the PbLi were measured for ascertaining whether it will show a consistent value for the many cycles of heating and cooling at various conditions of the loop operation. We can also investigate the contamination of PbLi according to the cyclic use. Of the liquid type breeder materials, PbLi is much safer than Li itself, as liquid metal can be ignited when it meets with water or air. There is still a concern regarding the use of PbLi, and it has not been fully proven whether it will react with water or air when it is in a molten state, as it contains lithium. Therefore, reaction tests of Li and PbLi with air and water were performed for safety reasons using the prepared test chamber

  10. Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

    International Nuclear Information System (INIS)

    The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

  11. Progress in studies of Li/sub 17/Pb/sub 83/ as liquid breeder for fusion reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.

    1983-09-01

    A review of the experimental and conceptual design work in progress at JRC-Ispra to investigate the feasibility of the eutectic Li/sub 17/Pb/sub 83/ as a liquid breeder for experimental power reactors is presented. Results of recent measurements to implement the data base of this material are given in the following areas: physical parameters, hydrogen solubility and recovery, chemical reactivity with air and water, compatibility with steel. The studies carried out on blanket concepts for the INTOR (International Tokamak Reactor)/NET (Next European Torus) projects are outlined and discussed.

  12. Development and analysis of fusion breeder blanket neutronics. Progress report, November 1, 1983-October 31, 1984

    International Nuclear Information System (INIS)

    The following activities are briefly described: (a) the IBM versions of the computer codes FORSS, PUFF-II, ONETRAN, TWOTRAN-II, and DOT4.3 were obtained from the Radiation Shielding Information Center (RSIC) and have been implemented on the UCLA local computer, the IBM 3033; (b) mathematical and computational models to describe the time-dependent transport and inventory of tritium in individual components of a fusion reactor system have been developed; (c) extensive cross-section sensitivity and uncertainty analysis was carried out to evaluate an estimate for the uncertainty associated with the TBR (both from 6Li and 7Li, individually) in four of the leading blanket concepts (the Li2O/HT-9 helium-cooled blanket, the 17Li-83Pb/PCA self-cooled blanket, the LiAlO2/He/FS/Be blanket, and the flibe/He/FS/Be blanket); (d) as far as the TBR obtain able in various blanket concepts is concerned, a comparative analysis was carried out to estimate the change in TBR in a particular blanket module when placed in a tokamak machine [R (first wall) approx. 2 m] as opposed to adopting the same blanket in a mirror machine [R (first wall) approx. 50 cm] with the same wall loading

  13. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 1: Self-cooled liquid metal breeder blanket. Vol. 2. Detailed version

    International Nuclear Information System (INIS)

    A self-cooled liquid metal breeder blanket for a fusion DEMO-reactor and the status of the development programme is described as a part of the European development programme of DEMO relevant test blankets for NET/ITER. Volume 1 (KfK 4907) contains a summary. Volume 2 (KfK 4908) a more detailed version of the report. Both volumes contain sections on previous studies on self-cooled liquid metal breeder blankets, the reference blanket design for a DEMO-reactor, a typical test blanket design including the ancillary loop system and the building requirements for NET/ITER together with the present status of the associated RandD-programme in the fields of neutronics, magnetohydrodynamics, tritium removal and recovery, liquid metal compatibility and purification, ancillary loop system, safety and reliability. An outlook is given regarding the required RandD-programme for the self-cooled liquid metal breeder blanket prior to tests in NET/ITER and the relevant test programme to be performed in NET/ITER. (orig.)

  14. Fabrication, properties, and tritium recovery from solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, C.E. (Argonne National Lab., IL (USA)); Kondo, T. (Japan Atomic Energy Research Inst., Tokyo (Japan)); Roux, N. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)); Tanaka, S. (Tokyo Univ. (Japan)); Vollath, D. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.))

    1991-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 133 refs., 1 fig.

  15. Breeding zone models of DEMO ceramic helium cooled blanket test module for testing in IVV-2M reactor

    International Nuclear Information System (INIS)

    The goal of DEMO ceramic helium cooled blanket test module (CHC BTM) is to demonstrate a breeding capability that would lead to tritium self-sufficiency in ITER reactor and to extract a high-grade heat suitable for electricity generation. Experimental validation of all the adopted design solutions is main important problem at design and calculation works carrying out in order to develop the CHC BTM. One important task for breeding zones feasibility validation is in-pile tests. Two models were developed and fabricated for testing in the fission IVV-2M reactor. Breeding zone is based on poloidal BIT-conception. The models structural material is ferrito-martensitic steel. Breeder material is lithium orthosilicate in pebble beds and pellet forms. Multiplier material is beryllium in pebble beds and porosity forms. The cooling is provided by helium at 10 MPa. The tritium produced in the breeder material is purged by the helium flow at 0.1-0.2 MPa. Designs of model description and experimental channel, results of neutronic and thermo-hydraulic calculations are presented in the paper. (orig.)

  16. An analysis of electron beam welds in a dual coolant liquid metal breeder blanket

    International Nuclear Information System (INIS)

    Numerical simulation of electron beam welding of blanket segments was performed using non-linear finite element code ABAQUS. The thermal and stress fields were assumed uncoupled, while preserving the temperature dependency of all material parameters. The martensite-austenite and austenite-martensite transformations were taken into account through volume shrinking/expansion effects, which is consistent with available data. The distributions of post welding residual stress in a complex geometry of the first wall are obtained. Also, the effects of preheating and post-welding heat treatment were addressed. Time dependent temperature and stress-strain fields obtained provide good insight into the welding process. They may be used directly to support reliability and life-time studies of blanket structures. On the other hand, they provide useful hints about the feasibility of the geometrical configurations as proposed by different design concepts. (orig.)

  17. Nitriding treatment of reduced activation ferritic steel as functional layer for liquid breeder blanket

    International Nuclear Information System (INIS)

    The development of functional layers such as a tritium permeation barrier and an anti-corrosion layer is the essential technology for the development of a molten salt type self cooled fusion blanket. In the present study, the characteristics of a nitriding treatment on a reduced activation ferritic steel, JLF-1 (Fe-9Cr-2W-0.1C) as the functional layer were investigated. The steel surface was nitrided by an ion nitriding treatment or a radical nitriding treatment. The nitridation characteristic of the steel surface was made clear based on the thermodynamic stability. The thermal diffusivity, the hydrogen permeability and the chemical stability in the molten salt Flinak were investigated. The results indicated that the nitriding treatment can improve the compatibility in the Flinak without the decrease of the thermal diffusivity, though there was little improvement as the hydrogen permeation barrier. (author)

  18. Achievements of the water cooled solid breeder test blanket module of Japan to the milestones for installation in ITER

    International Nuclear Information System (INIS)

    As the primary candidate of ITER Test Blanket Module (TBM) to be tested under the leadership of Japan, Water Cooled Solid Breeder (WCSB) TBM is being developed. Six TBMs will be tested in ITER simultaneously, under the leadership of different countries. To ensure the installation of reliable TBMs, it is necessary to show feasibility on the TBM milestones for installation in ITER. This paper shows the recent achievements toward the milestones of ITER TBMs prior to the installation, that consist of design integration in ITER, module qualification and safety assessment. With respect to the design integration, it is necessary to show the consistency with ITER design on time with ITER design progress, targeting the detailed design final report in 2012. Structure design of the interfacing components between the WCSB TBM structure and the interfacing components (Common Frame and Backside Shielding) that are placed in a test port of ITER has been developed. The design work also consists of procedures of fabrication and replacement of TBM, the consistency with ITER port structure and TBM interface structure, and the layouts of the auxiliary systems of TBMs including the tritium extraction system and water cooling system. As for the module qualification, it is necessary to show fabrication capability and the integrity of prototypical size mockup in corresponding operation condition before the delivery of the TBM to ITER. A real scale first wall mock-up was successfully fabricated by using Hot Isostatic Pressing (HIP) method by structural material of reduced activation martensitic ferritic steel, F82H. High heat flux test with real cooling water condition is planned using this mock-up. Other essential R and Ds for the WCSB TBM also showed steady progress on investigation of mechanical behavior of breeder pebble beds, development of advanced breeder/multiplier pebble, neutron measurement technology for TBM and purge gas tritium recovery technology. As for safety milestones

  19. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    International Nuclear Information System (INIS)

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m2 and a surface heat flux of 1 MW/m2. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO2 rods. The helium coolant pressure is 5 MPa, entering the module at 2970C and exiting at 5500C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter

  20. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A. (eds.)

    1986-09-01

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m/sup 2/ and a surface heat flux of 1 MW/m/sup 2/. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO/sub 2/ rods. The helium coolant pressure is 5 MPa, entering the module at 297/sup 0/C and exiting at 550/sup 0/C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter.

  1. Exchange reaction of hydrogen isotopes on proton conductor ceramic of hydrogen pump for blanket tritium recovery system

    International Nuclear Information System (INIS)

    Electrochemical hydrogen pump using ceramic proton conductor has been investigated to discuss its application for the blanket tritium recovery system of the nuclear fusion reactor. As the series of those work, the transportation experiments of H2-D2 mixture via ceramic proton conductor membrane have been carried out. Then, the phenomenon that was caused by the exchange reaction between the deuterium in the ceramic and the hydrogen in the gas phase has been observed. So, the ceramic proton conductor which doped deuterium was exposed to hydrogen under the control of zero current, and the effluent gas was analyzed. It is considered that the hydrogen in the gas phase is taken as proton to the ceramic by isotope exchange reaction, and penetrates to the ceramic by diffusion with replacement of deuteron. (author)

  2. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    This volume contains extensive data for the following chapters: (1) solid breeder tritium recovery, (2) solid breeder blanket designs, (3) alternate blanket concept screening, and (4) safety analysis. The following appendices are also included: (1) blanket design guidelines, (2) power conversion systems, (3) helium-cooled, vanadium alloy structure blanket design, (4) high wall loading study, and (5) molten salt safety studies

  3. Blanket comparison and selection study. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    1983-10-01

    This volume contains extensive data for the following chapters: (1) solid breeder tritium recovery, (2) solid breeder blanket designs, (3) alternate blanket concept screening, and (4) safety analysis. The following appendices are also included: (1) blanket design guidelines, (2) power conversion systems, (3) helium-cooled, vanadium alloy structure blanket design, (4) high wall loading study, and (5) molten salt safety studies. (MOW)

  4. Fusion breeder

    International Nuclear Information System (INIS)

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the US fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the US fusion program and the US nuclear energy program. The purpose of this paper is to suggest this policy change be made and tell why it should be made, and to outline specific research and development goals so that the fusion breeder will be developed in time to meet fissile fuel needs

  5. Beryllium data base for in-pile mockup test on blanket of fusion reactor, (1)

    International Nuclear Information System (INIS)

    Beryllium has been used in the fusion blanket designs with ceramic breeder as a neutron multiplier to increase the net tritium breeding ratio (TBR). The properties of beryllium, that is physical properties, chemical properties, thermal properties, mechanical properties, nuclear properties, radiation effects, etc. are necessary for the fusion blanket design. However, the properties of beryllium have not been arranged for the fusion blanket design. Therefore, it is indispensable to check and examine the material data of beryllium reported previously. This paper is the first one of the series of papers on beryllium data base, which summarizes the reported material data of beryllium. (author)

  6. Thermal and structural design issues of breeding blankets for testing in the Next European Torus (NET)

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.

    1988-05-01

    A review of the breeding blankets under study in Europe for testing in the Next European Torus is presented. In many concepts, the breeder modules are enclosed in boxes whose side walls in front of the plasma act as the first wall of the machine. Various types of breeder modules are investigated, involving both liquid and solid breeders, namely: - Pb-17Li liquid breeder concepts, the coolant being either water or Pb-17Li itself; - solid (ceramic) breeder concepts, the coolant being in all cases helium. The various ceramic concepts differ in the breeder/coolant arrangement (breeder-out-of-tube and breeder-in-tube), the orientation of the coolant tubes (poloidal or toroidal) and the breeder geometry (rods, plates or pebble bed). For each of these concepts the main design features are shown and the thermomechanical problems are discussed. The problems related to a coolant tube rupture are in many cases the most severe from the structural design point of view. The first wall box enclosing the breeder modules appears to be a weak secondary containment barrier. The liquid breeder-water cooled concept looks manageable from the thermal and structural design of point view. In the case of the self-cooled liquid breeder concept, the main problems are related to the magnetohydrodynamic effects. Solutions are envisaged to overcome these difficulties. In the case of ceramic breeders, the use of plates implies small dimensions in order to limit the thermal stresses and a poor exploitation of the permitted temperature operation window. Solutions involving rods associated with a multipass cooling scheme or pebble bed enable achievement of better thermomechanical conditions and, therefore, are preferred in the current investigations. However, they lead to design complications and require experimental verification which is in progress at the European laboratories.

  7. ITER breeding blanket module design and analysis

    International Nuclear Information System (INIS)

    The ITER breeding blanket employs a ceramic breeder and Be neutron multiplier both in small spherical pebble form. Radial-poloidal cooling panels are arranged in the blanket box to remove the nuclear heating in these materials and to reinforce the blanket structure. At the first wall, Be armor is bonded onto the stainless steel (SS) structure to provide a low Z plasma-compatible surface and to protect the first wall/blanket structure from the direct contact with the plasma during off-normal events. Thermo-mechanical analyses and investigation of fabrication procedure have been performed for this breeding blanket. To evaluate thermo-mechanical behavior of the pebble beds including the dependency of the effective thermal conductivity on stress, analysis methods have been preliminary established by the use of special calculation option of ABAQUS code, which are briefly summarized in this report. The structural response of the breeding blanket module under internal pressure of 4 MPa (in case of in-blanket LOCA) resulted in rather high stress in the blanket side (toroidal end) wall, thus addition of a stiffening rib or increase of the wall thickness will be needed. Two-dimensional elasto-plastic analyses have been performed for the Be/SS bonded interface at the first wall taking a fabrication process based on HIP bonding and thermal cycle due to pulsed plasma operation into account. The stress-strain hysteresis during these process and operation was clarified, and a procedure to assess and/or confirm the bonding integrity was also proposed. Fabrication sequence of the breeding blanket module was preliminarily developed based on the procedure to fabricate part by part and to assemble them one by one. (author)

  8. The fusion breeder

    International Nuclear Information System (INIS)

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the U.S. fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the U.S. fusion program and the U.S. nuclear energy program. There is wide agreement that many approaches will work and will produce fuel for five equal-sized LWRs, and some approach as many as 20 LWRs at electricity costs within 20% of those at today's price of uranium ($30/lb of U3O8). The blankets designed to suppress fissioning, called symbiotes, fusion fuel factories, or just fusion breeders, will have safety characteristics more like pure fusion reactors and will support as many as 15 equal power LWRs. The blankets designed to maximize fast fission of fertile material will have safety characteristics more like fission reactors and will support 5 LWRs. This author strongly recommends development of the fission suppressed blanket type, a point of view not agreed upon by everyone. There is, however, wide agreement that, to meet the market price for uranium which would result in LWR electricity within 20% of today's cost with either blanket type, fusion components can cost severalfold more than would be allowed for pure fusion to meet the goal of making electricity alone at 20% over today's fission costs. Also widely agreed is that the critical-pathitem for the fusion breeder is fusion development itself; however, development of fusion breeder specific items (blankets, fuel cycle) should be started now in order to have the fusion breeder by the time the rise in uranium prices forces other more costly choices

  9. Crucial issues on liquid metal blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Malang, S. (Kernforschungszentrum Karlsruhe (Germany)); Leroy, P. (CEA, CEN Saclay, 91 - Gif-sur-Yvette (France)); Casini, G.P. (CEC, Joint Research Centre (JRC), Ispra (Italy)); Mattas, R.F. (Argonne National Lab., IL (United States)); Strebkov, Yu. (Research and Development Inst. of Power Engineering, Moscow (USSR))

    1991-12-01

    Typical design concepts of liquid metal breeder blankets for power reactors are explained and characterized. The major problems of these concepts are described for both water-cooled blankets and self-cooled blankets. Three crucial issues of liquid metal breeder blankets are investigated. They are in the fields of magnetohydrodynamics, tritium control and safety. The influence of the magnetic field on liquid metal flow is of special interest for self-cooled blankets. The main problems in this field and the status of the related R and D work are described. Tritium permeation losses to the cooling water is a crucial issue for water-cooled blankets. Methods for its reduction are discussed. An inherent problem of all liquid breeder blankets is the potential release of activated products in the case of chemical reactions between the breeder material and water or reactive gases. The most important issues in this field are described. (orig.).

  10. Crystal chemistry of immobilization of fast breeder reactor (FBR) simulated waste in sodium zirconium phosphate (NZP) ceramic matrix

    Energy Technology Data Exchange (ETDEWEB)

    Chourasia, Rashmi [Department of Chemistry, Dr. H.S. Gour University, Sagar 470 003 (India); Shrivastava, O.P., E-mail: dr_ops11@rediffmail.co [Department of Chemistry, Dr. H.S. Gour University, Sagar 470 003 (India); Ambashta, R.D.; Wattal, P.K. [Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2010-02-15

    Fuel from the fast breeder reactor waste is reprocessed and subjected to cooling for a period of about one year. Fission and activation products of the fuel are the major constituents of this waste. Sodium zirconium phosphate (hereafter NZP) has been identified as a potential material for immobilization of long lived heat generating radio nuclides. It was found that most of the elements present in the radioactive waste could be immobilized in this ceramic matrix without significant changes of the three-dimensional framework of the host material. Simulated NZP waste forms synthesized by ceramic route at 1200 deg. C crystallize in the rhombohedral system (space group R-3c). The crystal chemistry of 0-35 wt.% waste loaded NZP waste forms have been investigated using General Structure Analysis System (GSAS) programming of the step analysis powder diffraction data. Rietveld refinement of crystal data on the waste oxide (WO{sub x}) loaded waste forms gives a satisfactory convergence of R-factors. The particle size along prominent reflecting planes ranges between 68 and 141 nm. The polyhedral distortions and effective valence calculations from bond strength data are also reported. Morphological examination by scanning electron microscopy (SEM) reveals that the size of almost rectangular parallelepiped shaped grains varies between 0.2 and 5 mum. The EDX analysis provides analytical evidence of immobilization of effluent cations in the matrix.

  11. Fusion Breeder Program interim report

    International Nuclear Information System (INIS)

    This interim report for the FY82 Fusion Breeder Program covers work performed during the scoping phase of the study, December, 1981-February 1982. The goals for the FY82 study are the identification and development of a reference blanket concept using the fission suppression concept and the definition of a development plan to further the fusion breeder application. The context of the study is the tandem mirror reactor, but emphasis is placed upon blanket engineering. A tokamak driver and blanket concept will be selected and studied in more detail during FY83

  12. Low activity aluminum blanket

    Energy Technology Data Exchange (ETDEWEB)

    Benenati, R.; Tichler, P.; Powell, J.R.

    1976-03-01

    The basic design of the breeding blanket consists of cylindrical aluminium canisters filled with a ceramic bed of moderating, shielding, and breeding materials all suitably cooled. A technical analysis of the blanket for an EPR design is given. Activation studies are presented. The effect of pulsed magnetic fields on module structure is investigated. (MOW)

  13. Neutronics Evaluation of Lithium-Based Ternary Alloys in IFE Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Jolodosky, A. [Univ. of California, Berkeley, CA (United States); Fratoni, M. [Univ. of California, Berkeley, CA (United States)

    2015-09-22

    Lithium is often the preferred choice as breeder and coolant in fusion blankets as it offers excellent heat transfer and corrosion properties, and most importantly, it has a very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and exacerbates plant safety concerns. For this reason, over the years numerous blanket concepts have been proposed with the scope of reducing concerns associated with lithium. The European helium cooled pebble bed breeding blanket (HCPB) physically confines lithium within ceramic pebbles. The pebbles reside within a low activation martensitic ferritic steel structure and are cooled by helium. The blanket is composed of the tritium breeding lithium ceramic pebbles and neutron multiplying beryllium pebbles. Other blanket designs utilize lead to lower chemical reactivity; LiPb alone can serve as a breeder, coolant, neutron multiplier, and tritium carrier. Blankets employing LiPb coolants alongside silicon carbide structural components can achieve high plant efficiency, low afterheat, and low operation pressures. This alloy can also be used alongside of helium such as in the dual-coolant lead-lithium concept (DCLL); helium is utilized to cool the first wall and structural components made up of low-activation ferritic steel, whereas lithium-lead (LiPb) acts as a self-cooled breeder in the inner channels of the blanket. The helium-cooled steel and lead-lithium alloy are separated by flow channel inserts (usually made out of silicon carbide) which thermally insulate the self-cooled breeder region from the helium cooled steel walls. This creates a LiPb breeder with a much higher exit temperature than the steel which increases the power cycle efficiency and also lowers the magnetohydrodynamic (MHD) pressure drop [6]. Molten salt blankets with a mixture of lithium, beryllium, and fluorides (FLiBe) offer good tritium breeding

  14. An Analysis of Ripple and Error Fields Induced by a Blanket in the CFETR

    Science.gov (United States)

    Yu, Guanying; Liu, Xufeng; Liu, Songlin

    2016-10-01

    The Chinese Fusion Engineering Tokamak Reactor (CFETR) is an important intermediate device between ITER and DEMO. The Water Cooled Ceramic Breeder (WCCB) blanket whose structural material is mainly made of Reduced Activation Ferritic/Martensitic (RAFM) steel, is one of the candidate conceptual blanket design. An analysis of ripple and error field induced by RAFM steel in WCCB is evaluated with the method of static magnetic analysis in the ANSYS code. Significant additional magnetic field is produced by blanket and it leads to an increased ripple field. Maximum ripple along the separatrix line reaches 0.53% which is higher than 0.5% of the acceptable design value. Simultaneously, one blanket module is taken out for heating purpose and the resulting error field is calculated to be seriously against the requirement. supported by National Natural Science Foundation of China (No. 11175207) and the National Magnetic Confinement Fusion Program of China (No. 2013GB108004)

  15. Production behavior of irradiation defects in solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  16. Tritium-assisted fusion breeders

    International Nuclear Information System (INIS)

    This report undertakes a preliminary assessment of the prospects of tritium-assisted D-D fuel cycle fusion breeders. Two well documented fusion power reactor designs - the STARFIRE (D-T fuel cycle) and the WILDCAT (Cat-D fuel cycle) tokamaks - are converted into fusion breeders by replacing the fusion electric blankets with 233U producing fission suppressed blankets; changing the Cat-D fuel cycle mode of operation by one of the several tritium-assisted D-D-based modes of operation considered; adjusting the reactor power level; and modifying the resulting plant cost to account for the design changes. Three sources of tritium are considered for assisting the D-D fuel cycle: tritium produced in the blankets from lithium or from 3He and tritium produced in the client fission reactors. The D-D-based fusion breeders using tritium assistance are found to be the most promising economically, especially the Tritium Catalyzed Deuterium mode of operation in which the 3He exhausted from the plasma is converted, by neutron capture in the blanket, into tritium which is in turn fed back to the plasma. The number of fission reactors of equal thermal power supported by Tritium Catalyzed Deuterium fusion breeders is about 50% higher than that of D-T fusion breeders, and the profitability is found to be slightly lower than that of the D-T fusion breeders

  17. Options and methods for instrumentation of Test Blanket Systems for experiment control and scientific mission

    International Nuclear Information System (INIS)

    Highlights: • This work defined options and methods to instrument ITER TBSs based on functional categories: safety, interlock and control and scientific exploitation based on the ITER research program. • Presented the general architecture of the HCLL and HCPB Test Blanket System Instrumentation and Control. • Defined safety and interlock sensors count and technology selection based on preliminary safety analysis. • Discussed the development status of scientific instrumentation, with focus on integration with design and fulfillment of TBM research program. - Abstract: Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept which uses the eutectic Pb-16Li as both breeder and neutron multiplier; the Helium-Cooled Pebble-Bed (HCPB) concept which features lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Each TBM is associated with several sub-systems required for their operation; together they form the Test Blanket System (TBS). This paper presents the state of HCLL and HCPB TBS instrumentation design. The discussion is based on the systems functional analysis, from which three main categories of instrumentation are defined: those relevant to safety functions; those relevant to interlock functions; those designed for the control and scientific exploitation of the devices based on the TBM program objectives

  18. Use of Nuclear Data Sensitivity and Uncertainty Analysis for the Design Preparation of the HCLL Breeder Blanket Mockup Experiment for ITER

    Directory of Open Access Journals (Sweden)

    I. Kodeli

    2008-01-01

    Full Text Available An experiment on a mockup of the test blanket module based on helium-cooled lithium lead (HCLL concept will be performed in 2008 in the Frascati Neutron Generator (FNG in order to study neutronics characteristics of the module and the accuracy of the computational tools. With the objective to prepare and optimise the design of the mockup in the sense to provide maximum information on the state-of-the-art of the cross-section data the mockup was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR, their sensitivity to the underlying basic cross-sections, as well as the corresponding uncertainties were calculated using the deterministic transport codes (DOORS package, the sensitivity/uncertainty code package SUSD3D, and the VITAMINJ/ COVA covariance matrix libraries. The cross-section reactions with largest contribution to the uncertainty of the calculated TPR were identified to be (n,2n and (n,3n reactions on lead. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross-sections.

  19. Impact of blanket tritium against the tritium plant of fusion reactor

    International Nuclear Information System (INIS)

    The breeder blanket and the blanket tritium recovery system are tested using test blanket modules during ITER campaign. And then, these are integrated with the tritium plant for the first time at a prototype reactor after ITER. In this work, impact to the tritium plant by integration of the solid breeder blanket was discussed. The method of tritium extraction from the blanket and the choice of the process for breeder blanket interface should be discussed not only from the viewpoint of tritium release but also from the viewpoint of the load of processing. (author)

  20. The State of the Art Report on the Development and Manufacturing Technology of Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. S.; Jeong, Y. H.; Park, S. Y.; Lee, M. H.; Choi, B. K.; Baek, J. H.; Park, J. Y.; Kim, J. H.; Kim, H. G.; Kim, K. H

    2006-07-15

    The main objective of the present R and D on breeder blanket is the development of test blanket modules (TBMs) to be installed and tested in International Thermonuclear Experimental Reactor (ITER). In the program of the blanket development, a blanket module test in the ITER is scheduled from the beginning of the ITER operation, and the performance test of TBM in ITER is the most important milestone for the development of the DEMO blanket. The fabrication of TBMs has been required to test the basic performance of the DEMO blanket, i.e., tritium production/recovery, high-grade heat generation and radiation shielding. Therefore, the integration of the TBM systems into ITER has been investigated with the aim to check the safety, reliability and compatibility under nuclear fusion state. For this reason, in the Test Blanket Working Group (TBWG) as an activity of the International Energy Association (IEA), a variety of ITER TBMs have been proposed and investigated by each party: helium-cooled ceramic (WSG-1), helium-cooled LiPb (WSG-2), water-cooled ceramic (WSG-3), self-cooled lithium (WSG-4) and self-cooled molten salt (WSG-5) blanket systems. Because we are still deficient in investigation of TBM development, the need of development became pressing. In this report, for the development of TBM sub-module and mock-up, it is necessary to analyze and examine the state of the art on the development of manufacturing technology of TBM in other countries. And we will be applied as basic data to establish a manufacturing technology.

  1. Tritium transport analysis in HCPB DEMO blanket with the FUS-TPC Code (KIT Scientific Reports ; 7642)

    OpenAIRE

    Franza, Fabrizio

    2013-01-01

    In thermonuclear fusion reactors, the fuel is an high temperature deuterium-tritium plasma, in which tritium is bred by lithium isotopes present inside solid ceramic breeder (e.g. Li-Orthosilicate) or inside liquid eutectic alloys (e.g. Pb-16Li alloy). In the breeding areas a significant fraction of the tritium produced is extracted out from the Breeding Zone by the He gas purging the breeding ceramic in the Helium Cooled Pebble Bed (HCPB) blanket concept or transported in solution by the owi...

  2. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

    The main function of this low technology blanket is to produce the necessary tritium for INTOR operation with minimum first wall coverage. The INTOR first wall, blanket, and shield are constrained by the dimensions of the reference design and the protection criteria required for different reactor components and dose equivalent after shutdown in the reactor hall. It is assumed that the blanket operation at commercial power reactor conditions and the proper temperature for power generation can be sacrificed to achieve the highest possible tritium breeding ratio with minimum additional research and developments and minimal impact on reactor design and operation. A set of blanket evaluation criteria has been used to compare possible blanket concepts. Six areas: performance, operating requirements, impact on reactor design and operation, safety and environmental impact, technology assessment, and cost have been defined for the evaluation process. A water-cooled blanket was developed to operate with a low temperature and pressure. The developed blanket contains a 24 cm of beryllium and 6 cm of solid breeder both with a 0.8 density factor. This blanket provides a local tritium breeding ratio of ∼2.0. The water coolant is isolated from the breeder material by several zones which eliminates the tritium buildup in the water by permeation and reduces the changes for water-breeder interaction. This improves the safety and environmental aspects of the blanket and eliminates the costly process of the tritium recovery from the water. 12 refs., 13 tabs

  3. Updated reference design of a liquid metal cooled tandem mirror fusion breeder

    International Nuclear Information System (INIS)

    Detailed studies of key techinical issues for liquid metal cooled fusion breeder (fusion-fission hybrid blankets) have been performed during the period 1983-4. Based upon the results of these studies, the 1982 reference liquid metal cooled tandem mirror fusion breeder blanket design was updated and is described. The updated reference blankets provides increased breeding and lower technological risk in comparison with the original reference blanket. In addition to the blanket design revisions, a plant concept, cost, and fuel cycle economics assessment is provided. The fusion breeder continues to promise an economical source of fissile fuel for the indefinite future

  4. Fusion breeder neutronics. Final report

    International Nuclear Information System (INIS)

    Research efforts in fusion breeder neutronics have been focused on two tasks that are strongly related. Efforts in Task 1 concentrate on examining the required conditions to sustain fuel self-sufficiency in fusion reactors operated on a D-T fuel cycle. In this respect, in-depth and detailed engineering analyses have been performed on various blanket and reactor concepts to verify the potential of each blanket concept to exhibit a tritium breeding ratio (TBR) in excess of unity by a margin that compensates for losses, radioactive decay and other inventory requirements. Efforts in Task 2 concentrate on evaluating the overall uncertainties (both experimental and analytical) associated with the TBR

  5. First wall and blanket concepts for experimental fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Biggio, M.; Cardella, A.; Daenner, W.; Farfaletti-Casali, F.; Ponti, C.; Rieger, M.; Vieider, G.

    1985-07-01

    The paper describes the progress of the studies on first wall and liquid breeder blankets for tritium production in the Next European Torus (NET). Two concepts of first wall/blanket segments are described, using 17Li83Pb as breeder and water as coolant. In both concepts the first wall is integrated in a steel box enveloping the breeder units which are cylindrical vessels with an inside heat transfer system. The thermomechanical and neutronics features of the two concepts are evaluated. Finally, the questions related to tritium permeation into coolant and tritium recovery from breeder are discussed on the basis of the analysis in progress in Europe.

  6. Water-cooled blanket concepts for the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    The primary goal of the Blanket Comparison and Selection Study (BCSS) was to select a limited number of blanket concepts for fusion power reactors, to serve as the focus for the U.S. Department of Energy blanket research and development program. The concepts considered most seriously by the BCSS can be grouped for discussion purposes by coolant: liquid metals and alloys, pressurized water, helium, and nitrate salts. Concepts using pressurized water as the coolant are discussed. Water-cooled concepts using liquid breeders-lithium and 17Li-83Pb (LiPb)-have severe fundamental safety problems. The use of lithium and water in the blanket was considered unacceptable. Initial results of tests at Hanford Engineering Development Laboratory using steam injected into molten LiPb indicate that use of LiPb and water together in a blanket is a very serious concern from the safety standpoint. Key issues for water-cooled blankets with solid tritium breeders (Li2O, or a ternary oxide such as LiAlO2) were identified and examined: reliability against leaks, control of tritium permeation into the coolant, retention of breeder physical integrity, breeder temperature predictability, determination of allowable temperature limits for breeders, and 6Li burnup effects (for LiAlO2). The BCSS's final rankings and associated rationale for all water-cooled concepts are examined. Key issues and factors for tokamak and tandem mirror reactor versions of water-cooled solid breeder concepts are discussed. The reference design for the top-ranked concept-LiAlO2 breeder, ferritic steel structure, and beryllium neutron multiplier-is presented. Finally, some general conclusions for water-cooled blanket concepts are drawn based on the study's results

  7. Breeding blanket development; Tritium release from breeder

    OpenAIRE

    土谷 邦彦; 河村 弘; 長尾 美春

    2006-01-01

    核融合炉ブランケットを設計するためには、微小球を用いたブランケット構造体の中性子照射に関する工学的データが必要不可欠である。工学的データのうち、トリチウム生成放出特性は、最も重要なデータの1つである。このため、トリチウム増殖材料の候補材であるチタン酸リチウム(Li2TiO3)微小球からのトリチウム生成放出試験を行い、トリチウム放出特性に対するスイープガス流量,照射温度,スイープガス中の水素添加量,熱中性子束の変化等の効果について調べた。本試験の結果、(1)Li2TiO3微小球充填体の外壁温度が100circC以上になった時、トリチウム放出が観測された。また、充填体の外壁温度が300sim400circCのとき、トリチウム生成・放出率(R/G)は1に到達した。(2)スイープガス流量を100sim900cm3/min(Li2TiO3微小球充填体の空塔速度:0.53sim4.8cm/s)の範囲で変化させても、定常時におけるLi2TiO3微小球充填体からのトリチウム放出に影響はなかった。(3)スイープガス中の水素添加量はトリチウム放出に影響することがわかった。...

  8. Neutronics and thermo-hydraulic design of supercritical-water cooled solid breeder TBM

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Jie; Wu, Yingwei, E-mail: wyw810@mail.xjtu.edu.cn; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-03-15

    Highlights: • A supercritical-water cooled solid breeder test blanket module (SWCB TBM) was designed. • The neutronics calculations show that the tritium breeding ratio (TBR) of SWCB TBM is 1.17. • The outlet temperature of SWCB TBM can reach as high as 500 °C. • Both thermal stress and deformation of the SWCB TBM design are within safety limits. - Abstract: In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li{sub 4}SiO{sub 4} lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER. The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500 °C. According to thermo-mechanics calculation results, the first wall with the width of 17 mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition.

  9. Neutronics and thermo-hydraulic design of supercritical-water cooled solid breeder TBM

    International Nuclear Information System (INIS)

    Highlights: • A supercritical-water cooled solid breeder test blanket module (SWCB TBM) was designed. • The neutronics calculations show that the tritium breeding ratio (TBR) of SWCB TBM is 1.17. • The outlet temperature of SWCB TBM can reach as high as 500 °C. • Both thermal stress and deformation of the SWCB TBM design are within safety limits. - Abstract: In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li4SiO4 lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER. The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500 °C. According to thermo-mechanics calculation results, the first wall with the width of 17 mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition

  10. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    The objectives of the Blanket Comparison and Selection Study (BCSS) can be stated as follows: (1) Define a small number (approx. 3) of blanket design concepts that should be the focus of the blanket R and D program. A design concept is defined by the selection of all materials (e.g., breeder, coolant, structure and multiplier) and other major characteristics that significantly influence the R and D requirements. (2) Identify and prioritize the critical issues for the leading blanket concepts. (3) Provide the technical input necessary to develop a blanket R and D program plan. Guidelines for prioritizing the R and D requirements include: (a) critical feasibility issues for the leading blanket concepts will receive the highest priority, and (b) for equally important feasibility issues, higher R and D priority will be given to those that require minimum cost and short time

  11. Progress on DCLL Blanket Concept

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Clement; Abdou, M.; Katoh, Yutai; Kurtz, Richard J.; Lumsdaine, A.; Marriott, Edward P.; Merrill, Brad; Morley, Neil; Pint, Bruce A.; Sawan, M.; Smolentsev, S.; Williams, Brian; Willms, Scott; Youssef, M.

    2013-09-01

    Under the US Fusion Nuclear Science and Technology Development program, we have selected the Dual Coolant Lead Lithium concept (DCLL) as a reference blanket, which has the potential to be a high performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. The self-cooled breeder PbLi is circulated for power conversion and for tritium breeding. A SiC-based flow channel insert (FCI) is used as a means for magnetohydrodynamic pressure drop reduction from the circulating liquid PbLi and as a thermal insulator to separate the high-temperature PbLi (~700°C) from the helium-cooled RAF/M steel structure. We are making progress on related R&D needs to address critical Fusion Nuclear Science and Facility (FNSF) and DEMO blanket development issues. When performing the function as the Interface Coordinator for the DCLL blanket concept, we had been developing the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. We had estimated the necessary ancillary equipment that will be needed at the ITER site and a detailed safety impact report has been prepared. This provided additional understanding of the DCLL blanket concept in preparation for the FNSF and DEMO. This paper will be a summary report on the progress of the DCLL TBM design and R&Ds for the DCLL blanket concept.

  12. Technical issues for beryllium use in fusion blanket applications

    International Nuclear Information System (INIS)

    Beryllium is an excellent non-fissioning neutron multiplier for fusion breeder and fusion electric blanket applications. This report is a compilation of information related to the use of beryllium with primary emphasis on the fusion breeder application. Beryllium resources, production, fabrication, properties, radiation damage and activation are discussed. A new theoretical model for beryllium swelling is presented

  13. An assessment of the base blanket for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Raffray, A.R.; Abdou, M.A.; Ying, A.

    1991-12-31

    Ideally, the ITER base blanket would provide the necessary tritium for the reactor to be self-sufficient during operation, while having minimal impact on the overall reactor cost, reliability and safety. A solid breeder blanket has been developed in CDA phase in an attempt to achieve such objectives. The reference solid breeder base blanket configurations at the end of the CDA phase has many attractive features such as a tritium breeding ratio (TBR) of 0.8--0.9 and a reasonably low tritium inventory. However, some concerns regarding the risk, cost and benefit of the base blanket have been raised. These include uncertainties associated with the solid breeder thermal control and the potentially high cost of the amount of Be used to achieve high TBR and to provide the necessary thermal barrier between the high temperature solid breeder and low temperature coolant. This work addresses these concerns. The basis for the selection of a breeding blanket is first discussed in light of the incremental risk, cost and benefits relative to a non-breeding blanket. Key issues associated with the CDA breeding blanket configurations are then analyzed. Finally, alternative schemes that could enhance the attractiveness and flexibility of a breeding blanket are explored.

  14. An assessment of the base blanket for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Raffray, A.R.; Abdou, M.A.; Ying, A.

    1991-01-01

    Ideally, the ITER base blanket would provide the necessary tritium for the reactor to be self-sufficient during operation, while having minimal impact on the overall reactor cost, reliability and safety. A solid breeder blanket has been developed in CDA phase in an attempt to achieve such objectives. The reference solid breeder base blanket configurations at the end of the CDA phase has many attractive features such as a tritium breeding ratio (TBR) of 0.8--0.9 and a reasonably low tritium inventory. However, some concerns regarding the risk, cost and benefit of the base blanket have been raised. These include uncertainties associated with the solid breeder thermal control and the potentially high cost of the amount of Be used to achieve high TBR and to provide the necessary thermal barrier between the high temperature solid breeder and low temperature coolant. This work addresses these concerns. The basis for the selection of a breeding blanket is first discussed in light of the incremental risk, cost and benefits relative to a non-breeding blanket. Key issues associated with the CDA breeding blanket configurations are then analyzed. Finally, alternative schemes that could enhance the attractiveness and flexibility of a breeding blanket are explored.

  15. Investigation of heat treatment conditions of structural material for blanket fabrication process

    International Nuclear Information System (INIS)

    This paper presents recent results of thermal hysteresis effects on ceramic breeder blanket structural material. Reduced activation ferritic/martensitic (RAF) steel is the leading candidates for the first wall structural materials of breeding blankets. RAF steel demonstrates superior resistance to high dose neutron irradiation, because the steel has tempered martensite structure which contains the number of sink site for radiation defects. This microstructure obtained by two-step heat treatment, first is normalizing at temperature above 1200 K and the second is tempering at temperature below 1100 K. Recent study revealed the thermal hysteresis has significant impacts on the post-irradiation mechanical properties. The breeding blanket has complicated structure, which consists of tungsten armor and thin first wall with cooling pipe. The blanket fabrication requires some high temperature joining processes. Especially hot isostatic pressing (HIP) is examined as a near-net-shape fabrication process for this structure. The process consists of heating above 1300 K and isostatic pressing at the pressure above 150 MPa followed by tempering. Moreover ceramics pebbles are packed into blanket module and the module is to be seamed by welding followed by post weld heat treatment in the final assemble process. Therefore the final microstructural features of RAFs strongly depend on the blanket fabrication process. The objective of this work is to evaluate the effects of thermal hysteresis corresponding to blanket fabrication process on RAFs microstructure in order to establish appropriate blanket fabrication process. Japanese RAFs F82H (Fe-0.1C-8Cr-2W-0.2V-0.05Ta) was investigated by metallurgical method after isochronal heat treatment up to 1473 K simulating high temperature bonding process. Although F82H showed significant grain growth after conventional solid HIP conditions (1313 K x 2 hr.), this coarse grained microstructure was refined by the post HIP normalizing at

  16. Fusion blanket materials development and recent R and D activities

    International Nuclear Information System (INIS)

    Development of structural materials plays an important role in the feasibility of fusion power plant. The candidate structural materials for future fusion reactors are Reduced Activation Ferritic Martensitic (RAFM) steel, nano structured ODS Steel, vanadium alloys and SiC/SiCf composite etc. RAFM steel is presently considered as the structural material for Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) because of its high void swelling resistance and improved thermal properties compared to austenitic steel. Development of RAFM steel in India is being carried out in full swing in collaboration with various research laboratories and steel industries. This paper presents an overview of the Indian activities on fusion blanket materials and describes in brief the efforts made to develop IN-RAFM steel as structural material for the LLCB TBM. In future, due to enhanced properties of vanadium base alloy and nano structured materials like ODS RAFMS, RAFM steel may be replaced by these materials for its application in DEMO relevant fusion reactor. Future R and D activities will be specifically towards the development of these structural materials for fusion reactor

  17. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    The economic impact of employing the highly ranked blankets in the Blanket Comparison and Selection Study (BCSS) was evaluated in the context of both a tokamak and a tandem mirror power reactor (TMR). The economic evaluation criterion was determined to be the cost of electricity. The influencing factors that were considered are the direct cost of the blankets and related systems; the annual cost of blanket replacement; and the performance of the blanket, heat transfer, and energy conversion systems. The technical and cost bases for comparison were those of the STARFIRE and Mirror Advanced Reactor Study conceptual design power plants. The economic evaluation results indicated that the nitrate-salt-cooled blanket concept is an economically attractive concept for either reactor type. The water-cooled, solid breeder blanket is attractive for the tokamak and somewhat less attractive for the TMR. The helium-cooled, liquidlithium breeder blanket is the least economically desirable of higher ranked concepts. The remaining self-cooled liquid-metal and the helium-cooled blanket concepts represent moderately attractive concepts from an economic standpoint. These results are not in concert with those found in the other BCSS evaluation areas (engineering feasibility, safety, and research and development (R and D) requirements). The blankets faring well economically had generally lower cost components, lower pumping power requirements, and good power production capability. On the other hand, helium- and lithium-cooled systems were preferred from the standpoints of safety, engineering feasibility, and R and D requirements

  18. Conceptual design of Tritium Extraction System for the European HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Highlights: ► HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM) to be tested in ITER. ► Tritium extraction by gas purging, removal and transfer to the Tritium Plant. ► Conceptual design of TES and revision of the previous configuration. ► Main components: adsorption column, ZrCo getter beds and PERMCAT reactor. - Abstract: The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80 bar as primary coolant. In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing. In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported.

  19. Conceptual design of Tritium Extraction System for the European HCPB Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Ciampichetti, A., E-mail: andrea.ciampichetti@enea.it [ENEA CR Brasimone, 40032 Camugnano (Italy); European TBM Consortium of Associates (Germany); Nitti, F.S.; Aiello, A. [ENEA CR Brasimone, 40032 Camugnano (Italy); European TBM Consortium of Associates (Germany); Ricapito, I. [Fusion for Energy, 08019 Barcelona (Spain); Liger, K. [CEA, DEN, DTN/STPA/LIPC, Cadarache, 13108 St. Paul-lez-Durance (France); European TBM Consortium of Associates (Germany); Demange, D. [Karlsruhe Institute of Technology, ITEP-TLK, Postfach 36 40, 76021 Karlsruhe (Germany); European TBM Consortium of Associates (Germany); Sedano, L.; Moreno, C. [EURATOM-CIEMAT Association, 28040 Madrid (Spain); European TBM Consortium of Associates (Germany); Succi, M. [SAES Getters Spa, 20020 Lainate (Italy)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM) to be tested in ITER. Black-Right-Pointing-Pointer Tritium extraction by gas purging, removal and transfer to the Tritium Plant. Black-Right-Pointing-Pointer Conceptual design of TES and revision of the previous configuration. Black-Right-Pointing-Pointer Main components: adsorption column, ZrCo getter beds and PERMCAT reactor. - Abstract: The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80 bar as primary coolant. In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing. In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported.

  20. Laser fusion driven breeder design study. Final report

    International Nuclear Information System (INIS)

    The results of the Laser Fusion Breeder Design Study are given. This information primarily relates to the conceptual design of an inertial confinement fusion (ICF) breeder reactor (or fusion-fission hybrid) based upon the HYLIFE liquid metal wall protection concept developed at Lawrence Livermore National Laboratory. The blanket design for this breeder is optimized to both reduce fissions and maximize the production of fissile fuel for subsequent use in conventional light water reactors (LWRs). When the suppressed fission blanket is compared with its fast fission counterparts, a minimal fission rate in the blanket results in a unique reactor safety advantage for this concept with respect to reduced radioactive inventory and reduced fission product decay afterheat in the event of a loss-of-coolant-accident

  1. Tritium self-sufficiency of HCPB blanket modules for DEMO considering time-varying neutron flux spectra and material compositions

    Energy Technology Data Exchange (ETDEWEB)

    Aures, A., E-mail: Alexander.Aures@ccfe.ac.uk; Packer, L.W.; Zheng, S.

    2013-10-15

    Highlights: • Simulations on the tritium breeding performance of HCPB blanket modules were done. • MCNP5 and FISPACT were used for coupled transport and activation calculations. • Material transmutation affects the neutron flux spectra within the blanket modules. • The consequences of time-dependent spectra on TBR and tritium self-sufficiency were investigated. -- Abstract: Significant transmutation of solid-type breeding blanket materials affects the time and spatial variation of neutron energy within such materials. This has an impact on simulation assumptions required to accurately assess tritium surplus quantities for conceptual power plant devices. This paper details an investigation, via simulation, of the consequences for the tritium breeding ratio and the tritium self-sufficiency of a DEMO concept with homogeneous Helium-Cooled Pebble Bed blanket modules containing Li{sub 4}SiO{sub 4} ceramic breeder material. For this purpose, a code was developed to couple MCNP5 and FISPACT to supply material compositions from activation calculations to the neutron transport calculation in an iterative loop covering several time steps. Simulation results are presented for a simple 1D spherical device model and a DEMO tokamak model.

  2. High power density self-cooled lithium-vanadium blanket.

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.; Majumdar, S.; Smith, D.

    1999-07-01

    A self-cooled lithium-vanadium blanket concept capable of operating with 2 MW/m{sup 2} surface heat flux and 10 MW/m{sup 2} neutron wall loading has been developed. The blanket has liquid lithium as the tritium breeder and the coolant to alleviate issues of coolant breeder compatibility and reactivity. Vanadium alloy (V-4Cr-4Ti) is used as the structural material because it can accommodate high heat loads. Also, it has good mechanical properties at high temperatures, high neutron fluence capability, low degradation under neutron irradiation, good compatibility with the blanket materials, low decay heat, low waste disposal rating, and adequate strength to accommodate the electromagnetic loads during plasma disruption events. Self-healing electrical insulator (CaO) is utilized to reduce the MHD pressure drop. A poloidal coolant flow with high velocity at the first wall is used to reduce the peak temperature of the vanadium structure and to accommodate high surface heat flux. The blanket has a simple blanket configuration and low coolant pressure to reduce the fabrication cost, to improve the blanket reliability, and to increase confidence in the blanket performance. Spectral shifter, moderator, and reflector are utilized to improve the blanket shielding capability and energy multiplication, and to reduce the radial blanket thickness. Natural lithium is used to avoid extra cost related to the lithium enrichment process.

  3. Global depletion analysis of Korean helium cooled solid breeder TBM model for demo fusion reactor

    International Nuclear Information System (INIS)

    The Korean HCSB (helium cooled solid breeder) TBM (test blanket module) is proposed with its specific compositions of lithium ceramic, beryllium and graphite in pebble form. In the Korean HCSB TBM, the amount of beryllium is reduced and the reduction is replaced by graphite for a neutron reflector, while tritium breeding ratio (TBR) remains almost unchanged with relatively low Li6 enrichment of ∼40%. However, the previous Korean HCSB was designed based on the LOCAL assumption, in which the surroundings are assumed by the reflective boundary condition. In this research, we establish a simple GLOBAL neutronics model based on demo fusion reactor and perform neutronics analyses including depletion (transmutation) calculation during 100 EFPDs (effective full power days) using the modified MONTEBURNS code.

  4. Preliminary thermal-hydraulic design and simulation for hybrid breeder blanket%聚变-快裂变增殖堆包层初步热工水力学设计分析

    Institute of Scientific and Technical Information of China (English)

    王小勇; 栗再新; 赵奉超; 赵周; 武兴华; 王琦杰

    2014-01-01

    Thermal-hydraulic design and analysis for the new conceptual design of fusion-fission breeding reactor using casing pipes for fuel assembly was done. Based on typical thermal-hydraulic design parameters, preliminary thermal-hydraulic design for the blanket was proposed. The corresponding temperature distribution and pressure distribution were obtained using thermal-hydraulic codes, CFX. The simulation results showed that maximum temperature of the materials were all below their corresponding temperature limits, coolant temperature at the outlet was higher than 773℃, and pressure drop of the coolant could satisfy engineering requirement. The reasonability of this thermal-hydraulic design was preliminarily verified.%对新提出的套管结构聚变-快裂变增殖堆包层概念设计方案进行了热工水力学分析和设计,给出了典型的热工设计参数,并结合大型热工水力学软件CFX对其进行了温度场和压力分布的模拟分析。分析结果表明,材料温度均已低于许用温度,冷却剂出口温度高于773K,冷却剂压降也符合工程上的要求,初步验证了增殖堆包层设计的合理性。

  5. Preliminary thermal-hydraulic design and simulation for hybrid breeder blanket%聚变-快裂变增殖堆包层初步热工水力学设计分析

    Institute of Scientific and Technical Information of China (English)

    王小勇; 栗再新; 赵奉超; 赵周; 武兴华; 王琦杰

    2014-01-01

    对新提出的套管结构聚变-快裂变增殖堆包层概念设计方案进行了热工水力学分析和设计,给出了典型的热工设计参数,并结合大型热工水力学软件CFX对其进行了温度场和压力分布的模拟分析。分析结果表明,材料温度均已低于许用温度,冷却剂出口温度高于773K,冷却剂压降也符合工程上的要求,初步验证了增殖堆包层设计的合理性。%Thermal-hydraulic design and analysis for the new conceptual design of fusion-fission breeding reactor using casing pipes for fuel assembly was done. Based on typical thermal-hydraulic design parameters, preliminary thermal-hydraulic design for the blanket was proposed. The corresponding temperature distribution and pressure distribution were obtained using thermal-hydraulic codes, CFX. The simulation results showed that maximum temperature of the materials were all below their corresponding temperature limits, coolant temperature at the outlet was higher than 773℃, and pressure drop of the coolant could satisfy engineering requirement. The reasonability of this thermal-hydraulic design was preliminarily verified.

  6. Fission-suppressed hybrid reactor: the fusion breeder

    International Nuclear Information System (INIS)

    Results of a conceptual design study of a 233U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed

  7. Fission-suppressed hybrid reactor: the fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Coops, M.S.

    1982-12-01

    Results of a conceptual design study of a /sup 233/U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed.

  8. Breeder Reprocessing Engineering Test

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, C.A.; Meacham, S.A.

    1984-01-01

    The Breeder Reprocessing Engineering Test (BRET) is a developmental activity of the US Department of Energy to demonstrate breeder fuel reprocessing technology while closing the fuel cycle for the Fast Flux Test Facility (FFTF). It will be installed in the existing Fuels and Materials Examination Facility (FMEF) at the Hanford Site near Richland, Washington, The major objectives of BRET are: (1) close the US breeder fuel cycle; (2) develop and demonstrate reprocessing technology and systems for breeder fuel; (3) provide an integrated test of breeder reactor fuel cycle technology - rprocessing, safeguards, and waste management. BRET is a joint effort between the Westinghouse Hanford Company and Oak Ridge National Laboratory. 3 references, 2 figures.

  9. Analysis of deficiencies in fast reactor blanket physics predictions

    International Nuclear Information System (INIS)

    This analysis addresses a deviation between experimental measurements and fast reactor blanket physics predictions. A review of worldwide results reveals that reaction rates in the blanket are underpredicted with the discrepancy increasing with penetration into the blanket. The analysis of this discrepancy involves two parts: quantifying possible error reductions using the most advanced methods and investigating deficiencies in current methodology. The source of these discrepancies was investigated by application of ''state-of-the-art'' group constant generation and flux prediction methodology to flux calculations for the Purdue University Fast Breeder Blanket Facility (FBBF). Refined group constant generation methods yielded a significant reduction in the blanket deviations; however, only about half of the discrepancy can be accounted for in this manner. Transport theory calculations were used to predict the blanket neutron transmission problem. The surprising result is that transport theory predictions utilizing diffusion theory group constants did not improve the blanket results. Transport theory predictions exhibited blanket underpredictions similar to the diffusion theory results. The residual blanket discrepancies not explained using advanced methods require a refinement of the theory. For this purpose an analysis of deficiencies in current methodology was performed

  10. Breeding blanket for DEMO

    International Nuclear Information System (INIS)

    This paper presents the main design features, their rationale, and the main critical issues for the development, of the four DEMO-relevant blanket concepts presently being investigated within the framework of the European Test-Blanket Development Programme. (orig.)

  11. Breeding blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Proust, E. (Commissariat a l' Energie Atomique (CEA), DRN/DMT/SERMA, CE, Saclay (France)); Anzidei, L. (ENEA/FUS, C.R.E., Frascati (Italy)); Casini, G. (Commission of the European Communities, Joint Research Center, Ispara (Italy)); Dalle Donne, M. (Kernforschungszentrum Karlsruhe GmbH (Germany)); Giancarli, L. (Commissariat a l' Energie Atomique (CEA), DRN/DMT/SERMA, CE, Saclay (France)); Malang, S. (Kernforschungszentrum Karlsruhe GmbH (Germany))

    1993-03-01

    This paper presents the main design features, their rationale, and the main critical issues for the development, of the four DEMO-relevant blanket concepts presently being investigated within the framework of the European Test-Blanket Development Programme. (orig.)

  12. Breeding blanket for Demo

    Energy Technology Data Exchange (ETDEWEB)

    Proust, E.; Giancarli, L. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie; Anzidei, L. [ENEA, Frascati (Italy). Centro Ricerche Energia; Casini, G. [Commission of the European Communities, Ispra (Italy). Joint Research Centre; Dalle Donne, M.; Malang, S. [Kernforschungszentrum Karlsruhe GmbH (Germany)

    1992-12-31

    This paper presents the main design features, their rationale, and the main critical issues for the development, of the four DEMO-relevant blanket concepts presently investigated within the framework of the European Test-Blanket Development Programme.

  13. Thermal and neutronic calculation for fast breeder reactor FBR

    International Nuclear Information System (INIS)

    This research included studying of thermal and neutronic calculation for fast breeder nuclear reactor, to putting the optimum design for this reactor. So a Soviet type (BN-350) was chosen, which has its core composed of two enrichment zones, and with blanket that contains depleted uranium. A group of thermal calculation programs was made by using personal computer, to obtain core and blanket reactor dimensions and volume fractions of reaction input material and number and dimensions of fuel rods which were used for neutron calculations. Several core and blanket enrichments were used to study neutron flux behaviour for two reactors different conditions. First when control rods exist in the core reactor and second when the rods are out of the core. Breeding ratio was also studied for different core and blanket enrichment. 30 tabs.; 24 figs.; 34 refs.; 3 apps

  14. Development of the Helium Cooled Lithium Lead blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aiello, G., E-mail: giacomo.aiello@cea.fr [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Aubert, J.; Jonquères, N. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Li Puma, A. [CEA-Saclay, DEN/DANS/DM2S/SERMA/LPEC, 91191 Gif Sur Yvette Cedex (France); Morin, A.; Rampal, G. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France)

    2014-10-15

    Highlights: • The HCLL blanket design has been modified to adapt it to the 2012 EFDA DEMO specifications. • The new design has been developed with the aim to capitalize on TBM experience in ITER. • A new attachment system for the modules has been proposed. - Abstract: The Helium Cooled Lithium Lead (HCLL) blanket is one of the candidate European blanket concepts selected for the DEMOnstration fusion power plant that should follow ITER. In a fusion power plant, the blanket is one of the key components because of its impact on the plant performance, availability, safety and economics. In 2012, the European Fusion Development Agreement (EFDA) agency issued new specifications for DEMO: this paper describes the work performed to adapt the previous 2007 HCLL-DEMO blanket design to those specifications. A new segmentation has been defined assuming straight surfaces for all blanket modules. Following the Multi Module Segment (MMS) option, all modules are attached to a common back supporting structure which also serves as manifold for Helium and PbLi distribution. A detailed CAD design of the central outboard module has been defined. Thermo-hydraulic and thermo-mechanical analyses on of the First Wall and Breeder Zone have been carried out. For the attachment of the modules to the common backplate, a new solution based on the use of Tie Rods, derived from the design of the corresponding HCLL Test Blanket Module for ITER, has been proposed. This paper also identifies the priorities for further development of the HCLL blanket design.

  15. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been performing the development of breeding blanket for fusion power plant, as a leading institute of the development of solid breeder blankets, according to the long-term R and D program of the blanket development established by the Fusion Council of Japan in 1999. This report is an overview of development plan, achievements of element technology development and future prospect and plan of the development of the solid breeding blanket in JAERI. In this report, the mission of the blanket development activity in JAERI, key issues and roadmap of the blanket development have been clarified. Then, achievements of the element technology development were summarized and showed that the development has progressed to enter the engineering testing phase. The specific development target and plan were clarified with bright prospect. Realization of the engineering test phase R and D and completion of ITER test blanket module testing program, with universities/NIFS cooperation, are most important steps in the development of breeding blanket of fusion power demonstration plant. (author)

  16. Water-cooled lithium-lead blanket

    International Nuclear Information System (INIS)

    The paper is an appendix to a study of the reactor relevance of the NET design concept. The present study examines whether the water-cooled lithium-lead blanket designed for NET can be directly extrapolated to a demonstration (DEMO) reactor. A fundamental requirement of the exercise is that the DEMO design should have a tritium breeding ratio which is higher than that in NET. The water-cooled lithium-lead blanket is discussed with respect to: neutronics design, design parameter survey and thermohydraulics, and engineering design. Results are reported of three-dimensional calculations using the Monte Carlo code MORSE-H to investigate possible neutron leakage between the poloidally disposed breeder tubes, and to determine the global tritium breeding ratio for the final double null machine design. (U.K.)

  17. DEMO blanket testing in ITER. Influence on reaching DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Shatalov, G. E-mail: geshat@nfi.kiae.ru

    2001-10-01

    ITER goal was specified as one step between now and the DEMO fusion reactor. One of the major issues is the tritium breeding blankets test relevant to future reactors. The major objectives of blanket modules (TBM) experiments in ITER are reduced in comparison with proposed test objectives in ITER-FDR. Thus, results of DEMO blanket designs testing in ITER will provide limited (but still useful) information that will need strong support from non-fusion facilities testing. The role of non-fusion tests is increased now to provide additional data required for DEMO blanket construction and qualification. A strategy of testing steps to DEMO blanket qualifications has to include parallel testing in ITER and in non-fusion devices. Experiments in fission reactors are able to provide essential data on materials radiation properties; tritium release, inventory and permeation; and thermomechanical behavior of the blanket breeder/multiplier. However, the volume in fission reactors is rather small and neutron spectra differ from the fusion reactor one. Nonetheless in the near future one depends primarily on fission reactor irradiation. The powerful accelerator based neutron source IFMIF could also provide useful information on radiation material properties. Plasma based neutron sources of different fusion devices could be the best choice for testing DEMO materials and blanket mock-ups. Timetable and costs of these devices are not clear now.

  18. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, a study was carried out to assess the impact of different reactor design choices on the reactor performance parameters. The design choices include the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, and the coolant choice for the nonbreeding inboard blanket. In addition, tritium breeding benchmark calculations were performed using different transport codes and nuclear data libraries. The importance of the TBR in the blanket design motivated the benchmark calculations

  19. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  20. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  1. New simulations to qualify eutectic lithium-lead as breeder material

    OpenAIRE

    Fraile García, Alberto; Cuesta Lopez, Santiago; Caro, Alfredo; Iglesias, R.; Perlado Martin, Jose Manuel

    2011-01-01

    Pb17Li is today a reference breeder material in diverse fusion R&D programs worldwide. One of the main issues is the problem of liquid metals breeder blanket behavior. The knowledge of eutectic properties like optimal composition, physical and thermodynamic behavior or diffusion coefficients of Tritium are extremely necessary for current designs. In particular, the knowledge of the function linking the tritium concentration dissolved in liquid materials with the tritium partial pressure at...

  2. Study of MHD Corrosion and Transport of Corrosion Products of Ferritic/Martensitic Steels in the Flowing PbLi and its Application to Fusion Blanket

    OpenAIRE

    Saeidi, Sheida

    2014-01-01

    Two important components of a liquid breeder blanket of a fusion power reactor are the liquid breeder/coolant and the steel structure that the liquid is enclosed in. One candidate combination for such components is Lead-Lithium (PbLi) eutectic alloy and advanced Reduced Activation Ferritic/Martensitic (RAFM) steel. Implementation of RAFM steel and PbLi in blanket applications still requires material compatibility studies as many questions related to physical/chemical interactions in the RAFM...

  3. Detection of Breeding Blankets Using Antineutrinos

    Science.gov (United States)

    Cogswell, Bernadette; Huber, Patrick

    2016-03-01

    The Plutonium Management and Disposition Agreement between the United States and Russia makes arrangements for the disposal of 34 metric tons of excess weapon-grade plutonium. Under this agreement Russia plans to dispose of its excess stocks by processing the plutonium into fuel for fast breeder reactors. To meet the disposition requirements this fuel would be burned while the fast reactors are run as burners, i.e., without a natural uranium blanket that can be used to breed plutonium surrounding the core. This talk discusses the potential application of antineutrino monitoring to the verification of the presence or absence of a breeding blanket. It is found that a 36 kg antineutrino detector, exploiting coherent elastic neutrino-nucleus scattering and made of silicon, could determine the presence of a breeding blanket at a liquid sodium cooled fast reactor at the 95% confidence level within 90 days. Such a detector would be a novel non-intrusive verification tool and could present a first application of coherent elastic neutrino-nucleus scattering to a real-world challenge.

  4. Fusion blanket testing in MFTF-α + T

    International Nuclear Information System (INIS)

    The Mirror Fusion Test Facility-α + T (MFTF-α + T) is an upgraded version of the current MFTF-B test facility at Lawrence Livermore National Laboratory, and is designed for near-term fusion-technology-integrated tests at a neutron flux of 2 MW/m2. Currently, the fusion community is screening blanket and related issues to determine which ones can be addressed using MFTF-α + T. In this work, the minimum testing needs to address these issues are identified for the liquid-metal-cooled blanket and the solid-breeder blanket. Based on the testing needs and on the MFTF-α + T capability, a test plan is proposed for three options; each option covers a six to seven year testing phase. The options reflect the unresolved question of whether to place the research and development (R and D) emphasis on liquid-metal or solid-breeder blankets. In each case, most of the issues discussed can be addressed to a reasonable extent in MFTF-α+T

  5. First wall and blanket module safety enhancement by material selection and design decision

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, B.J.

    1980-01-01

    A thermal/mechanical study has been performed which illustrates the behavior of a fusion reactor first wall and blanket module during a loss of coolant flow event. The relative safety advantages of various material and design options were determined. A generalized first wall-blanket concept was developed to provide the flexibility to vary the structural material (stainless steel vs titanium), coolant (helium vs water), and breeder material (liquid lithium vs solid lithium aluminate). In addition, independent vs common first wall-blanket cooling and coupled adjacent module cooling design options were included in the study. The comparative analyses were performed using a modified thermal analysis code to handle phase change problems.

  6. Safeguards in Prototype Fast Breeder Reactor Monju

    International Nuclear Information System (INIS)

    The assemblies loaded in the core and stored in the ex-vessel storage tank (EVST) are in liquid sodium in the Japanese prototype fast breeder reactor (FBR) Monju. Since it is difficult to apply a direct verification procedure for the fuel assemblies in these areas, a dual containment and surveillance system consisting of two monitoring devices such as surveillance camera and radiation monitor that are functionally independent has been applied. In addition, the Monju Remote Monitoring System was developed to strengthen the continuous surveillance and to reduce the load of the inspection activities. Furthermore, the ex-vessel transfer machine radiation monitor (EVRM) and the exit gate monitor (EXGM) were upgraded to strengthen the monitoring of spent blanket fuel assemblies and to improve the reliability of distinguishing between fuel assemblies and non-fuel items. As the result, the integrated safeguards was introduced in November 2009, and the effective safeguards activities have been implemented in Monju. (author)

  7. Direct LiT Electrolysis in a Metallic Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Luke [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-30

    A process that simplifies the extraction of tritium from molten lithium based breeding blankets was developed. The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fusion/fission reactors is critical in order to maintain low concentrations. This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Extraction is complicated due to required low tritium concentration limits and because of the high affinity of tritium for the blanket. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering hydrogen and deuterium through an electrolysis step at high temperatures.

  8. Direct Lit Electrolysis In A Metallic Lithium Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Colon-Mercado, H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Babineau, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Elvington, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Garcia-Diaz, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Teprovich, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Vaquer, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-10-13

    A process that simplifies the extraction of tritium from molten lithium based breeding blankets was developed.  The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fission/fusion reactors is critical in order to maintained low concentrations.  This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Because of the high affinity of tritium for the blanket, extraction is complicated at the required low levels. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering the hydrogen and deuterium thru an electrolysis step at high temperatures. 

  9. Safety and personnel access aspects of low activation fusion blankets

    International Nuclear Information System (INIS)

    The use of silicon carbide and carbon materials for structural applications in fusion reactor first wall and blanket regions has been proposed and a continuing effort spent on the development of the ceramics technology. The advantages identified are an extremely low induced radioactivity inventory, a high temperature operating capability, abundant raw material resource availability, and minimized plasma impurity effects. One of the unique features of the applications of these materials to fusion reactor blanket designs is that no alloying element is needed in order to assure the specified mechanical properties such as occurs in metal alloys. The major source of long term radioactivity in these materials is impurities. The impurity elements and their concentrations carried over to the blanket structure during fabrication can be minimized by proper fabrication procedures and techniques. The safety and personnel access aspects of such fusion blankets in conjunction with the impurity element concentration are the main subjects of this paper

  10. Flexible armored blanket development

    Energy Technology Data Exchange (ETDEWEB)

    Roth, E.S.

    1978-05-01

    An exploratory development contract was undertaken on December 23, 1977 which had as its purpose the development and demonstration of a flexible armored blanket design suitable for providing ballistic protection to nuclear weapons during shipment. Objectives were to design and fabricate a prototype blanket which will conform to the weapon shape, is troop-handleable in the field, and which, singly or in multiple layers, can defeat a range of kinetic energy armor piercing (AP) ammunition potentially capable of damaging the critical portion of the nuclear weapon. Following empirical testing, including the firing of threat ammunition under controlled laboratory and field test conditions, materials were selected and assembled into two blanket designs, each weighing approximately 54 kg/m{sup 2} (11 lbs/ft{sup 2}) and estimated to cost from $111 to $180 per ft{sup 2} in production. A firing demonstration to evidence blanket performance against terrorist/light infantry weapons, heavy infantry weapons, and aircraft cannon was conducted for representatives of the DOD and interested Sandia employees on April 12, 1978. The blankets performed better than anticipated defeating bullets up to 7.62 mm x 51 mm AP with one layer and projectiles up to 23 mm HEI with two layers. Based on these preliminary tests it is recommended that development work be continued with the following objectives: (1) the selection by the DOD of priority applications, (2) the specific design and fabrication of sufficient quantities of armored blankets for field testing, (3) the evaluation of the blankets by DOD operational units, with reports to Sandia Laboratories to enable final design.

  11. Compatibility of structural materials with fusion reactor coolant and breeder fluids

    International Nuclear Information System (INIS)

    Fusion reactors are characterized by a lithium-containing blanket, a heat transfer medium that is integral with the blanket and first wall, and a heat engine that couples to the heat transfer medium. A variety of lithium-containing substances have been identified as potential blanket materials, including molten lithium metal, molten LiF--BeF2, Pb--Li alloys, and solid ceramic compounds such as Li2O. Potential heat transfer media include liquid lithium, liquid sodium, molten nitrates, water, and helium. Each of these coolants and blankets requires a particular set of chemical and mechanical properties with respect to the associated reactor and heat engine structural materials. This paper discusses the materials factors that underlie the selection of workable combinations of blankets and coolants. It also addresses the materials compatibility problems generic to those blanket-coolant combinations currently being considered in reactor design studies

  12. A water cooled, lithium lead breeding blanket for a DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rieger, M.; Biggio, M.; Farfaletti-Casali, F.; Tominetti, S.; Wu, J.; Zucchetti, M. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Labbe, P.; Baraer, L.; Gervaise, G.; Giancarli, L.; Roze, M.; Severi, Y.; Quintric-Bossy, J. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1991-04-01

    The main features of a tritium breeding blanket for a Demonstration Power Reactor involving the eutectic Pb-17Li as liquid breeder and water as coolant are presented. The configuration of the blanket segments and breeder modules as well as their arrangement inside the reactor vacuum vessel are outlined. The main design aspects and the corresponding design limits are reviewed, namely those related to thermomechanics, neutronics, magneto-hydrodynamics, tritium permeation and recovery. First results of safety analysis, in particular those connected with the rupture of a coolant tube in the breeder module are presented and discussed. As a conclusion, the feasibility of the concept look attractive. A problem which requires further investigation is that of the tritium self-sufficiency. It is shown that a net tritium production near to one can be obtained if berylium tiles are placed in front of the plasma, provided that they are cooled by heavy water. (orig.).

  13. Tritium breeding mock-up experiments containing lithium titanate ceramic pebbles and lead irradiated with DT neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Jakhar, Shrichand; Abhangi, M.; Tiwari, S. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India); Makwana, R. [Department of Physics, MS University, Vadodara (India); Chaudhari, V.; Swami, H.L.; Danani, C.; Rao, C.V.S.; Basu, T.K. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India); Mandal, D.; Bhade, Sonali; Kolekar, R.V.; Reddy, P.J. [Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Bhattacharyay, R.; Chaudhuri, P. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India)

    2015-06-15

    Highlights: • Breeding benchmark experiment on LLCB TBM in ITER was performed. • Nuclear responses measured are TPR and reaction rate of {sup 115}In(n, n′){sup 115m}In reaction. • Measured responses are compared with calculations by MCNP and FENDL 2.1 library. • TPR measurements agree with calculations in the estimated error bar. • Measured {sup 115}In(n, n′){sup 115m}In reaction rates are underestimated by the calculations. - Abstract: Experiments were conducted with breeding blanket mock-up consisting of two layers of breeder material lithium titanate pebbles and three layers of pure lead as neutron multiplier. The radial dimensions of breeder, neutron multiplier and structural material layers are similar to the current design of the Indian Lead–Lithium cooled Ceramic Breeder (LLCB) blanket. The mock-up assembly was irradiated with 14 MeV neutrons from DT neutron generator. The local tritium production rates (TPR) from {sup 6}Li and {sup 7}Li in breeder layers were measured with the help of two different compositions of Li isotopes (60.69% {sup 6}Li and 7.54% {sup 6}Li) in Li{sub 2}CO{sub 3}. Tritium production in the multiplication layers were also measured with above mentioned two types of pellets to compare the experimental tritium production with calculations. TPR from {sup 6}Li at one location in the breeder layer was also measured by direct online measurement of tritons from {sup 6}Li(n, t){sup 4}He reaction using silicon surface barrier detector and {sup 6}Li to triton converter. Additional verification of neutron spectra (E{sub n} > 0.35 MeV) in the mock-up zones were obtained by measuring {sup 115}In(n, n′){sup 115m}In reaction rate and comparing it with calculated values in all five layers of mock-up. All the measured nuclear responses were compared with transport calculations using code MCNP with FENDL2.1 and FENDL3.0 cross-section libraries. The average C/E ratio for tritium production in enriched Li{sub 2}CO{sub 3} pellets was 1

  14. Multiplier, moderator, and reflector materials for lithium-vanadium fusion blankets.

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.; Smith, D. L.

    1999-10-07

    The self-cooled lithium-vanadium fusion blanket concept has several attractive operational and environmental features. In this concept, liquid lithium works as the tritium breeder and coolant to alleviate issues of coolant breeder compatibility and reactivity. Vanadium alloy (V-4Cr-4Ti) is used as the structural material because of its superior performance relative to other alloys for this application. However, this concept has poor attenuation characteristics and energy multiplication for the DT neutrons. An advanced self-cooled lithium-vanadium fusion blanket concept has been developed to eliminate these drawbacks while maintaining all the attractive features of the conventional concept. An electrical insulator coating for the coolant channels, spectral shifter (multiplier, and moderator) and reflector were utilized in the blanket design to enhance the blanket performance. In addition, the blanket was designed to have the capability to operate at high loading conditions of 2 MW/m{sup 2} surface heat flux and 10 MW/m{sup 2} neutron wall loading. This paper assesses the spectral shifter and the reflector materials and it defines the technological requirements of this advanced blanket concept.

  15. Special topics reports for the reference tandem mirror fusion breeder. Volume 4. Structural analysis

    Energy Technology Data Exchange (ETDEWEB)

    Orient, G.; Westmann, R.A.; Ghoniem, N.M.; Garner, J.K.; Gromada, R.G.

    1984-12-01

    This report presents a structural analysis of the reference fission suppressed fusion breeder blanket. An axisymmetric structural model is used to analyze thermal and pressure stresses in the blanket. Results indicate that the first wall must be decoupled from the back of the blanket to avoid large thermal stresses. The composite first wall appears to be adequate to resist buckling, and is further strengthened by radial diaphragms. Semieliptical closures for the module ends appear to be feasible, although the attachment of these end closures to the composite first wall has not been analyzed. Radiation effects have not been included in the structural model, but an assessment of creep and swelling indicates a 4 to 5 year blanket life at an assumed strain limit of 2%. Design modifications which will reduce thermal stresses and simplify manufacturing are recommended.

  16. Helium-cooled molten-salt fusion breeder

    International Nuclear Information System (INIS)

    We present a new conceptual design for a fusion reactor blanket that is intended to produce fissile material for fission power plants. Fast fission is suppressed by using beryllium instead of uranium to multiply neutrons. Thermal fission is suppressed by minimizing the fissile inventory. The molten-salt breeding medium (LiF + BeF2 + ThF4) is circulated through the blanket and to the on-line processing system where 233U and tritium are continuously removed. Helium cools the blanket and the austenitic steel tubes that contain the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion by molten salt. We estimate that a breeder having 3000 MW of fusion power will produce 6500 kg of 233U per year. This amount is enough to provide makeup for 20 GWe of light-water reactors per year or twice that many high-temperature gas-cooled reactors or Canadian heavy-water reactors. Safety is enhanced because the afterheat is low and blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times a light-water reactor of the same power. The estimated cost of the 233U produced is $40/g for fusion plants costing 2.35 times that of a light-water reactor if utility owned or $16/g if government owned

  17. A high tritium breeding ratio (TBR) blanket concept and requirements for nuclear data relating to TBR

    International Nuclear Information System (INIS)

    Significance of developing a blanket having a sufficiently larger tritium breeding ratio (TBR) than 1.0 is discussed. For this purpose, a high TBR blanket with a front breeder zone just before the multiplier is introduced together with conventional blankets. From discussion of TBR characteristics in these blankets, the necessity of improving on nuclear data, i.e. reducing uncertainties is presented as follows; σs, σnp and σnα of structural and coolant materials, and σn2n of the multiplier at higher energies above several MeV, and σnγ of these materials and σnαT of 6Li at energies from several hundred keV to thermal energy. (author)

  18. Experimental Investigation of Ternary Alloys for Fusion Breeding Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, B. William [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chiu, Ing L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-26

    Future fusion power plants based on the deuterium-tritium (DT) fuel cycle will be required to breed the T fuel via neutron reactions with lithium, which will be incorporated in a breeding blanket that surrounds the fusion source. Recent work by LLNL proposed the used of liquid Li as the breeder in an inertial fusion energy (IFE) power plant. Subsequently, an LDRD was initiated to develop alternatives ternary alloy liquid metal breeders that have reduced chemical reactivity with water and air compared to pure Li. Part of the work plan was to experimentally investigate the phase diagrams of ternary alloys. Of particular interest was measurement of the melt temperature, which must be low enough to be compatible with the temperature limits of the steel used in the construction of the chamber and heat transfer system.

  19. Development of the water cooled lithium lead blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Aiello, G.; Jonquères, N. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Li Puma, A. [CEA-Saclay, DEN/DANS/DM2S/SERMA/LPEC, 91191 Gif Sur Yvette Cedex (France); Morin, A.; Rampal, G. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France)

    2014-10-15

    Highlights: • The WCLL blanket design has been modified to adapt it to the 2012 EFDA DEMO specifications. • Preliminary CAD design of the equatorial outboard module of the WCLL blanket has been developed for DEMO. • Finite elements analyses have been carried out in order to assess the module thermal behavior in the straight part of the module. - Abstract: The water cooled lithium lead (WCLL) blanket, based on near-future technology requiring small extrapolation from present-day knowledge both on physical and technological aspect, is one of the breeding blanket concepts considered as possible candidates for the EU DEMOnstration power plant. In 2012, the EFDA agency issued new specifications for DEMO: this paper describes the work performed to adapt the WCLL blanket design to those specifications. Relatively small modules with straight surfaces are attached to a common Back Supporting Structure housing feeding pipes. Each module features reduced activation ferritic-martensitic steel as structural material, liquid Lithium-Lead as breeder, neutron multiplier and carrier. Water at typical Pressurized Water Reactors (PWR) conditions is chosen as coolant. A preliminary design of the equatorial outboard module has been achieved. Finite elements analyses have been carried out in order to assess the module thermal behavior. Two First Wall (FW) concepts have been proposed, one favoring the thermal efficiency, the other favoring the manufacturability. The Breeding Zone has been designed with C-shaped Double-Walled Tubes in order to minimize the Water/Pb-15.7Li interaction likelihood. The priorities for further development of the WCLL blanket concept are identified in the paper.

  20. Preliminary test for reprocessing technology development of tritium breeders

    International Nuclear Information System (INIS)

    In order to develop the reprocessing technology of lithium ceramics (Li2TiO3, CaO-doped Li2TiO3, Li4SiO4 and Li2O) as tritium breeder materials for fusion reactors, the dissolution methods of lithium ceramics to recover 6Li resource and the purification method of their lithium solutions to remove irradiated impurities (60Co) were investigated. In the present work, the dissolving rates of lithium from each lithium ceramic powder using chemical aqueous reagents such as HNO3, H2O2 and citric acid (C6H8O7 . H2O) were higher than 90%. Further the decontamination rate of 60Co added into the solutions dissolving lithium ceramics was higher than 97% using the activated carbon impregnated with 8-hydroxyquinolinol as chelate agent.

  1. Preparation and characterization of Li{sub 4}SiO{sub 4} ceramic pebbles by graphite bed method

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Ming; Zhang, Yingchun, E-mail: zycustb@163.com; Xiang, Maoqiao; Liu, Zhiang

    2015-06-15

    Highlights: • Lithium orthosilicate pebbles were fabricated by a new graphite bed process. • Two routes using different raw materials have been conducted in this work. • The fabricated pebbles exhibit a high relative density with uniform microstructure. • This method is short and simple as the pebbles could be fabricated in a continuous process. - Abstract: Lithium-based ceramics have long been recognized as tritium breeding materials in fusion reactor blankets. Lithium orthosilicate (Li{sub 4}SiO{sub 4}) is one of these materials and has been recommended by many ITER research teams as the first selection for the solid tritium breeder. In this paper, the fabrication of Li{sub 4}SiO{sub 4} pebbles used as tritium breeder by a graphite bed method was studied for the first time. Ceramic powders and deionized water were mixed and ball milled to obtain homogeneous suspensions. And then the ceramic suspensions were dispersed on spread graphite powder through nozzles. Spherical droplets with highly uniform size were formed by the surface tension of the liquid droplets. The droplets converted into green pebbles after drying. After calcination and sintering, Li{sub 4}SiO{sub 4} pebbles with desired size and shape were prepared. The obtained Li{sub 4}SiO{sub 4} pebbles had narrow size distribution and favorable sphericity. Thermal analysis, phase analysis and microstructure observation of the pebbles were carried out systematically. Properties of the prepared pebbles were also characterized for crushing load strength, density and porosity, etc. The values were found to be conforming to the desired properties for used as solid breeder.

  2. Blanket comparison and selection study. Final report. Volume 2

    International Nuclear Information System (INIS)

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li2O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N2) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concepts are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li2O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

  3. Blanket comparison and selection study. Final report. Volume 1

    International Nuclear Information System (INIS)

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li2O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N2) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li2O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

  4. Blanket comparison and selection study. Final report. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concepts are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

  5. Blanket comparison and selection study. Final report. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concepts are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

  6. Blanket comparison and selection study. Final report. Volume 3

    International Nuclear Information System (INIS)

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li2O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N2) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li2O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concepts are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

  7. Blanket comparison and selection study. Final report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

  8. Safety analysis of a loss-of-coolant accident in a breeding blanket for experimental fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rocco, P.; Casini, G.; Djerassi, H.; Papa, L.; Pautasso, G.; Renda, V.; Rouyer, J.L.

    1985-07-01

    A LOCA in a blanket design proposed for NET (Next European Torus) is investigated. The structural analysis of a damaged breeder unit shows that this first containment barrier has a high probability of survival to this accident. The radioactive sources involved are evaluated and an assessment is made of all containment barriers and associated protection systems.

  9. Trade-off study of liquid-metal self-cooled blankets

    International Nuclear Information System (INIS)

    A trade-off study of liquid-metal self-cooled blankets was carried out to define the performance of these blankets with respect to the main functions in a fusion reactor, and to determine the potential to operate at the maximum possible values of the performance parameters. The main purpose is to improve the reactor economics by maximizing the blanket energy multiplication factor, reduce the capital cost of the reactor, and satisfy the design requirements. The main parameters during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the 6Li enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, the impact of different reactor design choices on the performance parameters was analyzed. The effect of the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, the coolant choice for the nonbreeding inboard blanket, and the neutron source distribution were part of the trade-off study. In addition, tritium breeding benchmark calculations were performed to study the impact of the use of different transport codes and nuclear data libraries. The importance and the negative effect of high TBR on the energy multiplication motivated the benchmark calculations

  10. Can the breeder go commercial

    International Nuclear Information System (INIS)

    Contrary to some beliefs in the electric utility industry that ERDA is committed to developing a commercial breeder economy, it is pointed out that ERDA isn't even willing to pay the total cost of the R and D program--and unless there is a major commitment from the private sector (the electric utility industry, in particular) the breeder program will die. The schedule as of Fall 1976 called for: (1) Fast Flux Test Facility (scheduled to go critical in 1979, operate in 1980); (2) Clinch River Breeder Reactor Project (CRBRP) (1/3 commercial size plant hopefully operating by 1983); (3) Prototype Large Breeder Reactor (planned construction starting in 1981, operating in 1988); and (4) Commercial Breeder Reactor (CBR-1 design work to start in 1983, construction in 1986, and operation in 1993). The $257 million the utility industry has pledged to the CRBRP was just for openers. The $2 billion follow-on breeder project being designed calls for massive capital input from a utility (or utility consortium)--and if that is not forthcoming, then in the words of an ERDA official, ''we'll have to reassess the whole breeder program.''

  11. Blanket concept with liquid Li/sub 17/Pb/sub 83/ for tritium breeding in INTOR-NET

    Energy Technology Data Exchange (ETDEWEB)

    Airola, J.; Biggio, M.; Casini, G.; Farfaletti-Casali, F.; Li Bassi, P.; Ponti, C.; Rieger, M. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Piana, C. (Milan Univ. (Italy))

    1984-04-01

    A blanket concept with eutectic Li/sub 17/Pb/sub 83/ as liquid breeder, suited for tritium production in an experimental Tokamak power reactor is outlined and discussed. This design has been developed to satisfy the INTOR-Phase-I specifications, in particular: (I) modular arrangement of the blanket units inside the vacuum vessel; (II) no use of the heat deposited for electricity production, (III) a net tritium breeding of a least 60%. In this article the main results of the neutronics and thermohydraulics analysis are reviewed and the problems identified. Methods to keep liquid in the breeder during operation are proposed and discussed. The consequences of a coolant tube rupture in a breeder unit appears to be the most serious problem.

  12. Requirements for helium cooled pebble bed blanket and R and D activities

    Energy Technology Data Exchange (ETDEWEB)

    Carloni, D., E-mail: dario.carloni@kit.edu; Boccaccini, L.V.; Franza, F.; Kecskes, S.

    2014-10-15

    This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R and D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM. The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R and D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R and D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.

  13. Requirements for helium cooled pebble bed blanket and R and D activities

    International Nuclear Information System (INIS)

    This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R and D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM. The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R and D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R and D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine

  14. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

    The blanket of the present invention can keep the temperature of breeding materials within a predetermined range even if the breeding materials are consumed and the amount of heat generated from the breeding materials is reduced, thereby enabling to release tritium stably. That is, a neutron incident amount control means is disposed to the blanket for controlling the amount of neutrons incident to the breeding materials. Alternatively, a material to form hollow layers are disposed to the periphery of the breeding materials. With such constitution, the neutron incident amount control means enables to control the incident amount of neutrons from plasmas to the breeding materials, thereby enabling to suppress the change of the amount of heat generated in the breeding materials. In addition, the hollow layers formed at the periphery of the breeding materials enables selective filling of fluids having different heat transfer characteristics thereby enabling to control heat resistance between the breeding materials and cooling tubes. Accordingly, temperature of the breeding materials can be kept constant even in any of the cases. (I.S.)

  15. Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

    Energy Technology Data Exchange (ETDEWEB)

    Malara, C. [Institute Regional des Materiaux Avances, Ispra (Italy); Casini, G. [Systems Engineering and Informatics Institute, JRC Ispra, Ispra (Vatican City State, Holy See) (Italy); Viola, A. [Department of Chemical Engineering, University of Cagliari, Cagliari (Italy)

    1995-03-01

    The question of the tritium recovery in water-cooled Pb17Li blankets has been under investigation for several years at JRC Ispra. The method which has been more extensively analysed is that of slowly circulating the breeder out from the blanket units and of extracting the tritium from it outside the plasma vacuum vessel by helium gas purging or vacuum degassing in a suited process apparatus. A computerized model of the tritium behaviour in the blanket units and in the extraction system was developed. It includes four submodels: (1) tritium permeation process from the breeder to the cooling water as a function of the local operative conditions (tritium concentration in Pb17Li, breeder temperature and flow rate); (2) tritium mass balance in each breeding unit; (3) tritium desorption from the breeder material to the gas phase of the extraction system; (4) tritium extraction efficiency as a function of the design parameters of the recovery apparatus. In the present paper, on the basis of this model, a parametric study of the tritium permeation rate in the cooling water and of the tritium inventory in the blanket is carried out. Results are reported and discussed in terms of dimensionless groups which describe the relative effects of the overall resistance on tritium transfer to the cooling water (with and without permeation barriers), circulating Pb17Li flow rate and extraction efficiency of the tritium recovery unit. The parametric study is extended to the recovery unit in the case of tritium extraction by helium purge or vacuum degassing in a droplet spray unit. (orig.).

  16. Blanket concept of water-cooled lithium lead with beryllium for the SlimCS fusion DEMO reactor

    International Nuclear Information System (INIS)

    As an advanced option for SlimCS blanket, conceptual design study of water-cooled lithium lead (WCLL) blanket was performed. In SlimCS, the net tritium breeding ratio (TBR) supplied from WCLL blanket was not enough because the thickness of blanket in SlimCS was limited to about 0.5 m so as to allocate the conducting shell position near the plasma for high beta access and vertical stability of plasma. Therefore, the beryllium (Be) pebble bed was adopted as additional multiplier to reach a required TBR (≥ 1.05). Considering the operating temperature of blanket materials, a double pipe structure was adopted. The nuclear and thermal analysis were carried out by a nuclear-thermal-coupled code, ANIHEAT and DOHEAT so that blanket materials were appropriately arranged to satisfy the acceptable operation temperatures. The temperatures of materials were kept in appropriate range for the neutron wall load Pn = 5 MW/m2. It was found that the local TBR of WCLL with Be blanket was comparable with that of solid breeder blanket. (author)

  17. Neutronic studies of fissile and fusile breeding blankets

    International Nuclear Information System (INIS)

    In light of the need of convincing motivation substantiating expensive and inherently applied research (nuclear energy), first a simple comparative study of fissile breeding economics of fusion fission hybrids, spallators and also fast breeder reactors has been carried out. As a result, the necessity of maximization of fissile production (in the first two ones, in fast breeders rather the reprocessing costs should be reduced) has been shown, thus indicating the design strategy (high support ratio) for these systems. In spite of the uncertainty of present projections onto further future and discrepancies in available data even quite conservative assumptions indicate that hybrids and perhaps even earlier - spallators can become economic at realistic uranium price increase and successfully compete against fast breeders. Then on the basis of the concept of the neutron flux shaping aimed at the correlation of the selected cross-sections with the neutron flux, the indications for the maximization of respective reaction rates has been formulated. In turn, these considerations serve as the starting point for the guidelines of breeding blanket nuclear design, which are as follows: 1) The source neutrons must face the multiplying layer (of proper thickness) of possibly low concentration of nuclides attenuating the neutron multiplication (i.e. structure materials, nongaseous coolants). 2) For the most effective trapping of neutrons within the breeding zone (leakage and void streaming reduction) it must contain an efficient moderator (not valid for fissile breeding blankets). 3) All regions of significant slow flux should contain 6Li in order to reduce parasite neutron captures in there. (orig./HP)

  18. Physics aspects of metal fuelled fast reactors with thorium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Mohapatra, D.K., E-mail: dina@igcar.gov.in; Singh, S.S.; Riyas, A.; Mohanakrishnan, P.

    2013-12-15

    Metal fuelled fast breeder reactors (MFBR) with high breeding ratio will play a major role in meeting the high nuclear power growth envisaged in India. In this regard several conceptual reactor designs with alloys of U–Pu–Zr fuel have been suggested for commercial operations. This study focusses on the physics design aspects of a sodium cooled U–Pu–6%Zr fuelled 1000 MWe fast breeder reactor, which can attain a breeding ratio of nearly 1.5. The calculation results on reactor kinetics and safety parameters of the 1000 MWe MFBR are presented. The changes in the breeding ratio by introduction of thorium in the blankets of the MFBR are also investigated. Burnup analyses are carried out to compare the core burnup effects in MOX and metal fuelled FBRs. Since the MOX fuelled 500 MWe prototype fast breeder is getting constructed at IGCAR, for burnup comparisons a MFBR of similar design is considered. The results of this study indicate that the loss of reactivity in the metal core with burnup is less than half that of a MOX core and its breeding ratio remains nearly constant. It is also found that the isotopic composition of plutonium (Pu-vector composition) remains more steady with burnup in a metal core.

  19. Overview of requirements and design integration for the ITER EU Test Blanket Systems instrumentation

    International Nuclear Information System (INIS)

    The ITER project aims at building a fusion device with the general goal of demonstrating the scientific and technical feasibility of fusion power. The testing of Tritium Breeder Blanket concepts is one of the ITER missions and has been recognized as an essential milestone in the development of a future fusion reactor ensuring tritium self-sufficiency, extraction of high grade heat and electricity production. Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The strategy for the development of the instrumentation of the HCLL and HCPB Test Blanket Systems, which include the TBMs and their Ancillary Systems, is briefly recalled in this paper, along with the overview of the requirements coming from the harsh operational environment and the main challenges related to the integration with the complex design of the TBS components. (authors)

  20. High temperature blankets for non-electrical/electrical applications of fusion reactors: Annual report, [1983

    International Nuclear Information System (INIS)

    During FY '83 the Li2O solid-breeder, helium-cooled canister blanket emerged as the LLNL-UW choice for driving the low-temperature (2, high-temperature outer zone for driving the GA hydrogen synfuel process. Providing 3-dimensional neutronics analysis of power deposition and tritium breeding in both blankets was an important part of the UW-Rowe and Assoc. work. In both the LLNL-UW and MARS studies, the fusion driver as the Axi-Cell, A-cell version of the tandem mirror reactor (TMR). Physics parameters consistent with the synfuel interface were determined as part of the work. Defining and analyzing the thermal-electric interfaces between the TMR and the synfuel process continues to be of prime importance. The analysis of thermal transport and energy conversion in the interface, as well as thermal hydraulics analysis of the blanket, were part of the UW-Rowe Assoc. work

  1. Tritium recovery in Pb17Li-water cooled blanket systems

    Energy Technology Data Exchange (ETDEWEB)

    Malara, C. [Safety Technology Inst., Ispra (Italy); Casini, G. [Systems Engineering & Information Inst., Ispra (Italy); Viola, A. [Univ. of Cagliari (Italy)

    1994-12-31

    The question of tritium recovery in Pb17Li, water cooled blankets is under investigation since several years at JRC Ispra. The method which has been more extensively analyzed is that of slowly circulating the breeder out from the blanket units and of extracting the tritium from it outside the plasma vacuum vessel by helium gas purging in a suited process apparatus. The design features of the process systems are related to: (1) the very low tritium solubility in Pb17Li which implies high permeation rates through the containment structures; (2) the need of keeping as low as possible the tritium concentration in the cooling water both for safety and economical reasons. A computerized model of the tritium behavior in the blanket units and in the extraction system has been developed.

  2. Lithium ceramics: sol-gel preparation and tritium release; Ceramiques lithiees: elaboration sol-gel et relachement du tritium

    Energy Technology Data Exchange (ETDEWEB)

    Renoult, O.

    1994-04-01

    Ceramics based on lithium aluminate (LiA1O{sub 2}), lithium zirconate (Li{sub 2}ZrO{sub 3}) and lithium titanate (Li{sub 2}TiO{sub 3}) are candidates as tritium breeder blanket materials for forthcoming nuclear fusion reactors. Lithium silico-aluminate Li{sub 4+x}A1{sub 4-3x}Si{sub 2x}O{sub 8} (0 {<=} x {<=} 0,25) powders were synthetized from alkoxyde-hydroxyde sol-gel route. By direct sintering at 850-1100 deg C (without prior calcination), ceramics with controlled stoichiometry and homogenous microstructure were obtained. We have also prepared, using a comparable method, Li{sub 2}Zr{sub 1-x}Ti{sub x}O{sub 3} (x = 0, x = 0,1 et x = 1) materials. All these ceramics, with different microstructures and compositions, have been tested in out-of-reactor experiments. Concerning lithium aluminate microporous ceramics, the silicon substitution leads to a significant improvement of the tritrium release. Classical models taking into account independent surface mechanisms are not able to describe correctly the observed tritium release kinetics. We show, using a simple model, that the release kinetics is in fact limited by an intergranular diffusion followed by a desorption. The delay in tritium release, which occurs when the ceramic compacity increases, is explained in terms of an enhancement of the ionic T{sup +} diffusion path length. The energy required for desorption includes a leading term independent of hydrogen contained in the sweep gas. This term is attributed to the limiting recombination step of T{sup +} in molecular species HTO. For similar microstructures, the facility of tritium release for the different studied materials is explained by three properties: the crystal structure of the ceramic, the acidity of oxides and finally the presence of electronic non-stoichiometric defects. (author). 89 refs., 50 figs., 2 tabs., 1 annexe.

  3. Safeguards in the prototype fast breeder reactor MONJU

    Energy Technology Data Exchange (ETDEWEB)

    Usami, S.; Deshimaru, T.; Tomura, K. [Power Reactor and Nuclear Fuels Development Corporation, Ibaraki-ken (Japan)

    1995-12-31

    MONJU is a prototype fast breeder reactor in Japan designed to have a 280-MW(electric) output. The Power Reactor and Nuclear Fuel Development Corporation (PNC) started its construction in the autumn of 1985 in Tsuruga. The loading of the core fuel assemblies was started in October 1993, and the preoperational test is ongoing. MONJU uses 198 mixed-oxide (MOX) fuel assemblies as core fuel and 172 depleted uranium assemblies as blanket fuel. Assemblies loaded in-core and stored in the ex-vessel storage tank (EVST) reside in liquid sodium. These plutonium-containing fuel assemblies, MOX, and irradiated depleted uranium are regarded as in the difficult-to-access area, and the flows of fuel assemblies into and out of the area must be verified. Flow is verified by fuel flow monitors measuring radiation, which can limit inspector attendance during fuel handling.

  4. Tritium system design studies of fusion experimental breeder

    International Nuclear Information System (INIS)

    A summary of the tritium system design studies for the engineering outline design of a fusion experimental breeder (FEB-E) is presented. This paper is divided into three sections. In first section, the geometry, loading features and tritium concentrations in liquid lithium of tritium breeding zones of blanket are described. The tritium flow chart corresponding to the tritium fuel cycle system has been constructed, and the inventories in ten subsystems are calculated using SWITRIM code in section 2. Results show that the necessary initial tritium storage to start up FEB-E with fusion power of 143 MW is about 319 g. In final section, the tritium leakage issues under different operation circumstances have been analyzed. It was found that the potential danger of tritium leakage could be resulted from the exhausted gas of the diverter system. It is important to elevate the tritium burnup fraction and reduce the tritium throughput. (authors)

  5. Fast breeder reactor research

    International Nuclear Information System (INIS)

    , Italy, in April or May 1977. Recognizing the importance of international co-ope ration within the framework of IWGFR for preparing surveys, proposals and recommendations concerning sodium cooled fast breeder reactors, the Working Group prepared a number of joint documents with the help of experts from the participating countries, discussed them at the Eighth Annual Meeting and made recommendations on the preparation of subsequent joint documents. (author)

  6. Technical evaluation of major candidate blanket systems for fusion power reactor

    International Nuclear Information System (INIS)

    The key functions required for tritium breeding blankets for a fusion power reactor are: (1) self-sufficient tritium breeding, (2) in-situ tritium recovery and low tritium inventory, (3) high temperature cooling giving a high efficiency of electricity generation and (4) thermo-mechanical reliability and simplified remote maintenance to obtain high plant availability. Blanket performance is substantially governed by materials selection. Major options of structure/breeder/coolant/neutron multiplier materials considered for the present design study are PCA/Li2O/H2O/Be, Mo-alloy/Li2O/He/Be, Mo-alloy/LiAlO2/He/Be, V-alloy/Li/Li/none, and Mo-alloy/Li/He/none. In addition, remote maintenance of blankets, tritium recovery system, heat transport and energy conversion have been investigated. In this report, technological problems and critical R and D issues for power reactor blanket development are identified and a comparison of major candidate blanket concepts is discussed in terms of the present materials data base, economic performance, prospects for future improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies. (author)

  7. Fusion Blanket Coolant Section Criteria, Methodology, and Results

    Energy Technology Data Exchange (ETDEWEB)

    DeMuth, J. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meier, W. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jolodosky, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frantoni, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reyes, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-02

    The focus of this LDRD was to explore potential Li alloys that would meet the tritium breeding and blanket cooling requirements but with reduced chemical reactivity, while maintaining the other attractive features of pure Li breeder/coolant. In other fusion approaches (magnetic fusion energy or MFE), 17Li- 83Pb alloy is used leveraging Pb’s ability to maintain high TBR while lowering the levels of lithium in the system. Unfortunately this alloy has a number of potential draw-backs. Due to the high Pb content, this alloy suffers from very high average density, low tritium solubility, low system energy, and produces undesirable activation products in particular polonium. The criteria considered in the selection of a tritium breeding alloy are described in the following section.

  8. The current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

    The available thermodynamic information is reviewed for three categories of materials that meet essential criteria for use as breeding blankets in D-T fuelled fusion reactors: liquid lithium, solid lithium alloys, and lithium-containing ceramics. The leading candidate, liquid lithium, which also has potential for use as a coolant, has been studied more extensively than have the solid alloys or ceramics. Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries, etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phases of the Li-LiH, Li-LiD and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li3N, Li2O, and Li2C2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g. Li-Al and Li-Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li-M alloys can be estimated from lithium activity data for these alloys. There is essentially no refined thermodynamic information on the prospective ceramic blanket materials. The kinetics of tritium release from these materials is briefly discussed. Research areas are pointed out where additional thermodynamic information is needed for all three material categories. (author)

  9. Preliminary test for reprocessing technology development of tritium breeders

    Energy Technology Data Exchange (ETDEWEB)

    Hoshino, Tsuyoshi; Tsuchiya, Kunihiko; Hayashi, Kimio [Blanket Irradiation and Analysis Group, Directorates of Fusion Energy Research, Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Higashi Ibaraki-gun, Ibaraki 311-1393 (Japan); Nakamura, Mutsumi; Terunuma, Hitoshi [KAKEN Co., Ltd., 1044, Hori, Mito-city, Ibaraki 310-0903 (Japan); Tatenuma, Katsuyoshi [KAKEN Co., Ltd., 1044, Hori, Mito-city, Ibaraki 310-0903 (Japan)], E-mail: tatenuma@kakenlabo.co.jp

    2009-04-30

    In order to develop the reprocessing technology of lithium ceramics (Li{sub 2}TiO{sub 3}, CaO-doped Li{sub 2}TiO{sub 3}, Li{sub 4}SiO{sub 4} and Li{sub 2}O) as tritium breeder materials for fusion reactors, the dissolution methods of lithium ceramics to recover {sup 6}Li resource and the purification method of their lithium solutions to remove irradiated impurities ({sup 60}Co) were investigated. In the present work, the dissolving rates of lithium from each lithium ceramic powder using chemical aqueous reagents such as HNO{sub 3}, H{sub 2}O{sub 2} and citric acid (C{sub 6}H{sub 8}O{sub 7} . H{sub 2}O) were higher than 90%. Further the decontamination rate of {sup 60}Co added into the solutions dissolving lithium ceramics was higher than 97% using the activated carbon impregnated with 8-hydroxyquinolinol as chelate agent.

  10. Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

    International Nuclear Information System (INIS)

    A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the 210Po and 203Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the ITER Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket

  11. The impact of new experimental data on the design of Pb-17Li/water breeding blankets

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G. (Commission of the European Communities, Ispra (Italy). Joint Research Centre)

    1989-04-01

    The Pb-17Li/water-cooled blanket is one of the concepts being developed in Europe for testing in NET (Next European Torus). JRC-Ispra is strongly involved in this development. This paper describes the impact of the latest experimental results on the blanket design. The points considered are: breeder operating temperature and thermomechanical design: experiments on corrosion with steel 316L and liquid metal embrittlement tests have provided upper and lower limits for the breeder operating temperature (280-400/sup 0/C); tritium recovery from the breeder and permeation rate to the coolant: Ispra measurements indicate that solubility and diffusivity of hydrogen in Pb-17Li are lower as compared with the previous values used in blanket tritium analyses. The impact of these results on the design of the tritium recovery system is discussed; accident analyses: the experiments in progress at Ispra on the Pb-17Li/water interaction are reviewed and their application to a coolant pipe break accident is shown. (orig.).

  12. Research and development on liquid Pb-17Li breeder in europe

    Science.gov (United States)

    Casini, G.; Sannier, J.

    1991-03-01

    Research on eutectic Pb-17Li is part of the blanket studies carried-out in the frame of the EC-Fusion Technology Programme. Two blanket concepts using liquid Pb-17Li as breeder, one water-cooled and the other self-cooled, are being investigated and are among the candidates for testing in the Next Step machines. After a brief recall of the main features of both concepts, the paper presents the progress on the Pb-17Li data base acquisition, namely: — thermophysical properties, — solubility of metallic and non metallic elements (with a special attention to tritium), — chemical reactivity, — corrosion of structural materials and related mechanical effects, — tritium production and recovery.

  13. Research and development on liquid Pb-17Li breeder in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Sannier, J. (CEA Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France). Inst. de Recherche Technologique et de Developpement Industriel (IRDI))

    Research on eutectic Pb-17Li is part of the blanket studies carried-out in the frame of the EC-Fusion Technology Programme. Two blanket concepts using liquid Pb-17Li as breeder, one water-cooled and the other self-cooled, are being investigated and are among the candidates for testing in the Next Step machines. After a brief recall of the main features of both concepts, the paper presents the progress on the Pb-17Li data base acquisition, namely: - thermophysical properties, - solubility of metallic and non metallic elements (with a special attention to tritium), - chemical reactivity, - corrosion of structural materials and related mechanical effects, - tritium production and recovery. (orig.).

  14. Status and prospects of thermal breeders

    International Nuclear Information System (INIS)

    The main objective of this cooperative study and of this report is to evaluate the extent to which thermal breeders might complement or serve as an alternative to fast breeders in solving the long-term nuclear fuel supply problem. A secondary objective is to consider in a general way issues such as proliferation, safety, environmental impacts, economics, power plant availability, and fuel cycle versatility to determine whether thermal breeder reactors offer advantages or disadvantages with respect to such issues

  15. Analysis on tritium management in FLiBe blanket for LHD-type helical reactor FFHR2

    International Nuclear Information System (INIS)

    In FFHR2 (LHD-type helical reactor) design, FLiBe has been selected as a self-cooling tritium breeder for low reactivity with oxygen and water and lower conductivity. Considering the fugacity of the tritium, particular care and adequate mitigation measures should be applied for the effectively extracting tritium from breeder and controlling the tritium release to the environment. In this paper, a tritium analysis model of the FLiBe blanket system was developed and the preliminary analysis on tritium permeation and extraction for FLiBe blanket system were done. The results of the analysis showed that it was reasonable to select W alloy as heat exchanger (HX) material, the proportion of FLiBe flow in tritium recover system (TRS) was 0.2, the efficiency of TRS was 0.85 and tritium permeation reduction factor (TPRF) was 20 in blanket etc.. In addition, further R and D efforts were required for FFHR2 tritium system to guarantee the tritium self-sufficient and safety, for example reasonable quality of tritium permeation barriers on blanket, requirement for the TRS and fabrication technology of the heat exchanger etc.. (author)

  16. Analysis on tritium management in FLiBe blanket for force-free helical reactor FFHR2

    International Nuclear Information System (INIS)

    In FFHR2 design, FLiBe has been selected as a self-cooling tritium breeder for low reactivity with oxygen and water and lower conductivity. Considering the fugacity of the tritium, particular care and adequate mitigation measures should be applied for the effectively extract tritium from breeder and control the tritium release to the environment. In this paper, a tritium analysis model of the FLiBe blanket system was developed and the preliminary analysis on tritium permeation and extraction for FLiBe blanket system were done. The factors which affected tritium extraction and permeation were calculated and evaluated, such as the heat exchanger material, tritium permeation reduction factor (TPRF) in blanket, proportion of FLiBe flow in tritium recover system (TRS) and efficiency of TRS etc. The results of the analysis showed that further R and D efforts were required for FFHR2 tritium system to guarantee the tritium self-sufficient and safety, for example reasonable quality of tritium permeation barriers on blanket, requirement for the TRS and fabrication technology of the heat exchanger etc.. (author)

  17. Design study of blanket structure for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

  18. Improvement of hydrogen isotope exchange reactions on Li4SiO4 ceramic pebble by catalytic metals

    Institute of Scientific and Technical Information of China (English)

    Cheng Jian Xiao; Chun Mei Kang; Xiao Jun Chen; Xiao Ling Gao; Yang Ming Luo; Sheng Hu; Xiao Lin Wang

    2012-01-01

    Li4SiO4 ceramic pebble is considered as a candidate tritium breeding material of Chinese Helium Cooled Solid Breeder Test Blanket Module (CH HCSB TBM) for the International Thermonuclear Experimental Reactor (ITER).In this paper,Li4SiO4 ceramic pebbles deposited with catalytic metals,including Pt,Pd,Ru and Ir,were prepared by wet impregnation method.The metal particles on Li4SiO4 pebble exhibit a good promotion of hydrogen isotope exchange reactions in H2-DzO gas system,with conversion equilibrium temperature reduction of 200-300 ℃.The out-of-pile tritium release experiments were performed using 1.0 wt% Pt/Li4SiO4 and Li4SiO4 pebbles irradiated in a thermal neutron reactor.The thermal desorption spectroscopy shows that Pt was effective to increase the tritium release rate at lower temperatures,and the ratio of tritium molecule (HT) to tritiated water (HTO) of 1.0 wt% Pt/Li4SiO4 was much more than that of Li4SiO4,which released mainly as HTO.Thus,catalytic metals deposited on Li4SiO4 pebble may help to accelerate the recovery of bred tritium particularly in low temperature region,and increase the tritium molecule form released from the tritium breeding materials.

  19. Breeder Reactors, Understanding the Atom Series.

    Science.gov (United States)

    Mitchell, Walter, III; Turner, Stanley E.

    The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

  20. Fast breeder reactor protection system

    Science.gov (United States)

    van Erp, J.B.

    1973-10-01

    Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

  1. Fusion reactor blanket/shield design study

    International Nuclear Information System (INIS)

    A joint study of tokamak reactor first-wall/blanket/shield technology was conducted by Argonne National Laboratory (ANL) and McDonnell Douglas Astronautics Company (MDAC). The objectives of this program were the identification of key technological limitations for various tritium-breeding-blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium-breeding-blanket concepts were evaluated according to the proposed coolant. The ANL effort concentrated on evaluation of lithium- and water-cooled blanket designs while the MDAC effort focused on helium- and molten salt-cooled designs. A joint effort was undertaken to provide a consistent set of materials property data used for analysis of all blanket concepts. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first-wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented

  2. Breeder nutrition and offspring performance

    Directory of Open Access Journals (Sweden)

    F Calini

    2007-06-01

    Full Text Available Vertical integration in poultry industry strongly emphasizes the importance of cost control at all levels. In the usual broiler production operations, the costs involved with the production of the hatching egg or the day old chick are negligible if seen in the perspective of the cost per kg of live bird. From a research point of view, anyway, the greatest attention is usually given to the performance of broiler breeders, and most of the research in the field is focused on the improvement of their relative performance, mainly in terms of saleable chicks produced per hen, while less attention has been given to the quality of the chick and to the improvement of its growth performances, even if these last parameters have an effective impact on the overall economics of the poultry growing business. Most of the data available is quite dated, as can be seen from some recent reviews, and in general little attention is given to the impact of parental nutrition on the subsequent broiler performance. It is in fact more usual to find data about dam nutrition influence on egg fertility and hatchability than on subsequent progeny performance. The objectives of this review were to assess, on the basis of published reports, the effects of selected nutrients and anti-nutrients normally prevailing in commercial broiler breeder feeds - vitamins, micro-minerals, mycotoxins, - trying to pinpoint which could be the positive and the negative effects of both on the subsequent broiler performance, with a particular attention to the impact on immune function and carcass yield.

  3. Caracterização do resíduo de manta cerâmica usada para isolamento térmico e sua utilização na fabricação de argamassa Characterization of the waste of ceramic blanket used for thermal insulation and its use in the manufacture of mortar

    Directory of Open Access Journals (Sweden)

    Luiz Alberto Baptista Pinto Junior

    2010-09-01

    Full Text Available Esse trabalho tem os objetivos de caracterizar o resíduo de manta cerâmica, gerado no lingotamento contínuo dos aços, devido à diminuição da sua capacidade de isolamento térmico, e de estudar a viabilidade técnica da utilização desse resíduo na fabricação de argamassa. O resíduo foi submetido à análise química, difração de raios X e MEV e utilizado na preparação de argamassa. Foram feitas dosagens de 0%, 2%, e 6% em massa do resíduo, em substituição ao cimento, e realizado o ensaio de índice de consistência e compressão na argamassa. Dessa maneira, obtiveram-se resultados em três argamassas distintas. Os resultados indicam um aumento na resistência à compressão com adição do resíduo, em comparação com a argamassa-padrão, e o índice de consistência apresentou alterações pouco consideráveis.The goal of this study is to characterize the waste of ceramic blankets used in the thermal insulation of the continuous casting machine, which was generated due to the decrease in the capacity of thermal insulation. It also investigated the technical feasibility of the use of this waste as raw material to produce mortar. The waste was characterized by chemical analysis, X-ray diffraction, and SEM. Proportions of 0%, 2%, and 6% of the mass of the waste were used in substitution of cement in the production of mortar samples. Samples were tested to obtain the consistency index and mortar compression tests. The results indicate an increase in the compression resistance with the addition of waste in comparison with standard mortar. The consistency index was not affected by the additions.

  4. Comparison of early socialization practices used for litters of small-scale registered dog breeders and nonregistered dog breeders.

    Science.gov (United States)

    Korbelik, Juraj; Rand, Jacquie S; Morton, John M

    2011-10-15

    OBJECTIVE-To compare early socialization practices between litters of breeders registered with the Canine Control Council (CCC) and litters of nonregistered breeders advertising puppies for sale in a local newspaper. DESIGN-Retrospective cohort study. Animals-80 litters of purebred and mixed-breed dogs from registered (n = 40) and non-registered (40) breeders. PROCEDURES-Registered breeders were randomly selected from the CCC website, and nonregistered breeders were randomly selected from a weekly advertising newspaper. The litter sold most recently by each breeder was then enrolled in the study. Information pertaining to socialization practices for each litter was obtained through a questionnaire administered over the telephone. RESULTS-Registered breeders generally had more breeding bitches and had more litters than did nonregistered breeders. Litters of registered breeders were more likely to have been socialized with adult dogs, people of different appearances, and various environmental stimuli, compared with litters of nonregistered breeders. Litters from registered breeders were also much less likely to have been the result of an unplanned pregnancy. CONCLUSIONS AND CLINICAL RELEVANCE-Among those breeders represented, litters of registered breeders received more socialization experience, compared with litters of nonregistered breeders. People purchasing puppies from nonregistered breeders should focus on socializing their puppies between the time of purchase and 14 weeks of age. Additional research is required to determine whether puppies from nonregistered breeders are at increased risk of behavioral problems and are therefore more likely to be relinquished to animal shelters or euthanized, relative to puppies from registered breeders. PMID:21985351

  5. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    A mechanical configuration of ITER integrated primary first wall/shield blanket module were developed focusing on the welded attachment of its support leg to the back plate. A 100 mm x 150 mm space between the legs of adjacent modules was incorporated for the working space of welding/cutting tools. A concept of coolant branch pipe connection to accommodate deformation due to the leg welding and differential displacement of the module and the manifold/back plate during operation was introduced. Two-dimensional FEM analyses showed that thermal stresses in Cu-alloy (first wall) and stainless steel (first wall coolant tube and shield block) satisfied the stress criteria following ASME code for ITER BPP operation. On the other hand, three-dimensional FEM analyses for overall in-vessel structures exhibited excessive primary stresses in the back plate and its support structure to the vacuum vessel under VDE disruption load and marginal stresses in the support leg of module No.4. Fabrication procedure of the integrated primary first wall/shield blanket module was developed based on single step solid HIP for the joining of Cu-alloy/Cu-alloy, Cu-alloy/stainless steel, and stainless steel/stainless steel. (author)

  6. ITER blanket, shield and material data base

    International Nuclear Information System (INIS)

    As part of the summary of the Conceptual Design Activities (CDA) for the International Thermonuclear Experimental Reactor (ITER), this document describes the ITER blanket, shield, and material data base. Part A, ''ITER Blanket and Shield Conceptual Design'', discusses the need for ITER of a tritium breeding blanket to supply most of the tritium for the fuel cycle of the device. Blanket and shield combined must be designed to operate at a neutron wall loading of 1MW/m2, and to provide adequate shielding of the magnets to meet the neutron energy fluence goal of 3MWa/m2 at the first wall. After a summary of the conceptual design, the following topics are elaborated upon: (1) function, design requirement, and critical issues; (2) material selection; (3) blanket and shield segmentation; (4) blanket design description; (5) design analysis; (6) shield; (7) radiation streaming analysis; and (8) a summary of benchmark calculations. Part B, ''ITER Materials Evaluation and Data Base'', treats the compilation and assessment of the available materials data base used for the selection of the appropriate materials for all major components of ITER, including (i) structural materials for the first wall, (ii) Tritium breeding materials for the blanket, (iii) plasma facing materials for the divertor and first wall armor, and (4) electric insulators for use in the blanket and divertor. Refs, figs and tabs

  7. Li depletion effects on Li2TiO3 reaction with H2 in thermo-chemical environment relevant to breeding blanket for fusion power plants

    International Nuclear Information System (INIS)

    This is a report of the Working Group in the Subtask on Solid Breeder Blankets under the Implementing Agreement on a Co-operative Programme on Nuclear Technology of Fusion Reactors (International Energy Agency (IEA)). This Working Group (Task F and WG-F) was performed from 2000 to 2004 by a collaboration of European Union (EU) and Japan (JA). In this report, lithium depletion effects on the reaction of lithium titanate (Li2TiO3) with hydrogen (H2) in thermo-chemical environment were discussed. The reaction of Li2TiO3 ceramics with H2 was studied in a thermo-chemical environment simulating (excepting irradiation) that of the hottest pebble-bed zone of breeding-blanket actually designed for fusion power plants. This 'reduction' as performed at 900degC in Ar+0.1%H, purge gas (He+0.1%H2 being the designed reference') was found to be enhanced by TiO2 doping of the specimens of simulate 6Li-burn-up expected to reach 20% at their end-of-life. The reaction rates, however, were so slow to be not significantly extrapolated to the breeder material service time (years). In Ar+3%H2, faster reaction rates allowed a better identification of the process evolution (kinetics) by Temperature-Programmed Reduction' (TPR) and 'Oxidation' (TPO), and combined TG-DTA thermal analysis. The reduction of pure Li4/5TiO12/5 spinel phase to Li4/5TiO12/5-y was found to reach in one day the steady state at the O-vacancy concentration y=0.2. Complimentary microscopy (SEM) and spectroscopy (XRD, XPS) techniques were used to characterize the reaction products among which the presence of the orthorhombic LivTiO2 (0 ≤ v ≤ 1/2) and Li2TiO3 could be diagnosed. So that the complete spinel reduction to Li1/2TiO2 was obtained according to a scheme involving the Li1/2TiO2-Li4/5TiO12/5 spinel phase solid solution for which y=3v/(10-5v). The reduction rate of pure meta-titanate to Li2TiO3-x was found much lower (x approx. = 0.01) and even possibly due to the presence of the spinel phase whose quantitative

  8. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

    Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

  9. Portfolio: Ceramics.

    Science.gov (United States)

    Hardy, Jane; And Others

    1982-01-01

    Describes eight art activities using ceramics. Elementary students created ceramic tiles to depict ancient Egyptian and medieval European art, made ceramic cookie stamps, traced bisque plates on sketch paper, constructed clay room-tableaus, and designed clay relief masks. Secondary students pit-fired ceramic pots and designed ceramic Victorian…

  10. Multivariable optimization of fusion reactor blankets

    International Nuclear Information System (INIS)

    The optimization problem consists of four key elements: a figure of merit for the reactor, a technique for estimating the neutronic performance of the blanket as a function of the design variables, constraints on the design variables and neutronic performance, and a method for optimizing the figure of merit subject to the constraints. The first reactor concept investigated uses a liquid lithium blanket for breeding tritium and a steel blanket to increase the fusion energy multiplication factor. The capital cost per unit of net electric power produced is minimized subject to constraints on the tritium breeding ratio and radiation damage rate. The optimal design has a 91-cm-thick lithium blanket denatured to 0.1% 6Li. The second reactor concept investigated uses a BeO neutron multiplier and a LiAlO2 breeding blanket. The total blanket thickness is minimized subject to constraints on the tritium breeding ratio, the total neutron leakage, and the heat generation rate in aluminum support tendons. The optimal design consists of a 4.2-cm-thick BeO multiplier and 42-cm-thick LiAlO2 breeding blanket enriched to 34% 6Li

  11. Mechanical and thermal design of hybrid blankets

    International Nuclear Information System (INIS)

    The thermal and mechanical aspects of hybrid reactor blanket design considerations are discussed. This paper is intended as a companion to that of J. D. Lee of Lawrence Livermore Laboratory on the nuclear aspects of hybrid reactor blanket design. The major design characteristics of hybrid reactor blankets are discussed with emphasis on the areas of difference between hybrid reactors and standard fusion or fission reactors. Specific examples are used to illustrate the design tradeoffs and choices that must be made in hybrid reactor design. These examples are drawn from the work on the Mirror Hybrid Reactor

  12. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li17Pb83 and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the tritium breeding ratio to group structure and weighting spectrum increases as the thickness and Li enrichment decrease with up to 20% discrepancies for thin natural Li17Pb83 blankets. (author)

  13. Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Nygren, R.E.; Baxi, C.B.; Fogarty, P.; Ghoniem, N.; Khater, H.; McCarthy, K.; Merrill, B.; Nelson, B.; Reis, E.E.; Sharafat, S.; Schleicher, R.; Sze, D.K.; Ulrickson, M.; Willms, S.; Youssef, M.; Zinkel, S.

    1999-08-01

    Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W-5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. System study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kWh. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study.

  14. Divertor and gas blanket impurity control study

    Energy Technology Data Exchange (ETDEWEB)

    El Derini, Z; Stacey, Jr, W M

    1979-04-01

    A simple calculational model for the transport of particles across the scrap off region between the plasma and the wall in the presence of a divertor or a gas blanket has been developed. The model departs from previous work in including: (a) the entire impurity transport as well as its effect on the energy balance equations; (b) the recycling neutrals from the divertor, and (c) the reflected neutrals from the wall. Results obtained with this model show how the steady state impurity level in the plasma depends on the divertor parameters such as the neutral backflow from the divertor, the particle residence time and the scrape off thickness; and on the gas blanket parameters such as the neutral source strength and the gas blanket thickness. The variation of the divertor or gas blanket performance as a function of the heat and particle fluxes escaping from the plasma, the wall material and the cross field diffusion is examined and numerical examples are given.

  15. Exploratory Study of Blanket Liquid Curtain

    Institute of Scientific and Technical Information of China (English)

    HUGang; HUANGJinhua; FENGKaiming

    2003-01-01

    Blankets and other in-vessel components are easily damaged owing to their circumstance of high radiation and high heat. To protect them, first wall design should be considered. Owing to its high heat removal nd self-refreshing capability, liquid metal first wall has been seen as a potential first wall for a fusion reactor in the future. Blanketliquid curtain is actually a special liquid metal wall to protect blanket.

  16. Accelerator breeder with uranium, thorium target

    International Nuclear Information System (INIS)

    An accelerator breeder, that uses a low-enriched fuel as the target material, can produce substantial amounts of fissile material and electric power. A study of H2O- and D2O-cooled, UO2, U, (depleted U), or thorium indicates that U-metal fuel produces a good fissile production rate and electrical power of about 60% higher than UO2 fuel. Thorium fuel has the same order of magnitude as UO2 fuel for fissile-fuel production, but the generating electric power is substantially lower than in a UO2 reactor. Enriched UO2 fuel increases the generating electric power but not the fissile-material production rate. The Na-cooled breeder target has many advantages over the H2O-cooled breeder target

  17. Improved fuel element for fast breeder reactor

    International Nuclear Information System (INIS)

    The invention, in which the United States Department of Energy has participated as co-inventor, relates to breeder reactor fuel elements, and specifically to such elements incorporating 'getters', hereafter designated as fission product traps. The main object of the invention is the construction of a fast breeder reactor fuel pin, free from local stresses induced in the cladding by reactions with cesium. According to the invention, the fast breeder fuel element includes a cladding tube, sealed at both ends by a plug, and containing a fissile stack and a fertile stack, characterized by the interposition of a cesium trap between the fissile and fertile stacks. The trap is effective at reactor operating temperatures in retaining and separating the cesium generated in the fissile material and preventing cesium reaction with the fertile stack. Depending on the construction method adopted, the trap may consists of a low density titanium oxide or niobium oxide pellet

  18. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  19. Coatings for fast breeder reactor components

    International Nuclear Information System (INIS)

    Several types of metallurgical coatings are used in the unique environments of the fast breeder reactor. Most of the coatings have been developed for tribological applications, but some also serve as corrosion barriers, diffusion barriers, or radionuclide traps. The materials that have consistently given the best performance as tribological coatings in the breeder reactor environments have been coatings based on chromium carbide, nickel aluminide, or Tribaloy 700 (a nickel-base hard-facing alloy). Other coatings that have been qualified for limited applications include chromium plating for low temperature galling protection and nickel plating for radionuclide trapping

  20. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  1. Two dimensional distribution of tritium breeding ratio and induced activity in Japanese water cooled and helium cooled test blanket modules

    International Nuclear Information System (INIS)

    Solid breeder blankets are regarded as a near-at-hand blanket concept for a fusion power demonstration plant in Japan. Test blanket module (TBM) to be tested in ITER is the most important milestone to establish the fusion demonstration blanket. For the candidate TBM's, two types of TBM, water cooled solid breeder TBM, and a helium gas cooled solid breeder TBM have been proposed and designed in JAERI. For detailed performance study under operation and after shut down, detailed neutronics analysis gives the most important design conditions, such as, distribution of tritium breeding ratio, nuclear heating rate during operation, and induced activation and decay heat after termination of irradiation. In the analysis, neutron and gamma transportation was calculated by two dimensional analysis code, DOT3.5, for two TBMs. Nuclear reaction rate and induced activation rate were evaluated by APPLE-3 and ACT-4, respectively. The analysis model included configurations of thermo-mechanical test modules and surrounding common frames for both of He cooled and water cooled TBMs. By the neutronics analysis, TBR and contact dose rate by induced activation till one year after termination of the module testing have been evaluated. For the evaluation of induced activation level change and decay heat change, the transient decreases in one year after termination of the module testing have been calculated. The time duration of the module testing before termination of testing is assumed to be 133 continuous days of full power operation. The result of TBR analysis showed that TBR distribution in the toroidal direction of TBM is not significant, however, the neutron flux decreases in the region of sidewall of common frame made of SS and water. This result shows that there is relatively large neutron loss from the TBM to the common frame. Thus, it is considered that the TBR value observed in the TBM testing may be smaller than the estimation by one dimensional neutronics analysis which does

  2. Status and prospects of advanced fissile fuel breeders

    International Nuclear Information System (INIS)

    Fusion--fission hybrid systems, fast breeder systems, and accelerator breeder systems were compared on a common basis using a simple economic model. Electricity prices based on system capital costs only were computed, and were plotted as functions of five key breeder system parameters. Nominally, hybrid system electricity costs were about twenty-five percent lower than fast breeder system electricity costs, and fast breeder system electricity costs were about forty percent lower than accelerator breeder system electricity costs. In addition, hybrid system electricity costs were very insensitive to key parameter variations on the average, fast breeder system electricity costs were moderately sensitive to key parameter variations on the average, and accelerator breeder system electricity costs were the most sensitive to key parameter variations on the average

  3. Status of liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    This document represents a compilation of the information on the status of fast breeder reactor development. It is intended to provide complete and authoritative information for academic, energy, industrial and planning organizations in the IAEA Member States. The Report also provides extended reference and bibliography lists. A summarized overview of the national programmes of LMFBR development is given in Chapter II. Chapter III on LMFBR experience provides a brief description and purpose of all fast reactors - experimental, demonstration and commercial size - that have been or are planned for construction and operation. Fast reactor physics is dealt with in Chapter IV. Besides the basic facts and definitions of neutronics and the compilation and measurement of nuclear data, a broad range of the calculation methods, codes, and the state of the art is described. In Chapter V, fuels and materials are described. The emphasis is on the design and development experience gained with mixed oxide fuel pins and subassemblies. Structural materials, blanket elements and absorber materials are also discussed. Chaper VI presents a broad overview of the technical and engineering aspects of LMFBR power plants. LMFBR core design is described in detail, followed by the components of the main heat transport system, the refuelling equipment, and auxiliary systems. Chapter VII on safety is a compilation of the current safety design concepts of LMFBRs and new trends in safety criteria and safety goals. The chapter concludes with risk analyses of LMFBR technology. In Chapter VIII, the systems approach has been emphasized in the consideration of the whole LMFBR fuel cycle. Special emphasis is placed on safeguards aspects and the environmental impact of the LMFBR fuel cycle. Chapter IX describes deployment considerations of LMFBRs. Special emphasis is placed on economic aspects of the LMFBR power plant and its related fuel cycle. Finally, Chapter X provides an overall summary and a

  4. Compatibility of sodium with ceramic oxides employed in nuclear reactors

    International Nuclear Information System (INIS)

    This work is a review of experiments carried out up to the present time on the corrosion and compatibility of ceramic oxides with liquid sodium at temperatures corresponding to those in fast breeder reactors. The review also includes the results of a thermo-dynamic/liquid sodium reactions. The exercise has been conducted with a view to effecting experimental studies in the future. (Author)

  5. Possible types of breeders with thorium cycle

    International Nuclear Information System (INIS)

    Neutronics calculations of simplified homogeneous reactor models show the possibility that metal-fueled LMFBRs and coated particle fueled gas cooled reactors achieve reactor doubling times of around 10 years with the thorium cycle. Three concepts of gas-cooled thorium cycle breeders are discused. (Author)

  6. Possible types of breeders with thorium cycle

    International Nuclear Information System (INIS)

    Neutronics calculations of simplified homogeneous reactor models show the possibility that metal-fueled LMFBRs and coated particle fueled gas cooled reactors achieve doubling times of around 10 years with the thorium cycle. Three concepts of gas-cooled thorium cycle breeders are discussed. (Author)

  7. Tritium transport in lithium ceramics porous media

    Energy Technology Data Exchange (ETDEWEB)

    Tam, S.W.; Ambrose, V.

    1991-12-31

    A random network model has been utilized to analyze the problem of tritium percolation through porous Li ceramic breeders. Local transport in each pore channel is described by a set of convection-diffusion-reaction equations. Long range transport is described by a matrix technique. The heterogeneous structure of the porous medium is accounted for via Monte Carlo methods. The model was then applied to an analysis of the relative contribution of diffusion and convective flow to tritium transport in porous lithium ceramics. 15 refs., 4 figs.

  8. Tritium transport in lithium ceramics porous media

    Energy Technology Data Exchange (ETDEWEB)

    Tam, S.W.; Ambrose, V.

    1991-01-01

    A random network model has been utilized to analyze the problem of tritium percolation through porous Li ceramic breeders. Local transport in each pore channel is described by a set of convection-diffusion-reaction equations. Long range transport is described by a matrix technique. The heterogeneous structure of the porous medium is accounted for via Monte Carlo methods. The model was then applied to an analysis of the relative contribution of diffusion and convective flow to tritium transport in porous lithium ceramics. 15 refs., 4 figs.

  9. Design and trial fabrication of a dismantling apparatus for irradiation capsules of solid tritium breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, K. [Japan Atomic Energy Agency, Blanket Irradiation and Analysis Group, Fusion Research and Development Directorate, 4002 Narita-cho, Oarai-machi, Ibaraki-ken 311-1393 (Japan)], E-mail: hayashi.kimio@jaea.go.jp; Nakagawa, T.; Onose, S.; Ishida, T.; Nakamichi, M. [Japan Atomic Energy Agency, Blanket Irradiation and Analysis Group, Fusion Research and Development Directorate, 4002 Narita-cho, Oarai-machi, Ibaraki-ken 311-1393 (Japan); Takatsu, H. [Fusion Energy and Development Directorate, Japan Atomic Energy Agency, 801-1 Mukouyama, Naka-shi, Ibaraki-ken 311-0193 (Japan); Nakamura, M.; Noguchi, T. [Kaken, Inc., 873-3 Shikada, Hokota-shi, Ibaraki-ken, 311-1416 (Japan)

    2009-04-30

    Irradiation experiments of solid breeder materials including Li{sub 2}TiO{sub 3} have been being carried out in preparation for a test blanket module (TBM) of the International Thermonuclear Experimental Reactor (ITER). The present paper deals with design and trial-fabrication works for developing a dismantling apparatus for the irradiation capsules. The dismantling process leads to release of tritium which is left in free volumes of the capsule or in the breeder specimens. In the design of the dismantling apparatus, the released tritium is recovered safely by a purge-gas system during the cutting of the irradiation capsule by a band saw, and then the tritium is consolidated into a radioactive waste. Furthermore, an inner-box enclosing the dismantling apparatus works as a countermeasure of possible release of tritium in accidental events. Good performance of a trial fabrication model of the dismantling apparatus has been demonstrated by preliminary cutting runs using some mockups simulating the irradiation capsules. Thus, the present design of the apparatus, together with the trial mock-up runs, will contribute to the design of the TBM structure and to the planning of the dismantling process of the TBM.

  10. Design and trial fabrication of a dismantling apparatus for irradiation capsules of solid tritium breeder materials

    International Nuclear Information System (INIS)

    Irradiation experiments of solid breeder materials including Li2TiO3 have been being carried out in preparation for a test blanket module (TBM) of the International Thermonuclear Experimental Reactor (ITER). The present paper deals with design and trial-fabrication works for developing a dismantling apparatus for the irradiation capsules. The dismantling process leads to release of tritium which is left in free volumes of the capsule or in the breeder specimens. In the design of the dismantling apparatus, the released tritium is recovered safely by a purge-gas system during the cutting of the irradiation capsule by a band saw, and then the tritium is consolidated into a radioactive waste. Furthermore, an inner-box enclosing the dismantling apparatus works as a countermeasure of possible release of tritium in accidental events. Good performance of a trial fabrication model of the dismantling apparatus has been demonstrated by preliminary cutting runs using some mockups simulating the irradiation capsules. Thus, the present design of the apparatus, together with the trial mock-up runs, will contribute to the design of the TBM structure and to the planning of the dismantling process of the TBM.

  11. Neutronic implications of lead-lithium blankets

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W.R.

    1982-08-01

    Lead-lithium alloys have been proposed for use in several conceptual blanket designs for both inertial and magnetic confinement fusion reactors. In most cases, Pb/sub 83/Li/sub 17/, a eutectic with a melting point of 235/sup 0/C, is the chosen composition. The primary reasons for using Pb/sub 83/Li/sub 17/ instead of Li as the tritium breeding material are the perceived safety advantages, low tritium solubility, and favorable neutronic characteristics. This paper describes the neutronic characteristics of Pb/sub 83/Li/sub 17/ blankets with emphasis on the enhanced neutron leakage through chamber ports and the degradation in blanket performance parameters that occurs as a result of the enhanced leakage.

  12. On Ceramics.

    Science.gov (United States)

    School Arts, 1982

    1982-01-01

    Presents four ceramics activities for secondary-level art classes. Included are directions for primitive kiln construction and glaze making. Two ceramics design activities are described in which students make bizarrely-shaped lidded jars, feet, and footwear. (AM)

  13. Some new ideas for Tandem Mirror blankets

    International Nuclear Information System (INIS)

    The Tandem Mirror Reactor, with its cylindrical central cell, has led to numerous blanket designs taking advantage of the simple geometry. Also many new applications for fusion neutrons are now being considered. To the pure fusion electricity producers and hybrids producing fissile fuel, we are adding studies of synthetic fuel producers and fission-suppressed hybrids. The three blanket concepts presented are new ideas and should be considered illustrative of the breadth of Livermore's application studies. They are not meant to imply fully analyzed designs

  14. Lightweight IMM PV Flexible Blanket Assembly

    Science.gov (United States)

    Spence, Brian

    2015-01-01

    Deployable Space Systems (DSS) has developed an inverted metamorphic multijunction (IMM) photovoltaic (PV) integrated modular blanket assembly (IMBA) that can be rolled or z-folded. This IMM PV IMBA technology enables a revolutionary flexible PV blanket assembly that provides high specific power, exceptional stowed packaging efficiency, and high-voltage operation capability. DSS's technology also accommodates standard third-generation triple junction (ZTJ) PV device technologies to provide significantly improved performance over the current state of the art. This SBIR project demonstrated prototype, flight-like IMM PV IMBA panel assemblies specifically developed, designed, and optimized for NASA's high-voltage solar array missions.

  15. Advanced Ceramics

    International Nuclear Information System (INIS)

    The First Florida-Brazil Seminar on Materials and the Second State Meeting about new materials in Rio de Janeiro State show the specific technical contribution in advanced ceramic sector. The others main topics discussed for the development of the country are the advanced ceramic programs the market, the national technic-scientific capacitation, the advanced ceramic patents, etc. (C.G.C.)

  16. Cooperative and concentrated breeder development in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Hueper, R.

    The agreement of 1984 on cooperation for the fast breeder development, concluded by West Germany and France, Great Britain, Belgium and Italy, created the basis for abandoning the 'autarky' of national development efforts, which since then have been combined into a joint demonstration project. This European Fast Reactor, EFR, is in the phase of preparatory planning and is intended to replace the originally planned three installations SNR-2, SPX-2, and CDFR. There still are financing problems to be solved, and the conditions of further participation of Italy (and the Netherlands) are awaiting final decisions. The joint European experience in breeder development relies on operating results of more than 12 power reactors in the world, and the SNR-300 is expected to contribute a wealth of new experience after its commissioning.

  17. FOWL CHOLERA IN A BREEDER FLOCK

    OpenAIRE

    Z. Parveen, A. A. Nasir, K.Tasneem and A. Shah

    2003-01-01

    During January, 2003 Pasteurella multocida the causative agent of fowl cholera was isolated from a breeder flock in Lahore District. The age of the flock was 245 days. Increased mortality, swollen wattles and lameness were the clinical findings present in almost all the affected birds, while gross lesions were typical of fowl cholera. To prove the virulence of the organism, mice and six-week old cockerals were infected and P. multocida was reisolated.

  18. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  19. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    International Nuclear Information System (INIS)

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option

  20. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giese, R.F.

    1984-04-01

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option.

  1. Neutronics Evaluation of Lithium-Based Ternary Alloys in IFE Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Jolodosky, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fratoni, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-11-20

    Pre-conceptual fusion blanket designs require research and development to reflect important proposed changes in the design of essential systems, and the new challenges they impose on related fuel cycle systems. One attractive feature of using liquid lithium as the breeder and coolant is that it has very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and presents plant safety concerns. If the chemical reactivity of lithium could be overcome, the result would have a profound impact on fusion energy and associated safety basis. The overriding goal of this project is to develop a lithium-based alloy that maintains beneficial properties of lithium (e.g. high tritium breeding and solubility) while reducing overall flammability concerns. To minimize the number of alloy combinations that must be explored, only those alloys that meet certain nuclear performance metrics will be considered for subsequent thermodynamic study. The specific scope of this study is to evaluate the neutronics performance of lithium-based alloys in the blanket of an inertial confinement fusion (ICF) engine. The results of this study will inform the development of lithium alloys that would guarantee acceptable neutronics performance while mitigating the chemical reactivity issues of pure lithium.

  2. Review: BNL graphite blanket design concepts

    International Nuclear Information System (INIS)

    A review of the Brookhaven National Laboratory (BNL) minimum activity graphite blanket designs is made. Three designs are identified and discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a thick graphite screen (typically 30 cm or greater, depending on type as well as application-experimental power reactor or commercial reactor). Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy. This energy is then either radiated to a secondary blanket with coolant tubes, as in types A and B, or is removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the structural material of the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude by the graphite screen, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma, whatever the degree of radiation damage

  3. 47 CFR 22.353 - Blanketing interference.

    Science.gov (United States)

    2010-10-01

    ... Operational and Technical Requirements Technical Requirements § 22.353 Blanketing interference. Licensees of...: ER17NO94.007 where d is the radial distance to the boundary, in kilometers p is the radial effective radiated power, in kilowatts The maximum effective radiated power in the pertinent direction,...

  4. Aerogel Blanket Insulation Materials for Cryogenic Applications

    Science.gov (United States)

    Coffman, B. E.; Fesmire, J. E.; White, S.; Gould, G.; Augustynowicz, S.

    2009-01-01

    Aerogel blanket materials for use in thermal insulation systems are now commercially available and implemented by industry. Prototype aerogel blanket materials were presented at the Cryogenic Engineering Conference in 1997 and by 2004 had progressed to full commercial production by Aspen Aerogels. Today, this new technology material is providing superior energy efficiencies and enabling new design approaches for more cost effective cryogenic systems. Aerogel processing technology and methods are continuing to improve, offering a tailor-able array of product formulations for many different thermal and environmental requirements. Many different varieties and combinations of aerogel blankets have been characterized using insulation test cryostats at the Cryogenics Test Laboratory of NASA Kennedy Space Center. Detailed thermal conductivity data for a select group of materials are presented for engineering use. Heat transfer evaluations for the entire vacuum pressure range, including ambient conditions, are given. Examples of current cryogenic applications of aerogel blanket insulation are also given. KEYWORDS: Cryogenic tanks, thermal insulation, composite materials, aerogel, thermal conductivity, liquid nitrogen boil-off

  5. Advanced Polymer For Multilayer Insulating Blankets

    Science.gov (United States)

    Haghighat, R. Ross; Shepp, Allan

    1996-01-01

    Polymer resisting degradation by monatomic oxygen undergoing commercial development under trade name "Aorimide" ("atomic-oxygen-resistant imidazole"). Intended for use in thermal blankets for spacecraft in low orbit, useful on Earth in outdoor applications in which sunlight and ozone degrades other plastics. Also used, for example, to make threads and to make films coated with metals for reflectivity.

  6. ITER driver blanket, European Community design

    Energy Technology Data Exchange (ETDEWEB)

    Simbolotti, G. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Zampaglione, V. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Ferrari, M. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Gallina, M. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Mazzone, G. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Nardi, C. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Petrizzi, L. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Rado, V. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Violante, V. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Daenner, W. (NET Team, Max-Planck-Inst. fuer Plasmaphysik, Garching (Germany)); Lorenzetto, P. (NET Team, Max-Planck-Inst. fuer Plasmaphysik, Garching (Germany)); Gierszewski, P. (CFFTP, Mississauga, ON (Canada)); Gratt

    1993-07-01

    Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly afftect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. With the framework of the ITER Conceptual Design Activities (CDA, 1988-1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper. (orig.)

  7. Fidget Blankets: A Sensory Stimulation Outreach Program.

    Science.gov (United States)

    Kroustos, Kelly Reilly; Trautwein, Heidi; Kerns, Rachel; Sobota, Kristen Finley

    2016-01-01

    Behavioral and Psychological Symptoms of Dementia (BPSD) include behaviors such as aberrant motor behavior, agitation, anxiety, apathy, delusions, depression, disinhibition, elation, hallucinations, irritability, and sleep or appetite changes. A student-led project to provide sensory stimulation in the form of "fidget blankets" developed into a community outreach program. The goal was to decrease the use of antipsychotics used for BPSD.

  8. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  9. Blanket design study for a Commercial Tokamak Hybrid Reactor (CTHR)

    International Nuclear Information System (INIS)

    The results are presented of a study on two blanket design concepts for application in a Commercial Tokamak Hybrid Reactor (CTHR). Both blankets operate on the U-Pu cycle and are designed to achieve tritium self-sufficiency while maximizing the fissile fuel production within thermal and mechanical design constraints. The two blanket concepts that were evaluated were: (1) a UC fueled, stainless steel clad and structure, helium cooled blanket; and (2) a UO2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li2O and LiH, were evaluated for use in both blanket concepts. The use of lead as a neutron multiplier or reflector and graphite as a reflector was also considered for both blankets

  10. What determines hatchling weight: breeder age or incubated egg weight?

    OpenAIRE

    AB Traldi; Menten JFM; CS Silva; PV Rizzo; PWZ Pereira; J Santarosa

    2011-01-01

    Two experiments were carried out to determine which factor influences weight at hatch of broiler chicks: breeder age or incubated egg weight. In Experiment 1, 2340 eggs produced by 29- and 55-week-old Ross® broiler breeders were incubated. The eggs selected for incubation weighed one standard deviation below and above average egg weight. In Experiment 2, 2160 eggs weighing 62 g produced by breeders of both ages were incubated. In both experiments, 50 additional eggs within the weight interval...

  11. Hatching distribution of eggs varying in weight and breeder age

    Directory of Open Access Journals (Sweden)

    SL Vieira

    2005-06-01

    Full Text Available Broiler chicks from one incubator hatch within long periods of time, which leads to dehydration and reduction in yolk sac reserves of those chicks that have hatched earlier and potentially impairs early performance. The present research investigated the hatching distribution at intervals of incubation using eggs of different weights within one breeder age or eggs from widely different breeder ages. Eggs from breeders at 27 and 59 weeks of age (54 and 69 g and from breeders at 40 weeks of age, which were graded as light (58 g and heavy (73 g, were placed in a commercial incubator. There were a total of 1,184 eggs distributed in four treatments and eight replicates: eggs from 27-week-old breeders (27B, eggs from 59-week-old breeders (59B, light eggs from 40-week-old breeders (40BL and heavy eggs from 40-week-old breeders (40BH. Replicates were comprised of 37 eggs that were placed in each incubator tray. The treatments were physically separated from each other using a plate. Eggs were transferred to a hatcher after 432 hours of incubation and the first chick hatched at 449 hours of incubation. Afterwards, the number of completely hatched chicks from each replicate was recorded at six-hour intervals until 503 hours of incubation, when the hatchings stopped. Hatched chicks were removed from the trays after each measurement. Data were submitted to an analysis of variance with repeated measures. There was a significant interaction between breeder age and incubation length. The hatching onset of eggs from the old breeders was later compared to young breeders. Hatchability (%incubated eggs was lower for the old breeders; however, differences in hatchability as a percentage of the hatched eggs were not so evident. Complete hatchability occurred only at 503 hours of incubation; however, more than 90% eggs had hatched 18 hours earlier.

  12. Ceramic joining

    Energy Technology Data Exchange (ETDEWEB)

    Loehman, R.E. [Sandia National Lab., Albuquerque, NM (United States)

    1996-04-01

    This paper describes the relation between reactions at ceramic-metal interfaces and the development of strong interfacial bonds in ceramic joining. Studies on a number of systems are described, including silicon nitrides, aluminium nitrides, mullite, and aluminium oxides. Joints can be weakened by stresses such as thermal expansion mismatch. Ceramic joining is used in a variety of applications such as solid oxide fuel cells.

  13. Ceramic Processing

    Energy Technology Data Exchange (ETDEWEB)

    EWSUK,KEVIN G.

    1999-11-24

    Ceramics represent a unique class of materials that are distinguished from common metals and plastics by their: (1) high hardness, stiffness, and good wear properties (i.e., abrasion resistance); (2) ability to withstand high temperatures (i.e., refractoriness); (3) chemical durability; and (4) electrical properties that allow them to be electrical insulators, semiconductors, or ionic conductors. Ceramics can be broken down into two general categories, traditional and advanced ceramics. Traditional ceramics include common household products such as clay pots, tiles, pipe, and bricks, porcelain china, sinks, and electrical insulators, and thermally insulating refractory bricks for ovens and fireplaces. Advanced ceramics, also referred to as ''high-tech'' ceramics, include products such as spark plug bodies, piston rings, catalyst supports, and water pump seals for automobiles, thermally insulating tiles for the space shuttle, sodium vapor lamp tubes in streetlights, and the capacitors, resistors, transducers, and varistors in the solid-state electronics we use daily. The major differences between traditional and advanced ceramics are in the processing tolerances and cost. Traditional ceramics are manufactured with inexpensive raw materials, are relatively tolerant of minor process deviations, and are relatively inexpensive. Advanced ceramics are typically made with more refined raw materials and processing to optimize a given property or combination of properties (e.g., mechanical, electrical, dielectric, optical, thermal, physical, and/or magnetic) for a given application. Advanced ceramics generally have improved performance and reliability over traditional ceramics, but are typically more expensive. Additionally, advanced ceramics are typically more sensitive to the chemical and physical defects present in the starting raw materials, or those that are introduced during manufacturing.

  14. Design requirements for SiC/SiC composites structural material in fusion power reactor blankets

    International Nuclear Information System (INIS)

    This paper recalls the main features of the TAURO blanket, a self-cooled Pb-17Li concept using SiC/SiC composites as structural material, developed for FPR. The objective of this design activity is to compare the characteristics of present-day industrial SiC-SiC composites with those required for a fusion power reactor blanket (FPR) and to evaluate the main needs of further R and D. The performed analyses indicated that the TAURO blanket would need the availability of SiC/SiC composites approximately 10 mm thick with a thermal conductivity through the thickness of approximately 15 Wm-1K-1 at 1000 C and a low electrical conductivity. A preliminary MHD analysis has indicated that the electrical conductivity should not be greater than 500 Ω-1m-1. Irradiation effects should be included in these figures. Under these conditions, the calculated pressure drop due to the high Pb-17Li velocity (approximately 1 m s-1) is much lower then 0.1 MPa. The characteristics and data base of the recently developed 3D-SiC/SiC composite, Cerasep trademark N3-1, are reported and discussed in relation to the identified blanket design requirements. The progress on joining techniques is briefly reported. For the time being, the best results have been obtained using Si-based brazing systems initially developed for SiC ceramics and whose major issue is the higher porosity of the SiC/SiC composites. (orig.)

  15. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  16. Large scale breeder reactor pump dynamic analyses

    International Nuclear Information System (INIS)

    The lateral natural frequency and vibration response analyses of the Large Scale Breeder Reactor (LSBR) primary pump were performed as part of the total dynamic analysis effort to obtain the fabrication release. The special features of pump modeling are outlined in this paper. The analysis clearly demonstrates the method of increasing the system natural frequency by reducing the generalized mass without significantly changing the generalized stiffness of the structure. Also, a method of computing the maximum relative and absolute steady state responses and associated phase angles at given locations is provided. This type of information is very helpful in generating response versus frequency and phase angle versus frequency plots

  17. Diffusive heat blanketing envelopes of neutron stars

    CERN Document Server

    Beznogov, M V; Yakovlev, D G

    2016-01-01

    We construct new models of outer heat blanketing envelopes of neutron stars composed of binary ion mixtures (H - He, He - C, C - Fe) in and out of diffusive equilibrium. To this aim, we generalize our previous work on diffusion of ions in isothermal gaseous or Coulomb liquid plasmas to handle non-isothermal systems. We calculate the relations between the effective surface temperature Ts and the temperature Tb at the bottom of heat blanketing envelopes (at a density rhob= 1e8 -- 1e10 g/cc) for diffusively equilibrated and non-equilibrated distributions of ion species at different masses DeltaM of lighter ions in the envelope. Our principal result is that the Ts - Tb relations are fairly insensitive to detailed distribution of ion fractions over the envelope (diffusively equilibrated or not) and depend almost solely on DeltaM. The obtained relations are approximated by analytic expressions which are convenient for modeling the evolution of neutron stars.

  18. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author)

  19. A Precambrian proximal ejecta blanket from Scotland

    Science.gov (United States)

    Amor, Kenneth; Hesselbo, Stephen P.; Porcelli, Don; Thackrey, Scott; Parnell, John

    2008-04-01

    Ejecta blankets around impact craters are rarely preserved onEarth. Although impact craters are ubiquitous on solid bodiesthroughout the solar system, on Earth they are rapidly effaced,and few records exist of the processes that occur during emplacementof ejecta. The Stac Fada Member of the Precambrian Stoer Groupin Scotland has previously been described as volcanic in origin.However, shocked quartz and biotite provide evidence for high-pressureshock metamorphism, while chromium isotope values and elevatedabundances of platinum group metals and siderophile elementsindicate addition of meteoritic material. Thus, the unit isreinterpreted here as having an impact origin. The ejecta blanketreaches >20 m in thickness and contains abundant dark green,vesicular, devitrified glass fragments. Field observations suggestthat the deposit was emplaced as a single fluidized flow thatformed as a result of an impact into water-saturated sedimentarystrata. The continental geological setting and presence of groundwatermake this deposit an analogue for Martian fluidized ejecta blankets.

  20. Stellar model atmospheres with magnetic line blanketing

    CERN Document Server

    Kochukhov, O; Shulyak, D

    2004-01-01

    Model atmospheres of A and B stars are computed taking into account magnetic line blanketing. These calculations are based on the new stellar model atmosphere code LLModels which implements direct treatment of the opacities due to the bound-bound transitions and ensures an accurate and detailed description of the line absorption. The anomalous Zeeman effect was calculated for the field strengths between 1 and 40 kG and a field vector perpendicular to the line of sight. The model structure, high-resolution energy distribution, photometric colors, metallic line spectra and the hydrogen Balmer line profiles are computed for magnetic stars with different metallicities and are discussed with respect to those of non-magnetic reference models. The magnetically enhanced line blanketing changes the atmospheric structure and leads to a redistribution of energy in the stellar spectrum. The most noticeable feature in the optical region is the appearance of the 5200 A depression. However, this effect is prominent only in ...

  1. Analysis of Consistency of Printing Blankets using Correlation Technique

    Directory of Open Access Journals (Sweden)

    Lalitha Jayaraman

    2010-01-01

    Full Text Available This paper presents the application of an analytical tool to quantify material consistency of offset printing blankets. Printing blankets are essentially viscoelastic rubber composites of several laminas. High levels of material consistency are expected from rubber blankets for quality print and for quick recovery from smash encountered during the printing process. The present study aims at determining objectively the consistency of printing blankets at three specific torque levels of tension under two distinct stages; 1. under normal printing conditions and 2. on recovery after smash. The experiment devised exhibits a variation in tone reproduction properties of each blanket signifying the levels of inconsistency also in thicknessdirection. Correlation technique was employed on ink density variations obtained from the blanket on paper. Both blankets exhibited good consistency over three torque levels under normal printing conditions. However on smash the recovery of blanket and its consistency was a function of manufacturing and torque levels. This study attempts to provide a new metrics for failure analysis of offset printing blankets. It also underscores the need for optimizing the torque for blankets from different manufacturers.

  2. TCT hybrid preconceptual blanket design studies

    Energy Technology Data Exchange (ETDEWEB)

    Aase, D.T.; Bampton, M.C.C.; Doherty, T.J.; Leonard, B.R.; McCann, R.A.; Newman, D.F.; Perry, R.T.; Stewart, C.W.

    1978-01-01

    The conceptual design of a tokamak fusion-fission (hybrid) reactor, which produces electric power and fissile material, has been performed in a cooperative effort between Princeton's Plasma Physics Laboratory (PPPL) and Battelle's Pacific Northwest Laboratories (PNL). PPPL, who had overall project lead responsibility, designed the fusion driver system. Its core consists of a tokamak plasma maintained in the two-component torus (TCT) mode by both D and T beams and having a single null poloidal divertor. The blanket concept selected by PPPL consists of a neutron multiplying converter region, containing natural Uranium Molybdenum (U-Mo) slugs followed by a fuel burning blanket region of molten salt containing PuF/sub 3/. PNL analyzed this concept to determine its structural, thermal and hydraulic performance characteristics. An adequate first wall cooling method was determined, utilizing low pressure water in a double wall design. A conceptual layout of the converter region tubes was performed, providing adequate helium cooling and the desired movement of U-Mo slugs. A thermal hydraulic analysis of the power-producing blanket regions indicated that either more helium coolant tubes are needed or the salt must be circulated to obtain adequate heat removal capability.

  3. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m2 and a particle heat flux of 1 MW/m2. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma

  4. Prototype fast breeder reactor main options

    International Nuclear Information System (INIS)

    Fast reactor programme gets importance in the Indian energy market because of continuous growing demand of electricity and resources limited to only coal and FBR. India started its fast reactor programme with the construction of 40 MWt Fast Breeder Test Reactor (FBTR). The reactor attained its first criticality in October 1985. The reactor power will be raised to 40 MWt in near future. As a logical follow-up of FBTR, it was decided to build a prototype fast breeder reactor, PFBR. Considering significant effects of capital cost and construction period on economy, systematic efforts are made to reduce the same. The number of primary and secondary sodium loops and components have been reduced. Sodium coolant, pool type concept, oxide fuel, 20% CW D9, SS 316 LN and modified 9Cr-1Mo steel (T91) materials have been selected for PFBR. Based on the operating experience, the integrity of the high temperature components including fuel and cost optimization aspects, the plant temperatures are recommended. Steam temperature of 763 K at 16.6 MPa and a single TG of 500 MWe gross output have been decided. PFBR will be located at Kalpakkam site on the coast of Bay of Bengal. The plant life is designed for 30 y and 75% load factor. In this paper the justifications for the main options chosen are given in brief. (author). 2 figs, 2 tabs

  5. Coincidence measurements of FFTF breeder fuel subassemblies

    International Nuclear Information System (INIS)

    A prototype coincidence counter developed to assay fast breeder reactor fuel was used to measure four fast-flux test facility subassemblies at the Hanford Engineering Development Laboratory in Richland, Washington. Plutonium contents in the four subassemblies ranged between 7.4 and 9.7 kg with corresponding 240Pu-effective contents between 0.9 and 1.2 kg. Large count rates were observed from the measurements, and plots of the data showed significant multiplication in the fuel. The measured data were corrected for deadtime and multiplication effects using established formulas. These corrections require accurate knowledge of the plutonium isotopics and 241Am content in the fuel. Multiplication-corrected coincidence count rates agreed with the expected count rates based on spontaneous fission-neutron emission rates. These measurements indicate that breeder fuel subassemblies with 240Pu-effective contents up to 1.2 kg can be nondestructively assayed using the shift-register electronics with the prototype counters. Measurements using the standard Los Alamos National Laboratory shift-register coincidence electronics unit can produce an assay value accurate to +-1% in 1000 s. The uncertainty results from counting statistics and deadtime-correction errors. 3 references, 8 figures, 8 tables

  6. The ITER Blanket System Design Challenge

    International Nuclear Information System (INIS)

    Full text: The blanket system is one of the most technically challenging components of the ITER machine, having to accommodate high heat fluxes from the plasma, large electromagnetic loads during off-normal events and demanding interfaces with many key components (in particular the vacuum vessel and in-vessel coils) and the plasma. Plasma scenarios impose demanding requirements on the blanket in terms of heat fluxes on various areas of the first wall during different phases of operation (inboard and outboard midplane for start-up/shut-down scenarios and the top region close to the secondary X-point during flat top) as well as large electro-magnetic (EM) loads and transient energy deposition during off-normal plasma events (such as disruptions and vertical displacement events (VDE)). The high heat fluxes resulting in some areas have necessitated the use of “enhanced heat flux” panels capable of accommodating an incident heat flux of up to 5 MW/m2 in steady state. The other regions utilize “normal heat flux” panels, which have been developed and tested for a heat flux of the order of 1 — 2 MW/m2. The FW shaping design requires a compromise between the conflicting requirements for accommodation of steady state and transient loads (energy deposition during off-normal events). A shaped surface increases the heat loads which are due to plasma particles following the field lines compared to a perfectly toroidal surface. The blanket provides a major contribution to the shielding of the vacuum vessel and coils. A challenging criterion is the need to limit the integrated heating in the toroidal field coil (TFC) to ∼ 14 kW. This is particularly severe on the inboard leg where approximately 80% of the total nuclear heat on the TFC is deposited. Several design modifications were considered and analyzed to help achieve this, including increasing the inboard blanket radial thickness and reducing the assembly gaps. This paper summarizes the latest progress in the

  7. Activation Calculation for a Fusion Experimental Breeder FEB-E

    Institute of Scientific and Technical Information of China (English)

    FENGKaiming

    2002-01-01

    A fusion breeder might be an essential intermediate application of fusion energy at earlier term, since it has the potential to provide plenty of commercial fissile fuel. Based on fusion physics and technologies available at present and in the near future, the realistic fusion experimental breeder, FEB-E was designed.

  8. Glass-ceramic joining and coating of SiC/SiC for fusion applications

    International Nuclear Information System (INIS)

    The aim of this work is the joining and the coating of SiC/SiC composites by a simple, pressureless, low cost technique. A calcia-alumina glass-ceramic was chosen as joining and coating material, because its thermal and thermomechanical properties can be tailored by changing the composition, it does not contain boron oxide (incompatible with fusion applications) and it has high characteristic temperatures (softening point at about 1400 C). Furthermore, the absence of silica makes this glass-ceramic compatible with ceramic breeder materials (i.e. lithium-silicates, -alluminates or -zirconates). Coatings and joints were successfully obtained with Hi-Nicalon fiber-reinforced CVI silicon carbide matrix composite. Mechanical shear strength tests were performed on joined samples and the compatibility with a ceramic breeder material was examined. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-08-01

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

  11. Advanced Acoustic Blankets for Improved Aircraft Interior Noise Reduction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this project advanced acoustic blankets for improved low frequency interior noise control in aircraft will be developed and demonstrated. The improved...

  12. Water chemistry of breeder reactor steam generators

    International Nuclear Information System (INIS)

    The water quality requirements will be described for breeder reactor steam generators, as well as specifications for balance of plant protection. Water chemistry details will be discussed for the following power plant conditions: feedwater and recirculation water at above and below 5% plant power, refueling or standby, makeup water, and wet layup. Experimental data will be presented from tests which included a departure from nucleate boiling experiment, the Few Tube Test, with a seven tube evaporator and three tube superheater, and a verification of control and on-line measurement of sodium ion in the ppB range. Sampling and instrumentation requirements to insure adherence to the specified water quality will be described. Evaporator cleaning criteria and data from laboratory testing of chemical cleaning solutions with emphasis on flow, chemical composition, and temperature will be discussed

  13. Development of hi-tech ceramics fabrication technology

    International Nuclear Information System (INIS)

    There are some ceramic materials being used in the nuclear energy such as nuclear fuel, coolant pump seals, tritium breeder materials, a high temperature absorber, and the solid electrolyte for recovering tritium. In addition, lots of researches recently have been conducted on the development of highly functional ceramics such as highly efficient shielding materials, functional graded materials and radioactive isotopes-separating materials. Therefore, one of the objectives of this project is to develop ultra-fine and pure powder manufacturing technology. Tritium breeder materials, LiAlO2, Li2ZrO3 and Li2TiO3 were made with a combustion process of mixed fuels that is developed indigenously in this project. Additionally, this study also focused on the development of promising low temperature electrolytes of ceria. By using the ceria powder made by the combustion process of GNP was investigated their sinterability and the electrolytic characteristics. (author). 167 refs., 74 tabs., 91 figs

  14. Systematic methodology for estimating direct capital costs for blanket tritium processing systems

    International Nuclear Information System (INIS)

    This paper describes the methodology developed for estimating the relative capital costs of blanket processing systems. The capital costs of the nine blanket concepts selected in the Blanket Comparison and Selection Study are presented and compared

  15. Engineering ceramics

    CERN Document Server

    Bengisu, Murat

    2001-01-01

    This is a comprehensive book applying especially to junior and senior engineering students pursuing Materials Science/ Engineering, Ceramic Engineering and Mechanical Engineering degrees. It is also a reference book for other disciplines such as Chemical Engineering, Biomedical Engineering, Nuclear Engineering and Environmental Engineering. Important properties of most engineering ceramics are given in detailed tables. Many current and possible applications of engineering ceramics are described, which can be used as a guide for materials selection and for potential future research. While covering all relevant information regarding raw materials, processing properties, characterization and applications of engineering ceramics, the book also summarizes most recent innovations and developments in this field as a result of extensive literature search.

  16. Progress in Solid Tritium Breeder Materials%固态氚增殖剂研究进展

    Institute of Scientific and Technical Information of China (English)

    赵林杰; 肖成建; 陈晓军; 龚宇; 彭述明; 龙兴贵

    2015-01-01

    增殖包层作为实现可控核聚变燃料“自持”的关键,不仅能实现氚的增殖,而且起着能量转换的作用,氚增殖剂是其中最重要的功能材料。本文从材料体系的制备、性能以及改性总结了固态氚增殖剂的发展趋势。同时,基于当前的研究现状对固态氚增殖剂的发展进行了展望。%The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction.Tritium breeding material is one of the most important functional materials.Herein,we reviewed the trends in solid tritium breeder development,including the fabrication,properties and modification.Meanwhile,the focus of the solid tritium breeder materials were prospected based on the current research situa-tion.

  17. Tailored ceramics

    International Nuclear Information System (INIS)

    In polyphase tailored ceramic forms two distinct modes of radionuclide immobilization occur. At high waste loadings the radionuclides are distributed through most of the ceramic phases in dilute solid solution, as indicated schematically in this paper. However, in the case of low waste loadings, or a high loading of a waste with low radionuclide content, the ceramic can be designed with only selected phases containing the radionuclides. The remaining material forms nonradioactive phases which provide a degree of physical microstructural isolation. The research and development work with polyphase ceramic nuclear waste forms over the past ten years is discussed. It has demonstrated the critical attributes which suggest them as a waste form for future HLW disposal. From a safety standpoint, the crystalline phases in the ceramic waste forms offer the potential for demonstrable chemical durability in immobilizing the long-lived radionuclides in a geologic environment. With continued experimental research on pure phases, analysis of mineral analogue behavior in geochemical environments, and the study of radiation effects, realistic predictive models for waste form behavior over geologic time scales are feasible. The ceramic forms extend the degree of freedom for the economic optimization of the waste disposal system

  18. ITER blanket manifold system: Integration, assembly and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Alex, E-mail: alex.martin@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Dellopoulos, George [F4E, EU ITER Domestic Agency, Barcelona (Spain); Edwards, Paul; Furmanek, Andreas; Gicquel, Stefan; Macklin, Brian [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Martin, Patrick [RÜECKER LYPSA, Carretera del Prat, 65, Cornellá de Llobregat (Spain); Merola, Mario; Norman, Mark; Raffray, Rene [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-10-15

    Highlights: •The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. •-The blanket manifold system has been redesigned to improve leak detection and localization. •-The redesign of the blanket manifold system into a system based on individual pipes has proven to be a major engineering challenge. -- Abstract: The ITER Tokamak Cooling Water System (TCWS) provides coolant for blankets and divertor. The blanket system consists of 440 blanket modules (BMs). The blanket manifold consists of a system of seamless pipes arranged in bundles and routed in poloidal direction from the upper ports of the Vacuum Vessel (VV) to the bottom of the machine. In each of the 18 upper ports there are 20 inlet and 20 outlet pipes, which split at the port exit in two directions, supplying cooling water to either the inboard or the outboard blanket modules. The manifold is routed between the VV and BMs. Branch pipes provide the connection between the manifold and the blanket cooling circuits through a coaxial connector welded to the shield block. A complex, sequential installation sequence has been developed in order to enable the assembly. Once installed the manifold is considered a semi-permanent component, but since failure would prevent ITER operation a maintenance strategy has been planned.

  19. 75 FR 51482 - Woven Electric Blankets From China

    Science.gov (United States)

    2010-08-20

    ... publishing the notice in the Federal Register of March 11, 2010 (75 FR 11557). The hearing was held in... COMMISSION Woven Electric Blankets From China Determination On the basis of the record \\1\\ developed in the... United States is materially injured by reason of imports from China of woven electric blankets,...

  20. ITER blanket manifold system: Integration, assembly and maintenance

    International Nuclear Information System (INIS)

    Highlights: •The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. •-The blanket manifold system has been redesigned to improve leak detection and localization. •-The redesign of the blanket manifold system into a system based on individual pipes has proven to be a major engineering challenge. -- Abstract: The ITER Tokamak Cooling Water System (TCWS) provides coolant for blankets and divertor. The blanket system consists of 440 blanket modules (BMs). The blanket manifold consists of a system of seamless pipes arranged in bundles and routed in poloidal direction from the upper ports of the Vacuum Vessel (VV) to the bottom of the machine. In each of the 18 upper ports there are 20 inlet and 20 outlet pipes, which split at the port exit in two directions, supplying cooling water to either the inboard or the outboard blanket modules. The manifold is routed between the VV and BMs. Branch pipes provide the connection between the manifold and the blanket cooling circuits through a coaxial connector welded to the shield block. A complex, sequential installation sequence has been developed in order to enable the assembly. Once installed the manifold is considered a semi-permanent component, but since failure would prevent ITER operation a maintenance strategy has been planned

  1. Accelerator breeder nuclear fuel production: concept evaluation of a modified design for ORNL's proposed TME-ENFP

    International Nuclear Information System (INIS)

    Recent advances in accelerator beam technology have made it possible to improve the target/blanket design of the Ternary Metal Fueled Electronuclear Fuel Producer (TMF-ENFP), an accelerator-breeder design concept proposed by Burnss et al. for subcritical breeding of the fissile isotope 233U. In the original TMF-ENFP the 300-mA, 1100-MeV proton beam was limited to a small diameter whose power density was so high that a solid metal target could not be used for producing the spallation neutrons needed to drive the breeding process. Instead the target was a central column of circulating liquid sodium, which was surrounded by an inner multiplying region of ternary fuel rods (239Pu, 232Th, and 238U) and an outer blanket region of 232Th rods, with the entire system cooled by circulating sodium. In the modified design proposed here, the proton beam is sufficiently spread out to allow the ternary fuel to reside directly in the beam and to be preceded by a thin (nonstructural) V-Ti steel firThe spread beam mandated a change in the design configuration (from a cylindrical shape to an Erlenmeyer flask shape), which, in turn, required that the fuel rods (and blanket rods) be replaced by fuel pebbles. The fuel residence time in both systems was assumed to be 90 full power days. A series of parameter optimization calculations for the modified TMF-ENFP led to a semioptimized system in which the initial 239Pu inventory of the ternary fuel was 6% and the fuel pebble diameter was 0.5 cm. With this system the 233Pu production rate of 5.8 kg/day reported for the original TMF-ENFP was increased to 9.3 kg/day, and the thermal power production at beginning of cycle was increased from 3300 MW(t) to 5240 MW(t). 31 refs., 32 figs., 6 tabs

  2. Appendix for blanket - University of Wisconsin: tritium issues

    International Nuclear Information System (INIS)

    The selection of the liquid metal alloys, Li17Pb83, as the tritium breeder with helium serving as the heat transfer fluid suggests two alternative techniques for the removal of tritium from the breeder. The low solubility of tritium in this liquid breeder requires only a simple vacuum degassing technique for tritium removal. Because of this high tritium partial pressure, tritium removal in the present design could potentially be achieved by either (a) slow circulation of the liquid LiPb alloy to an external degassing system, or (b) noncirculation of the liquid breeder so that the tritium permeates through the walls of the coolant tubes into the circulating helium for subsequent recovery. Both of these techniques were investigated with special attention given to the resultant tritium inventories in the liquid breeder and the helium system, and the potential for tritium permeation at the steam generator (SG)

  3. LMFBR Blanket Physics Project progress report No. 6

    International Nuclear Information System (INIS)

    Progress is summarized in experimental and analytical investigations of the neutronics and photonics of benchmark mockups of LMFBR blankets. During the reporting period work was devoted primarily to a wide range of analytical/numerical investigations, including blanket fuel management/economics studies, evaluation of improved blanket designs, and assessment of state-of-the-art methods for gamma heating calculations. Experimental work included preparations for resumption of MIT Reactor operations, primarily fabrication of improved steel reflector assemblies for blanket mockups, and development of an improved radiophotoluminescent readout device for LiF thermoluminescent detectors. The most significant finding was that the neutronic and economic performance of radial blanket assemblies are essentially independent of core size (rating) for radially-power-flattened cores. Hence the methodology and results of current experiments and calculations should be valid for the large commercial LMFBR's of the future

  4. LMFBR Blanket Physics Project progress report No. 6

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J. (ed.)

    1975-06-30

    Progress is summarized in experimental and analytical investigations of the neutronics and photonics of benchmark mockups of LMFBR blankets. During the reporting period work was devoted primarily to a wide range of analytical/numerical investigations, including blanket fuel management/economics studies, evaluation of improved blanket designs, and assessment of state-of-the-art methods for gamma heating calculations. Experimental work included preparations for resumption of MIT Reactor operations, primarily fabrication of improved steel reflector assemblies for blanket mockups, and development of an improved radiophotoluminescent readout device for LiF thermoluminescent detectors. The most significant finding was that the neutronic and economic performance of radial blanket assemblies are essentially independent of core size (rating) for radially-power-flattened cores. Hence the methodology and results of current experiments and calculations should be valid for the large commercial LMFBR's of the future.

  5. MIT LMFBR blanket research project. Final summary report

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J.

    1983-08-01

    This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record.

  6. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record

  7. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

    This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics

  8. Structural Ceramics Database

    Science.gov (United States)

    SRD 30 NIST Structural Ceramics Database (Web, free access)   The NIST Structural Ceramics Database (WebSCD) provides evaluated materials property data for a wide range of advanced ceramics known variously as structural ceramics, engineering ceramics, and fine ceramics.

  9. Tritium permeation and recovery for the helium-cooled molten salt fusion breeder

    International Nuclear Information System (INIS)

    Design concepts are presented to control tritium permeation from a molten salt/helium fusion breeder reactor. This study assumes tritium to be a gas dissolved in molten salt, with TF formation suppressed. Tritium permeates readily through the hot steel tubes of the reactor and steam generator and will leak into the steam system at the rate of about one gram per day in the absence of special permeation barriers, assuming that 1% of the helium coolant flow rate is processed for tritium recovery at 90% efficiency per pass. The proposed permeation barrier for the reactor tubes is a 10 μm layer of tungsten which, in principle, will reduce tritium blanket permeation by a factor of about 300 below the bare-steel rate. A research and development effort is needed to prove feasibility or to develop alternative barriers. A 1 mm aluminum sleeve is proposed to suppress permeation through the steam generator tubes. This gives a calculated reduction factor of more than 500 relative to bare steel, including a factor of 30 due to an assumed oxide layer. The permeation equations are developed in detail for a multi-layer tube wall including a frozen salt layer and with two fluid boundary-layer resistances. Conditions are discussed for which Sievert's or Henry's Law materials become flux limiters. An analytical model is developed to establish the tritium split between wall permeation and reactor-tube flow

  10. Influence of chemisorption products of carbon dioxide and water vapour on radiolysis of tritium breeder

    International Nuclear Information System (INIS)

    Highlights: • Chemisorption products affect formation proceses of radiation-induced defects. • Radiolysis of chemisorption products increase amount of radiation-induced defects. • Irradiation atmosphere influence radiolysis of lithium orthosilicate pebbles. - Abstract: Lithium orthosilicate pebbles with 2.5 wt% excess of silica are the reference tritium breeding material for the European solid breeder test blanket modules. On the surface of the pebbles chemisorption products of carbon dioxide and water vapour (lithium carbonate and hydroxide) may accumulate during the fabrication process. In this study the influence of the chemisorption products on radiolysis of the pebbles was investigated. Using nanosized lithium orthosilicate powders, factors, which can influence the formation and radiolysis of the chemisorption products, were determined and described as well. The formation of radiation-induced defects and radiolysis products was studied with electron spin resonance and the method of chemical scavengers. It was found that the radiolysis of the chemisorption products on the surface of the pebbles can increase the concentration of radiation-induced defects and so could affect the tritium diffusion, retention and the released species

  11. Activation characteristics and waste management options for some candidate tritium breeders

    International Nuclear Information System (INIS)

    Activation and transmutation characteristics are calculated for the candidate breeder compositions Li2O, LiAlO2, Li2SiO3, Li2ZrO3, LiVO3 and 17Li-83Pb. Irradiation conditions comprise a 2.5 y continuous exposure to the neutron flux appropriate to the outboard blanket zone of the EEF reference reactor with an assumed first wall neutron loading of 5 MW m-2. Results are presented for specific activity, surface γ-dose rate, ingestion and inhalation doses and compositional changes. Neglecting any retained tritium, activity is least for Li2 and LiVO3 and greatest for Li2ZrO3 and 17Li-83Pb. The silicate and aluminate are intermediate in level. Following reactor service, all the materials should be suitable, after appropriate conditioning, for geological disposal as Intermediate Level Waste. Alternatively, they could be considered for recycling to reclaim the unused lithium. In all cases, recycling is probably feasible within 10 y of removal from service and should be easier for the oxide silicate and vanadate. (orig.)

  12. Research about the Influence of Environmental Factors on Breeders Quality

    Directory of Open Access Journals (Sweden)

    Adina Popescu

    2011-10-01

    Full Text Available Along the growth period of the breeders, the monitoring of environmental parameters is a fundamental condition toensure the quality of the breeders used for reproduction. The results from the research presented in this paper wereobtained following experimental type investigations developed in vegetation and cold season within Carja 1-Vasluifish farm, on chemical and biological samples which were analyzed within the research laboratory of the Departmentof Aquaculture, Environmental Science and Cadastre. Were analyzed parameters which influence bio-productivity:temperature, oxygen, pH, the concentration of nitrites, nitrates, phosphates, the density and abundance ofphytoplankton and zooplankton, the individual weight and health condition of breeders. Analyzed parametersincluded mean values recorded in the optimal range for fish waters, as reflected in the numerical density andabundance of plankton and the average weight of Asian cyprinids breeders with a plankton nutritional spectrum.

  13. DT neutron irradiation experiment for evaluation of tritium recovery from WCCB blanket

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Kentaro, E-mail: ochiai.kentaro@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki (Japan); Kawamura, Yoshinori [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki (Japan); Hoshino, Tsuyoshi [Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori (Japan); Edao, Yuki; Takakura, Kosuke; Ohta, Masayuki; Sato, Satoshi; Konno, Chikara [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki (Japan)

    2014-10-15

    Highlights: • We have performed the tritium recovery experiment with the DT neutron source. • The recovered tritium corresponded to the calculated tritium production. • The recovered HTO was recovered at lower temperature and high water moisture. • The recovered HT increases at higher temperature and dry hydrogen circumstance. - Abstract: The Li{sub 2}TiO{sub 3} is one of candidate breeding materials of a water cooled ceramic breeding (WCCB) blanket. In order to clarify the tritium recovery property of the WCCB blanket with the Li{sub 2}TiO{sub 3} breeding material, we have performed the tritium recovery online experiment with the DT neutron source at the Fusion Neutronics Source facility in Japan Atomic Energy Agency (JAEA-FNS). We irradiated an experimental assembly simulating the WCCB blanket and recovered the tritium recovered from the Li{sub 2}TiO{sub 3} pebbles put into the assembly with a heater system, sweep gases and bubblers. The activity of the recovered tritium was measured with a liquid scintillation counter. From our tritium recovery online experiment and calculation, the followings were found out: (1) the recovered tritium corresponded to the calculated tritium production within the experimental error in the range of 573–1073 K and (2) the recovered HTO tended to be easily recovered at lower temperature and high water moisture. The recovered HT increases at higher temperature and dry hydrogen circumstance. However, the maximum level of the tritium gas recovery is around 90% even at higher temperature and 1% H{sub 2} circumstance.

  14. Exploding the myths about the fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, S.

    1979-01-01

    This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

  15. Group size adjustment to ecological demand in a cooperative breeder

    OpenAIRE

    Zöttl, Markus; Frommen, Joachim G.; Taborsky, Michael

    2013-01-01

    Environmental factors can determine which group size will maximize the fitness of group members. This is particularly important in cooperative breeders, where group members often serve different purposes. Experimental studies are yet lacking to check whether ecologically mediated need for help will change the propensity of dominant group members to accept immigrants. Here, we manipulated the perceived risk of predation for dominant breeders of the cooperatively breeding cichlid fish Neolampro...

  16. Liquid metal tribology in fast breeder reactors

    International Nuclear Information System (INIS)

    Liquid Metal Cooled Fast Breeder Reactors (LMFBR) require mechanisms operating in various sodium liquid and sodium vapor environments for extended periods of time up to temperatures of 900 K under different chemical properties of the fluid. The design of tribological systems in those reactors cannot be based on data and past experience of so-called conventional tribology. Although basic tribological phenomena and their scientific interpretation apply in this field, operating conditions specific to nuclear reactors and prevailing especially in the nuclear part of such facilities pose special problems. Therefore, in the framework of the R and D-program accompanying the construction phase of SNR 300 experiments were carried out to provide data and knowledge necessary for the lay-out of friction systems between mating surfaces of contacting components. Initially, screening tests isolated material pairs with good slipping properties and maximum wear resistance. Those materials were subjected to comprehensive parameter investigations. A multitude of laboratory scale tests have been performed under largely reactor specific conditions. Unusual superimpositions of parameters were analyzed and separated to find their individual influence on the friction process. The results of these experiments were made available to the reactor industry as well as to factories producing special tribo-materials. (orig.)

  17. Manufacture of blanket shield modules for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lorenzetto, P. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany)]. E-mail: Patrick.Lorenzetto@tech.efda.org; Boireau, B. [AREVA Centre Technique de Framatome, BP181, F-71200 Le Creusot (France); Boudot, C. [AREVA Centre Technique de Framatome, BP181, F-71200 Le Creusot (France); Bucci, P. [CEA, DTEN/S3ME/LMIC, 17 rue des Martyrs, F-38054 Grenoble (France); Furmanek, A. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Ioki, K. [ITER IT, Boltzmannstr. 2, D-85748 Garching (Germany); Liimatainen, J. [Metso Powdermet, P.O. Box 306, FIN-33101 Tampere (Finland); Peacock, A. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Sherlock, P. [NNC Ltd., Booths Hall, Knutsford, Cheshire WA16 8QZ (United Kingdom); Taehtinen, S. [VTT Industrial Systems, P.O. Box 1704, Espoo, FIN-02044 VTT (Finland)

    2005-11-15

    A research and development programme for the ITER blanket shield modules has been implemented in Europe to provide input for the design and the manufacture of the full-scale production components. It involves in particular the fabrication and testing of mock-ups (small scale and medium scale) and full-scale prototypes of shield blocks (SB) and first wall (FW) panels. The manufacturing feasibility of FW panels has been demonstrated for two copper alloy candidates. Two designs have been developed for the manufacture of the SB, one for a conventional fabrication route and one for a fabrication route based on the hot isostatic press technology. This paper presents the fabrication routes developed in Europe for the manufacture of the ITER Shield modules.

  18. Simulation of sludge blanket height in clarifiers

    Institute of Scientific and Technical Information of China (English)

    ZHOU Zhen; WU Zhi-chao; WANG Zhi-wei; GU Guo-wei

    2009-01-01

    Sludge blanket height (SBH) is an important parameter in the clarifier design,operation and control.Based on an overview and classification of SBH algorithms,a modifed SBH algorithm is proposed by incorporating a threshold concentration limit into a relative concentration sharp change algorithm to eliminate the disturbance of compression interfaces on the correct simulation of SBH.Pilot-scale test data are adopted to compare reliability of three SBH algorithms reported in literature and the modified SBH algorithm developed in this paper.Calculated results demonstrate that the three SBH algorithms give results with large deviation (>50%) from measured SBH,especially under low solid flux conditions.The modified algorithm is computationally efficient and reliable in matching the measured data.It is incorporated into a onedimensional clarifier model for stable simulation of pilot-scale experimental clarifier data and into dynamic simulation of a full-scale wastewater treatment plant (WWTP) clarifier data.

  19. Spacecraft thermal blanket cleaning: Vacuum bake of gaseous flow purging

    Science.gov (United States)

    Scialdone, John J.

    1990-01-01

    The mass losses and the outgassing rates per unit area of three thermal blankets consisting of various combinations of Mylar and Kapton, with interposed Dacron nets, were measured with a microbalance using two methods. The blankets at 25 deg C were either outgassed in vacuum for 20 hours, or were purged with a dry nitrogen flow of 3 cu. ft. per hour at 25 deg C for 20 hours. The two methods were compared for their effectiveness in cleaning the blankets for their use in space applications. The measurements were carried out using blanket strips and rolled-up blanket samples fitting the microbalance cylindrical plenum. Also, temperature scanning tests were carried out to indicate the optimum temperature for purging and vacuum cleaning. The data indicate that the purging for 20 hours with the above N2 flow can accomplish the same level of cleaning provided by the vacuum with the blankets at 25 deg C for 20 hours, In both cases, the rate of outgassing after 20 hours is reduced by 3 orders of magnitude, and the weight losses are in the range of 10E-4 gr/sq cm. Equivalent mass loss time constants, regained mass in air as a function of time, and other parameters were obtained for those blankets.

  20. Effective Thermal Property Estimation of Unitary Pebble Beds Based on a CFD-DEM Coupled Method for a Fusion Blanket

    Science.gov (United States)

    Chen, Lei; Chen, Youhua; Huang, Kai; Liu, Songlin

    2015-12-01

    Lithium ceramic pebble beds have been considered in the solid blanket design for fusion reactors. To characterize the fusion solid blanket thermal performance, studies of the effective thermal properties, i.e. the effective thermal conductivity and heat transfer coefficient, of the pebble beds are necessary. In this paper, a 3D computational fluid dynamics discrete element method (CFD-DEM) coupled numerical model was proposed to simulate heat transfer and thereby estimate the effective thermal properties. The DEM was applied to produce a geometric topology of a prototypical blanket pebble bed by directly simulating the contact state of each individual particle using basic interaction laws. Based on this geometric topology, a CFD model was built to analyze the temperature distribution and obtain the effective thermal properties. The current numerical model was shown to be in good agreement with the existing experimental data for effective thermal conductivity available in the literature. supported by National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2015GB108002, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  1. Development of electron beam ion source charge breeder for rare isotopes at Californium Rare Isotope Breeder Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Kondrashev, S.; Dickerson, C.; Levand, A.; Ostroumov, P. N.; Pardo, R. C.; Savard, G.; Vondrasek, R. [Physics Division, Argonne National Laboratory, Argonne, Illinois 60439 (United States); Alessi, J.; Beebe, E.; Pikin, A. [Collider-Accelerator Department, Brookhaven National Laboratory, Upton, New York 11973 (United States); Kuznetsov, G. I.; Batazova, M. A. [Budker Institute of Nuclear Physics, Novosibirsk 630090 (Russian Federation)

    2012-02-15

    Recently, the Californium Rare Isotope Breeder Upgrade (CARIBU) to the Argonne Tandem Linac Accelerator System (ATLAS) was commissioned and became available for production of rare isotopes. Currently, an electron cyclotron resonance ion source is used as a charge breeder for CARIBU beams. To further increase the intensity and improve the purity of neutron-rich ion beams accelerated by ATLAS, we are developing a high-efficiency charge breeder for CARIBU based on an electron beam ion source (EBIS). The CARIBU EBIS charge breeder will utilize the state-of-the-art EBIS technology recently developed at Brookhaven National Laboratory (BNL). The electron beam current density in the CARIBU EBIS trap will be significantly higher than that in existing operational charge-state breeders based on the EBIS concept. The design of the CARIBU EBIS charge breeder is nearly complete. Long-lead components of the EBIS such as a 6-T superconducting solenoid and an electron gun have been ordered with the delivery schedule in the fall of 2011. Measurements of expected breeding efficiency using the BNL Test EBIS have been performed using a Cs{sup +} surface ionization ion source for external injection in pulsed mode. In these experiments we have achieved {approx}70% injection/extraction efficiency and breeding efficiency into the most abundant charge state of {approx}17%.

  2. Development of electron beam ion source charge breeder for rare isotopes at Californium Rare Isotope Breeder Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Kondrashev S.; Alessi J.; Dickerson, C.; Levand, A.; Ostroumov, P.N.; Pardo, R.C.; Savard, G.; Vondrasek, R.; Beebe, E.; Pikin, A.; Kuznetsov, G.I.; Batazova, M.A.

    2012-02-03

    Recently, the Californium Rare Isotope Breeder Upgrade (CARIBU) to the Argonne Tandem Linac Accelerator System (ATLAS) was commissioned and became available for production of rare isotopes. Currently, an electron cyclotron resonance ion source is used as a charge breeder for CARIBU beams. To further increase the intensity and improve the purity of neutron-rich ion beams accelerated by ATLAS, we are developing a high-efficiency charge breeder for CARIBU based on an electron beam ion source (EBIS). The CARIBU EBIS charge breeder will utilize the state-of-the-art EBIS technology recently developed at Brookhaven National Laboratory (BNL). The electron beam current density in the CARIBU EBIS trap will be significantly higher than that in existing operational charge-state breeders based on the EBIS concept. The design of the CARIBU EBIS charge breeder is nearly complete. Long-lead components of the EBIS such as a 6-T superconducting solenoid and an electron gun have been ordered with the delivery schedule in the fall of 2011. Measurements of expected breeding efficiency using the BNL Test EBIS have been performed using a Cs{sup +} surface ionization ion source for external injection in pulsed mode. In these experiments we have achieved {approx}70% injection/extraction efficiency and breeding efficiency into the most abundant charge state of {approx}17%.

  3. Demonstration Tokamak Hybrid Reactor (DTHR) blanket design study, December 1978

    International Nuclear Information System (INIS)

    This work represents only the second iteration of the conceptual design of a DTHR blanket; consequently, a number of issues important to a detailed blanket design have not yet been evaluated. The most critical issues identified are those of two-phase flow maldistribution, flow instabilities, flow stratification for horizontal radial inflow of boiling water, fuel rod vibrations, corrosion of clad and structural materials by high quality steam, fretting and cyclic loads. Approaches to minimizing these problems are discussed and experimental testing with flow mock-ups is recommended. These implications on a commercial blanket design are discussed and critical data needs are identified

  4. Non-destructive assay of EBR-II blanket elements using resonance transmission analysis

    International Nuclear Information System (INIS)

    Resonance transmission analysis utilizing a faltered reactor beam was examined as a means of determining the 239Pu content in Experimental Breeder Reactor-II depleted uranium blanket elements. The technique uses cadmium and gadolinium falters along with a 239Pu fission chamber to isolate the 0.3 eV resonance in 239Pu. In the energy range of this resonance (0.1 eV to 0.5 ev), the total microscopic cross-section of 239Pu is significantly greater than the cross-sections of 238U and 235U. This large difference allows small changes in the 239Pu content of a sample to result in large changes in the mass signal response. Tests with small stacks of depleted uranium and 239Pu foils indicate a significant change in response based on the 239Pu content of the foil stack. In addition, the tests indicate good agreement between the measured and predicted values of 239Pu up to approximately two weight percent

  5. Stability of LMR oxide pins and blanket rods during run-beyond-cladding-break (RBCB) operation

    International Nuclear Information System (INIS)

    Since 1981, the U.S. Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan have collaborated on an operational reliability testing program in the Experimental Breeder Reactor II. The tests were designed to determine the irradiation behavior of liquid-metal reactor (LMR) oxide pins and blanket rods during steady-state, transient, and run-beyond-claddin-breach (RBCB) operation. Phase I tests completed in 1987 involved current LMR oxide designs and claddings; the phase II tests begun in 1988 concentrate on advanced LMR designs, large-diameter pins (7.5 mm), and advance cladding alloys. The cladding breaches in these tests have been readily detected by fission-gas and delayed-neutron (DN) precursor release. The condition of the fuel pin has been monitored by these releases during RBCB operation. A variety of failures have been intentionally studied in the RBCB portion of the program for operating times of up to 142 full-power days; also, several failure types have been incidentally experienced during the transient tests. Types of failure have included those induced by gas-pressure loading either naturally or by prethinning of the cladding defects, and fuel-cladding mechanical interaction (FCMI)-induced failures or secondary failures caused by the formation of low-density fuel-sodium reaction product (FSRP). This paper summarizes this experience with regard to LMR oxide fuel stability during RBCB operation

  6. Evaluation of compatibility of flowing liquid lithium curtain for blanket with core plasma in fusion reactors

    International Nuclear Information System (INIS)

    A global model analysis of the compatibility of flowing liquid lithium curtain for blanket with core plasma has been performed. The relationships between the surface temperature of lithium curtain and mean effective plasma charges, fuel dilution and produced fusion power have been obtained. Results show that under normal circumstances, the evaporation of liquid lithium does not affect Zeff seriously, but affects fuel dilution and fusion power sensitively. The authors have investigated the relationships between the flow velocity of liquid lithium and its surface temperature rise based on the conditions of the option II of the fusion experimental breeder (FEB-E) design with reversed shear configuration and fairly high power density. The authors concluded that the effects of evaporation from liquid lithium curtain for FEB-E on plasma are negligible even if the flow velocity of liquid lithium is as low as 0.5 m·s-1. Finally, the sputtering yield of liquid lithium saturated by hydrogen isotopes is briefly discussed

  7. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt / 13.2 MWe sodium cooled, loop type mixed carbide fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors and to serve as an irradiation facility for development of fuel and structural material for future fast reactors. The reactor achieved first criticality in October 1985 with small indigenously designed and fabricated Mark I core (70% PuC-30% UC). The reactor power was subsequently raised in steps to 17.4 MWt by addition of Mark II fuel subassemblies (55% PuC-45% UC) and with the Mark I fuel operating at the designed linear heat rating of 400 W/cm. The turbo-generator was synchronized with the grid in July 1997. The achieved peak burn-up is 137 000 MWd/t so far without any fuel-clad failure. Presently the reactor is being operated at a nominal power of 15.7 MWt for irradiation of a test fuel subassembly of the Prototype Fast Breeder Reactor, which is coming up at Kalpakkam. It is also planned to irradiate test subassemblies made of metallic fuel for future fast reactor program. Being a small reactor, all feed back coefficients of reactivity including void coefficient are negative and hence the reactor is inherently safe. This was confirmed by carrying out physics tests. The capability to remove decay heat under various incidental conditions including natural convection was demonstrated by carrying out engineering tests. Thermo couples are provided for on-line monitoring of fuel SA outlet temperature by dedicated real time computer and processed to generate trip signals for the reactor in case of power excursion, increase in clad hot spot temperature and subassembly flow blockage. All pipelines and capacities in primary main circuit are provided with segmented outer envelope to minimize and contain radioactive sodium leak while ensuring forced cooling through reactor to remove decay heat in case of failure of primary boundary. In secondary circuit, provision is

  8. Status of fast breeder development in Germany

    International Nuclear Information System (INIS)

    The German Minister for Research and Technology (BMFT), Dr. Heinz Riesenhuber, announced on March 20, 1991 that SNR 300, the fast breeder power plant at Kalkar, shall be abandoned. This message followed a top level meeting between BMFT officials and senior managers of Siemens, RWE, PreuBenElektra und Bayernwerk. BMFT, vendor Siemens and the three utilities had carried the interim finance costs of DM 105 million yearly since 1989. The licensing procedure had been obstructed during a long time by the responsible authorities. For several years the licensing process for the last permits on nuclear operation of KKW Kalkar had been held up by the government of the state of North Rhine-Westphalia (NWR). Licensing of nuclear power plants is the responsibility of the states, according to the German Atomic Act. The state of NRW turned against the SNR 300 project when the Social Democratic Party (SPD) started questioning nuclear power in 1985. Until then 17 partial licenses for SNR 300 had been granted, each time including an overall project approval. One of the consequences of the demise of SNR-300 was that Interatom GmbH, a subsidiary of Siemens AG, has been integrated into the division KWU of the Siemens AG on 1 October, 1991. For SNR 300 the turn-key contracts to the supplier company were cancelled by the operator on April 10, 1991 following the political termination of the SNR-300 Project. On August 23, 1991 after the termination of the SNR project, KfK decided to shutdown the KNK II reactor for final decommissioning

  9. The breeder reactor in electricity supply

    International Nuclear Information System (INIS)

    Forecasts are made of Britain's energy prospects in the year 2000. It is concluded that fossil fuels and renewable energy sources will leave an energy gap and that dependence on nuclear power will be substantial. There will, however have been a rapid depletion of readily available uranium ore reserves and a growing availability of plutonium from thermal reactors. Britain's resources of plutonium and depleted uranium which the fast breeder reactor can use - will equal many thousand million tonnes of coal. Our nuclear programme should therefore include one or two FBRs. Resources should be pooled internationally and plants built to prove alternative options and consolidate an already highly developed technology. In Britain our earliest nuclear (Magnox) stations have served as well. In Scotland, where next year an estimated 30% of electricity output will be nuclear, Hunterston 'B' AGR has had a splendid first year of operation, and pumped storage capacity in Scotland has extended the benefits of low-cost generation. The FBR has many very satisfactory engineering features and is eminently controllable and well behaved. It is compact, with relatively low cooling-water requirements and it could be built, one hopes, close to our load centres. There can be confidence that it will be proved safe. An order for an FBR, on the earliest timescale that fits in with evidence of successful operation of the Dounreay PFR and an agreed international programme, would serve the national interest and ensure the survival of our plant manufacturers, so badly hit by the effects of stagnation of the U.K. economy. (author)

  10. Characterization of the thermal conductivity for ceramic pebble beds

    Science.gov (United States)

    Lo Frano, R.; Aquaro, D.; Scaletti, L.; Olivi, N.

    2015-11-01

    The evaluation of the thermal conductivity of breeder materials is one of the main goals to find the best candidate material for the fusion reactor technology. The aim of this paper is to evaluate experimentally the thermal conductivity of a ceramic material by applying the hot wire method at different temperatures, ranging from 50 to about 800°C. The updated experimental facility, available at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa, used to determine the thermal conductivity of a ceramic material (alumina), will be described along with the measurement acquisition system. Moreover it will be also provided an overview of the current state of art of the ceramic pebble bed breeder thermos-mechanics R&D (e.g. Lithium Orthosilicate (Li4SiO4) and Lithium Metatitanate (Li2TiO3)) focusing on the up-to-date analysis. The methodological approach adopted is articulated in two phase: the first one aimed at the experimental evaluation of thermal conductivity of a ceramic material by means of hot wire method, to be subsequently used in the second phase that is based on the test rig method, through which is measured the thermal conductivity of pebble bed material. In this framework, the experimental procedure and the measured results obtained varying the temperature, are presented and discussed.

  11. Test blanket module maintenance operations between port plug and ancillary equipment unit in ITER

    International Nuclear Information System (INIS)

    In collaboration between the FZK and KFKI-RMKI, in the frame of the activities of the EU Breeder Blanket Programme a concept for test blanket module (TBM) integration, maintenance schedules and all required special purpose equipments has been developed. During the first 10 years of ITER operation four different plasma scenarios will be used. Hence it will be possible to investigate the characteristics (e.g. tritium breeding performance) of different TBM concepts which will be installed during operation for the different phases of ITER operation in the equatorial ports 2, 16 and 18. In every port two TBMs will be accommodated, in the port 16 will be the European helium-cooled pebble bed blanket. In different phases of ITER operation different TBMs will be used. Therefore a complex maintenance process is necessary for the exchange of TBMs. Two TBMs are mounted onto one common frame, into a port plug (PP), which offers a standardised interface to the vacuum vessel (VV). It is cantilevered with a flange to VV port extension. This attachment system is the same in every equatorial port, so the exchange process of this structure with the TBMs is also the standard operation of ITER. Several components of the helium cooling system of the EU breeder modules, valves, pipes, gas mixers, thermal sleeves, pipes for tritium extraction, measurement system are integrated into the ancillary equipment unit (AEU), which during the operation will connect the port plug to the subsystems. The bigger part of the AEU is accommodated in the port cell and the rest part of it is penetrated into the interspace inside the bioshield and reach the back plane of the installed PP. The remote handling operations for connection/disconnection of an interface between the PP of the EU-TBMs and the AEU are investigated with the goal to reach a quick and simple TBM exchange procedure. The current design of the EU-TBMs foresees up to 18 supply lines for both TBMs. These lines have to be connected here. A

  12. Test blanket module maintenance operations between port plug and ancillary equipment unit in ITER

    International Nuclear Information System (INIS)

    In collaboration between the FZK and KFKI-RMKI, in the frame of the activities of the EU Breeder Blanket Programme a concept for Test Blanket Module (TBM) integration, maintenance schedules and all required special purpose equipments has been developed. During the first 10 years of ITER operation 4 different plasma scenarios will be used. Hence it will be possible to investigate the characteristics (e.g. tritium breeding performance) of different TBM concepts which will be installed during operation for the different phases of ITER operation in the equatorial ports 2, 16 and 18. In every port will be two TBMs accomodated, in the port 16 will be the the European Helium Cooled Pebble Bed blanket. In the different phases of ITER operation different TBMs will be used. Therefore a complex maintenance process is necessary for exchange the TBMs. Two TBMs are mounted into one common frame, into a Port Plug (PP), which offers a standardised interface to the Vacuum Vessel (VV). It is cantilevered with a flange to VV Port Extension. This attachment system is the same in every equatorial port, so the exchange process of this structure with the TBMs are also standard operation of ITER. Several components of the Helium cooling system of the EU breeder modules, valves, pipes, gas mixers, thermal sleeves, pipes for tritium extraction, measurement system, etc. All of them is integrated into the Ancillary Equipment Unit (AEU) which during operation will connect the port plug to the sub systems. The bigger part of the AEU is accomodated in the Port Cell and the rest part of it is penetrate to the interspace inside the bioshield and reach the back plane of the installed PP. The remote handling operations for connection / disconnection of an interface between the PP of the EU-TBMs and the AEU are investigated with the goal to reach a quick and simple TBM exchange procedure. The current design of the EU-TBMs foresees up to 18 supply lines for both TBMs. These lines have to be connected

  13. 18 CFR 284.402 - Blanket marketing certificates.

    Science.gov (United States)

    2010-04-01

    ... effective for an affiliated marketer with respect to transactions involving affiliated pipelines when an affiliated pipeline receives its blanket certificate pursuant to § 284.284. (2) Should a marketer...

  14. Advanced Acoustic Blankets for Improved Aircraft Interior Noise Reduction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of the proposed Phase II research effort is to develop heterogeneous (HG) blankets for improved sound reduction in aircraft structures. Phase I...

  15. Lightweight IMM Multi-Junction Photovoltaic Flexible Blanket Assembly Project

    Data.gov (United States)

    National Aeronautics and Space Administration — DSS's recently completed successful NASA SBIR Phase 1 program has established a TRL 3/4 classification for an innovative IMM PV Integrated Modular Blanket Assembly...

  16. Electrical connectors for blanket modules in ITER

    International Nuclear Information System (INIS)

    Highlights: • Analysis of static and cyclic strength for L-shaped and Z-shaped ES has been performed. • Analysis results do show that for L-shaped ES static and cyclic strength criteria are not satisfied. • Static and cyclic strength criteria are met well by ES with Z-shaped elastic elements. • ES with Z-shaped elastic elements has been adopted as a new baseline design for ITER. - Abstract: Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2 mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137 kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9 T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak. The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was

  17. Blankets for tritium catalyzed deuterium (TCD) fusion reactors

    International Nuclear Information System (INIS)

    The TCD fusion fuel cycle - where the 3He from the D(D,n)3He reaction is transmuted, by neutron capture in the blanket, into tritium which is fed back to the plasma - was recently recognized as being potentially more promising than the Catalyzed Deuterium (Cat-D) fuel cycle for tokamak power reactors. It is the purpose of the present work to assess the feasibility of, and to identify promising directions for designing blankets for TCD fusion reactors

  18. Impact of prescribed burning on blanket peat hydrology

    OpenAIRE

    Holden, J; Palmer, SM; Johnston, K; Wearing, C.; Irvine, B; Brown, LE

    2015-01-01

    Fire is known to impact soil properties and hydrological flowpaths. However, the impact of prescribed vegetation burning on blanket peatland hydrology is poorly understood. We studied ten blanket peat headwater catchments. Five were subject to prescribed burning, while five were unburnt controls. Within the burnt catchments we studied plots where the last burn occurred ∼2 (B2), 4 (B4), 7 (B7) or greater than 10 years (B10+) prior to the start of measurements. These were compared with plots at...

  19. Compatibility of sodium with ceramic oxides employed in nuclear reactors; Compatibilidad del sodio con oxidos ceramicos utilizados en reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Acena Moreno, V.

    1981-07-01

    This work is a review of experiments carried out up to the present time on the corrosion and compatibility of ceramic oxides with liquid sodium at temperatures corresponding to those in fast breeder reactors. The review also includes the results of a thermo-dynamic/liquid sodium reactions. The exercise has been conducted with a view to effecting experimental studies in the future. (Author)

  20. Studies on steps affecting tritium residence time in solid blanket

    International Nuclear Information System (INIS)

    For the self sustaining of CTR fuel cycle, the effective tritium recovery from blankets is essential. This means that not only tritium breeding ratio must be larger than 1.0, but also high recovering speed is required for the short residence time of tritium in blankets. Short residence time means that the tritium inventory in blankets is small. In this paper, the tritium residence time and tritium inventory in a solid blanket are modeled by considering the steps constituting tritium release. Some of these tritium migration processes were experimentally evaluated. The tritium migration steps in a solid blanket using sintered breeding materials consist of diffusion in grains, desorption at grain edges, diffusion and permeation through grain boundaries, desorption at particle edges, diffusion and percolation through interconnected pores to purging stream, and convective mass transfer to stream. Corresponding to these steps, diffusive, soluble, adsorbed and trapped tritium inventories and the tritium in gas phase are conceivable. The code named TTT was made for calculating these tritium inventories and the residence time of tritium. An example of the results of calculation is shown. The blanket is REPUTER-1, which is the conceptual design of a commercial reversed field pinch fusion reactor studied at the University of Tokyo. The experimental studies on the migration steps of tritium are reported. (Kako, I.)

  1. Axial blanket for 16NGF Angra 1 fuel type

    Energy Technology Data Exchange (ETDEWEB)

    Sadde, Luciano Martins; Faria, Eduardo Fernandes [Industrias Nucleares do Brasil (INB), Resende, RJ (Brazil)]. E-mails: sadde@inb.gov.br; faria@inb.gov.br; Sang-Keun You [Korea Nuclear Fuel Co. Ltd. (KNFC), Taejon (Korea, Republic of)]. E-mail: skyou@knfc.co.kr

    2007-07-01

    Angra-1, Kori-2 and Krsko are nuclear power plants with the same design. However, the fuel assemblies have some differences in design due to the countries strategies and the differences in the fabrication process. The 16NGF (16x16 Next Generation Fuel) was developed by INB, KNFC and Westinghouse in order to be used in these three nuclear power plants and the 'Axial Blanket' is one of the new features for the 16NGF design. The main purpose of the Axial Blanket Optimization study is to determine which axial blanket enrichment and length would provide the better fuel cycle cost benefit. All of the calculations were performed using Gadolinium as Burnable Absorber and solid pellets type for Axial Blanket. The results indicate 1.8 w/o U235 enrichment and 8 inches length as the best option of Axial Blanket from the fuel cycle cost benefit standpoint. The economy is about 1.8%. The difference in the reload cost in the range between 1.5 and 2.6 w/o U235 enrichment and for the 6 and 8 inches length is not so significant. Due that, from the Fq limit standpoint and also for longer cycle length requirements, a higher axial blanket enrichment (2.6 w/o) and shorter length (6 inches) is recommended. (author)

  2. Hot isostatic pressing of ceramic waste from spent nuclear fuel

    International Nuclear Information System (INIS)

    Argonne National Laboratory has developed a process to immobilize waste salt containing fission products, uranium, and transuranic elements as chlorides in a glass-bonded ceramic waste form. This salt was generated in the electrorefining operation used in electrometallurgical treatment of spent Experimental Breeder Reactor-II fuel. The ceramic waste process culminated with a hot isostatic pressing operation. This paper reviews the installation and operation of a hot isostatic press in a radioactive environment. Processing conditions for the hot isostatic press are presented for non-irradiated material and irradiated material. Sufficient testing was performed to demonstrate that a hot isostatic press could be used as the final step of the processing of ceramic waste for the electrometallurgical spent fuel treatment process

  3. The integrated-blanket-coil concept applied to the poloidal field and blanket systems of a tokamak reactor

    International Nuclear Information System (INIS)

    A novel concept is proposed for combining the blanket and coil functions of a fusion reactor into a single component. This concept, designated the ''integrated-blanket-coil'' (IBC) concept, is applied to the poloidal field and blanket systems of a tokamak reactor. An examination of resistive power losses in the IBC suggests that these losses can be limited to 10% of the fusion thermal power. By assuming a sandwich construction for the IBC walls, magnetohydrodynamic (MHD)-induced pressure drops and associated pressure stresses are shown to be modest and well below design limits. For the stainless steel reference case examined, the MHD-induced pressure drop was estimated to be about 1/3 MPa and the associated primary membrane stress was estimated to be about 47 MPa. The preliminary analyses indicate that the IBC concept offers promise as a means for making fusion reactors more compact by combining blanket and coil functions in a single component

  4. Safety Evaluation of the EVOLVE Blanket Concept

    International Nuclear Information System (INIS)

    This article summarizes the results of the safety evaluation of the Evaporation of Lithium and Vapor Extraction (EVOLVE) W-alloy first wall (FW) and blanket concept. We have analyzed the EVOLVE design response during a confinement bypass accident. A confinement bypass accident was chosen because, based on previous safety studies, this accident can produce environmental releases by breaching the primary radioactive confinement boundary of EVOLVE, which is the EVOLVE vacuum vessel (VV). As a consequence of a bypass accident, air from a room adjoining the reactor enters the plasma chamber by way of a failed VV port. This air reacts with the high temperature metals inside of the VV to release energy in the case of a lithium spill, or to mobilize radioactive material by oxidation, and then transport this material to the environment by natural convection airflow through the failed VV port. We use the MELCOR code to analyze the response of EVOLVE during this accident. Based on these results, the EVOLVE concept can meet the no-evacuation dose goal set by the DOE Fusion Safety Standard if the EVOLVE confinement building ventilation system is closed within two hours of the onset of this accident

  5. Current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

    Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phase of the Li-LiH, Li-LiD, and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li3N, Li2O, and Li2C2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g., Li--Al and Li--Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li--M alloys can be estimated from lithium activity data for these alloys

  6. Flow characteristics of the Cascade granular blanket

    International Nuclear Information System (INIS)

    Analysis of a single granule on a rotating cone shows that for the 350 half-angle, double-cone-shaped Cascade chamber, blanket granules will stay against the chamber wall if the rotational speed is 50 rpm or greater. The granules move axially down the wall with a slight (5-mm or less) sinusoidal oscillation in the circumferential direction. Granule chute-flow experiments confirm that two-layered flow can be obtained when the chute is inclined slightly above the granular material angle of repose. The top surface layer is thin and fast moving (supercritical flow). A thick bottom layer moves more slowly (subcritical flow controlled at the exit) with a velocity that increases with distance from the bottom of the chute. This is a desirable velocity profile because in the Cascade chamber about one-third of the fusion energy is deposited in the form of x rays and fusion-fuel-pellet debris in the top surface (inner-radius) layer

  7. Environmental assessment for Breeder Reprocessing Engineering Test (BRET): Revision 1

    International Nuclear Information System (INIS)

    This Environmental Assessment (EA) is for the proposed installation and operation of an integrated breeder fuel reprocessing test system in the shielded cells of the Fuels and Materials Examination Facility (FMEF) at Hanford and the associated modifications to the FMEF to accommodate BRET. These modifications would begin in FY-1986 subject to Congressional authorization. Hot operations would be scheduled to start in the early 1990's. The system, called the Breeder Reprocessing Engineering Test (BRET), is being designed to provide a test capability for developing the demonstrating fuel reprocessing, remote maintenance, and safeguards technologies for breeder reactor fuels. This EA describes (1) the action being proposed, (2) the existing environment which would be affected, (3) the potential environmental impacts from normal operations and severe accidents from the proposed action, (4) potential conflicts with federal, state, regional, and/or local plans for the area, and (5) environmental implications of alternatives considered to the proposed action. 41 refs., 10 figs., 31 tabs

  8. Fast breeders role in the energy supply of the EC

    International Nuclear Information System (INIS)

    The investigation summarized in this article was initiated by a work team of the International Society of Power Generators (UNIPEDE) and the EC-commission. The first part presents the results of the possible introduction of fast breeder reactors in the EC for power generation and describes its effects on the demand for natural uranium. The second part describes the present development level of reprocessing of breeder reactor fuel, a part of the fuel cycle which is of very special importance. With the assumption of a rather undisturbed utilization of nuclear energy the investigation comes to the result that the development of the fast breeders and their fuel cycle in the EC must be promoted in any case. And, in the future, the available means should be used for a balanced development of both the reactor system and the fuel cycle. (orig.)

  9. On the history of the Fast Breeder Project

    International Nuclear Information System (INIS)

    The evolution of the Fast Breeder Project from its beginning at the Karlsruhe Nuclear Research Center to the present cooperation of various organisations especially in the Federal Republic of Germany, the Netherlands, Belgium and France is described in its historical context. Where as the emphasis was on physical studies of fast neutron cores in the early phase, technological and safety problems gained importance in the subsequent development. The increasing collaboration with industry and the support by government funds resulted in the design and start of construction of the prototype SNR 300. The objectives and the reasoning underlying important intermediate decisions are described. In the meantime, licensing and funding problems have become decisive for the project schedule. The present report also gives an account of the international and national political aspects which influence the breeder reactor development. In the annex all fast breeder publications of the Karlsruhe Nuclear Research Center are listed. (orig.)

  10. The United States of America fast breeder reactor program

    International Nuclear Information System (INIS)

    The reasons for the development of the fast breeder reactor in the United States are outlined, and the LMFBR program is discussed in detail, under the following headings: program objectives, reactor physics, fuel and materials development, fuel recycle, safety, components, plant experience program (Near Commercial Breeder Reactor). The special facilities to be used at each stage of the program are described. It is planned that the Near Commercial Breeder Reactor will be complete in 1986, and commercial plants should follow in rapid succession. An alternate fast reactor concept (Gas Cooled Fast Reactor) is outlined. The Environmental Impact Statement for the proposed program is summarized, and the cost benefit analysis supplied as part of the Environment Statement is also summarized. (U.K.)

  11. APT Blanket System Loss-of-Helium-Gas Accident Based on Initial Conceptual Design - Helium Supply Rupture into Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    The model results are used to determine if beam power shutdown is necessary (or not) as a result of the LOHGA accident to maintain the blanket system well below any of the thermal-hydraulic constraints imposed on the design. The results also provide boundary conditions to the detailed bin model to study the detailed temperature response of the hot blanket module structure. The results for these two cases are documented in the report.

  12. Neutronic analysis of denaturing plutonium in a thorium fusion breeder and power flattening

    International Nuclear Information System (INIS)

    The purpose of this study is to denature nuclear weapon grade quality plutonium in a thorium fusion breeder. Ten fuel rods containing the mixture of ThO2 and PuO2 are placed in a radial direction in the fissile zone where ThO2 is mixed with variable amounts of PuO2 to obtain a quasi-constant nuclear heat production density. The plutonium composition volume fractions in the fuel rods are gradually increased from 0.1% to 1% by 0.1% increments. The fissile fuel zone is cooled with four various coolants with a volume fraction ratio of 1 (Vcoolant/Vfuel = 1). These coolants are helium gas, flibe 'Li2BeF4', natural lithium and eutectic lithium 'Li17Pb83'. Nuclear weapon grade quality 239Pu in the fuel composition is denatured due to the accumulation of the 240Pu isotope in the fissile zone after 18 months of plant operations. Under a first wall fusion neutron current load of 2.222 x 1014 (14.1 MeV n/cm2 s), which corresponds to 5 MW/m2, by a plant factor of 100%, at the end of the plant operation, the fissile fuel enrichment quality between 6.0% and 10% is obtained depending on the coolant types. During the plant operation, the tritium breeding ratio (TBR) should be at least 1.05. In the selected blanket, only the flibe coolant is already self sustaining at start up. The TBR increases steadily due to the higher neutron multiplication rate during the plant operation period. The highest TBR is obtained for the eutectic lithium coolant 1.4035, followed by the flibe coolant 1.3095, helium gas coolant 1.2172 and natural lithium coolant 1.0553 at the end of the operation period of 48 months. The energy multiplication factor M changed between 2.1731 and 6.6241 depending on coolant type during the operation period. The peak to average fission power density ratio Γ in the blanket decreases by ∼15%, which allows a more uniform power generation in the fissile zone. The isotopic percentage of 240Pu reaches higher than 5% in all coolant types. This is very important for

  13. Prevalence of Campylobacter jejuni in poultry breeder flocks

    Directory of Open Access Journals (Sweden)

    Ludovico Dipineto

    2010-01-01

    Full Text Available The aim of this work is to present the preliminary results of a study about the prevalence of Campylobacter jejuni in poultry breeder flocks. It was examined three different breeder flocks of Bojano in Molise region. A total of 360 cloacal swabs and 80 enviromental swabs was collected. Of the 3 flocks studied, 6.9% tested were positive for Campylobacter spp. The most-prevalent isolated species is C. jejuni (8.2%. Only 3 of the 360 cloacal swabs samples examined were associated with C. coli. The environmental swabs resulted negative. This results confirms again that poultry is a reservoir of this germ.

  14. Are Kirindy sifaka capital or income breeders? It depends.

    Science.gov (United States)

    Lewis, R J; Kappeler, P M

    2005-11-01

    The capital and income breeding framework has only recently been used to explain variation in female reproductive strategies in primates. The application of this framework to primates and other mammals with long reproductive cycles has not been consistent. We evaluated data on Verreaux's sifaka (Propithecus verreauxi verreauxi) in the Kirindy Forest of western Madagascar to determine whether they are capital or income breeders. We found that Verreaux's sifaka can be classified as either capital or income breeders, depending on how these concepts are operationalized. These conflicting findings highlight why the capital/income framework is currently problematic and must be standardized if it is to be a useful framework for primatologists.

  15. A study on the environmentally benign fusion breeder-transmuter

    International Nuclear Information System (INIS)

    The present study is an attempt to demonstrate the fusion breeder as a concept environmentally benign, which should help to promote the idea of fusion energy. Thus a sketch of design for a fusion hybrid aimed at satisfying the requirements of: 1. economy (thanks to fissile fuel production), 2. safety (low power density), 3. environment (reduction of impact) is presented. The emphasis which is put on the reliability of performed neutronic calculations (e.g. resonance self-shielding) permits one to recognize the advantages of fusion breeder as confirmed and its development as deserving a significant support. (author)

  16. U.S. reference paper on national decisions on breeder development and deployment

    International Nuclear Information System (INIS)

    Factors involved in making national decisions on the deployment of breeder reactor systems are identified in terms of a nation's potential for electrification, capital resources, the available industrial and manpower infrastructure and importance attached to energy independence and the degree to which a breeder program can help realize this objective in the time scale of interest. The specific factors analysed are: the high capital cost of the breeder and the one-time transition costs to bring the breeder to maturity the high breeder research, development and demonstration costs, the impact of discount rate, and the fuel cycle costs, e.g. indigeneous facilities or purchase of services. A principal conclusion of this paper is that nations may find it more economical to continue to deploy LWRs for a number of years rather than to consider the breeder option because of the initial high breeder capital cost and high breeder R and D costs

  17. Development of metallic fuels for Indian Fast Breeder Reactors

    International Nuclear Information System (INIS)

    The neutronic performance of metal fuel based on binary U-Pu alloy or ternary U-Pu-Zr alloys are better than conventional uranium plutonium mixed oxide or high density carbide ceramic fuel. The growing energy demand in India needs faster growth of nuclear power and warrants introduction of fast reactors based on metallic fuels in future. Physics calculation showed that fast reactor based on metallic fuels results in higher breeding ratio and lower doubling time compare to mixed oxide or carbide fuels. Moreover inclusion of pyro-processing of the fuel in the fuel cycle is expected to make metal fuel option more economical. As part of metal fuel development programme for future FBR's in India, capsule irradiation of metal fuel based on sodium bonded U-Pu-Zr alloy and metal (Zircaloy) bonded binary U-Pu (Pu ∼ 15 %) alloy are being actively pursued. For this purpose two design concepts have been proposed : one based on sodium bonded ternary alloy fuel of U-Pu-Zr (2-10 wt%) in modified T91 cladding material and the other is U-Pu binary alloy mechanically bonded to modified T91 cladding material with 'Zircaloy' as a liner between the fuel alloy and the clad. The Zircaloy liner act as a barrier in reducing the fuel clad chemical interaction. It also helps in transfer of heat from the fuel to the clad. The smear density of metal bonded pin will be between 70% - 85% and that for sodium bonded pin will be ∼ 70%. In metal bonded fuel pin design two/four semi-circular grooves of diameter ∼1.0 mm, will be provided in diametrically opposite directions in the fuel cross section to accommodate fuel swelling. A comparison has been made on the relative merits and demerits of these two fuel pin designs. The material for the axial blanket will be 'U' or U-Zr (Zr upto 10wt %) alloy based on the results of the out-of-pile thermal cycling behavior and irradiation performance. In the present investigation out-of-pile experiments have been carried out to address some of the issues of

  18. Status and prospects of thermal breeders and their effect on fuel utilization

    International Nuclear Information System (INIS)

    The report evaluates the extent to which thermal breeders and near-breeders might complement fast breeders or serve as an alternative in solving the long-term nuclear fuel supply problem. It considers in a general way issues such as proliferation, safety, environmental impacts, economics, power plant availability and fuel cycle versatility in order to determine whether thermal breeder reactors offer advantages or disadvantages with respect to such issues

  19. Multiple recycling of fuel in prototype fast breeder reactor in a closed fuel cycle with pressurized heavy-water reactor external feed

    Indian Academy of Sciences (India)

    G Pandikumar; A John Arul; P Puthiyavinayagam; P Chellapandi

    2015-10-01

    A fast breeder reactor (FBR) closed fuel cycle involves recycling of the discharged fuel, after reprocessing and refabrication, in order to utilize the unburnt fuel and the bred fissile material. Our previous study in this regard for the prototype fast breeder reactor (PFBR) indicated the possibility of multiple recycling with self-sufficiency. It was found that the change in Pu composition becomes negligible (less than 1%) after a few cycles. The core-1 Pu increases by 3% from the beginning of cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th by only 0.3%. In this work, the possibility of multiple recycling of PFBR fuel with external plutonium feed from pressurized heavy-water reactor (PHWR) is examined. Modified in-core cooling and reprocessing periods are considered. The impact of multiple recycling on PFBR core physics parameters due to the changes in the fuel composition has been brought out. Instead of separate recovery considered for the core and axial blankets in the earlier studies, combined fuel recovery is considered in this study. With these modifications and also with PHWR Pu as external feed, the study on PFBR fuel recycling is repeated. It is observed that the core-1 initial Pu inventory increases by 3.5% from cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th is only 0.35%. A comparison of the studies done with different external plutonium options viz., PHWR and PFBR radial blanket has also been made.

  20. DeBeNe Test Facilities for Fast Breeder Development

    International Nuclear Information System (INIS)

    This report gives an overview and a short description of the test facilities constructed and operated within the collaboration for fast breeder development in Germany, Belgium and the Netherlands. The facilities are grouped into Sodium Loops (Large Facilities and Laboratory Loops), Special Equipment including Hot Cells and Reprocessing, Test Facilities without Sodium, Zero Power Facilities and In-pile Loops including Irradiation Facilities

  1. Symposium on key questions about the fast breeder reactor

    International Nuclear Information System (INIS)

    Except for several introductions on various aspects of the fast breeder reactor development this paper contains the full texts of the discussions held in the sub-groups panels on resp. technical matters, environment and health, society, politics and economics. The main issues of each discussion are summarized

  2. Feeding broiler breeder flocks in relation to bird welfare aspects

    NARCIS (Netherlands)

    Jong, de I.C.; Krimpen, van M.M.

    2011-01-01

    To ensure health and reproductive capacity of the birds, broiler breeders are fed restricted during the rearing period, and to a lesser extent also during the production period. Although restricted feeding improves health and thereby bird welfare, on the other hand the birds are chronically hungry a

  3. Clinch River Breeder Reactor Plant Project: construction schedule

    International Nuclear Information System (INIS)

    The construction schedule for the Clinch River Breeder Reactor Plant and its evolution are described. The initial schedule basis, changes necessitated by the evaluation of the overall plant design, and constructability improvements that have been effected to assure adherence to the schedule are presented. The schedule structure and hierarchy are discussed, as are tools used to define, develop, and evaluate the schedule

  4. Status of fast breeder reactor development in the United States

    International Nuclear Information System (INIS)

    The energy policy of the United States is aimed at shifting as rapidly as practicable from an oil dependent economy to one that relies heavily on other fuels and energy sources. Nuclear power Is now and is expected to continue to be an important factor in achieving this goal. If nuclear power is to contribute to a solution of future energy needs, demonstration of the breeder reactor as a viable source of essentially inexhaustible energy supply is essential. The US DOE program for development of the fast breeder reactor has witnessed some notable events in the past year. Foremost among these Is the successful operational testing of the Fast Flux Test Facility (FFTF), located at.the Hanford Engineering Development Laboratory. The reactor reached full design power of 400 MW(t) on December 21, 1980, and has performed remarkably close to design specifications. Design of the Clinch River Breeder Reactor Plant (CRBRP), a 375 MW(e) LMFBR, is now over 80 percent complete. About $530 million in components have been ordered; component deliveries total approximately $124 million; work-in-process totals another $204 million. Construction of the plant, however, has been suspended since 1977. With the concurrence of the U.S. Congress and approvals from the appropriate authorities work on the safety review and site clearing for construction can resume. The Conceptual Design Study for a large, 1000 MW(e) LMFBR Large Developmental Plant was recently completed on a schedule commensurate with submission of a full report to the Congress at the end of March, 1981. This report is the culmination of a study which began in October, 1978 and involved contributions from U.S. reactor manufacturers and US DOE laboratories. The US DOE is carrying forward a comprehensive technology development program. This effort provides direct support to the FFTF and CRBRP projects and to the LDP. It also supports technology development which is generic to the overall LMFBR program. Funding for breeder

  5. The use of lithium oxide as the breeder in fusion reactors

    International Nuclear Information System (INIS)

    Lithium oxide as a fusion blanket material has neutronic advantages but various design limitations. The study was undertaken to investigate the design implications, to demonstrate how the limitations can be overcome and to provide guidance for future development. The study included lithium oxide properties, tritium control, coolant chemistry, blanket engineering and blanket neutronics. (author)

  6. The breeder spent fuel packaging and transportation program

    International Nuclear Information System (INIS)

    The Breeder Spent Fuel Handling and Transportation Program of the United States Department of Energy (DOE) was established in 1983 in order to develop a reliable planning base for interface development at the back end of the liquid metal fast breeder reactor (LMFBR) fuel cycle. It began by addressing the immediate interface needs between the planned Clinch River Breeder Reactor, near Oak Ridge, Tennessee, and the proposed Breeder Reprocessing Engineering Test Facility at Richland, Washington, and concluded by providing a developmental plan leading to a sodium-cooled spent breeder fuel transportation cask for a mature 20-reactor LMFBR industry in the year 2025. During the formulation of this plan, as well as during the technology development that constituted the programme, liaison between the DOE and the concerned private industry operations was maintained by frequent meetings. As a result of functional considerations, it was decided that a legal truck-weight stainless steel multi-assembly package would both be economical and would have unlimited routine possibilities and facility access. As the detailed conceptual design emerged, it included remotely workable, spring-loaded, captive bolts to reduce occupational exposure, internal integral impact limiters and a structurally promising depleted uranium gamma shield. Modular baskets of a boron-aluminium alloy, produced by Fonderies Montupet of France, would enhance criticality control and heat transfer, as well as allowing for either a spent fuel or high level waste payload. While preliminary calculations have qualified the structure and shielding, heat transfer from a six-assembly payload still poses problems. Details are discussed in the paper. (author)

  7. Fabrication and quality control of MOX fuel for Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Full text: Uranium-Plutonium mixed oxide (MOX) fuel for both thermal and fast reactors have been fabricated by Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, India. MOX fuel bundles fabricated by AFFF have been loaded in Boiling Water Reactors (BWRs) and Pressurised Heavy Water Reactors (PHWRs) and have been discharged after successful irradiation. An experimental fuel subassembly containing 37 MOX pins is being irradiated in Fast Breeder Test Reactor (FBTR) at Kalpakkam near Chennai and has seen a burn up of more than 80000 MWD/T. MOX fuel pins containing 44% Pu02 have been recently loaded as a part of the hybrid core of FBTR. AFFF has now taken up the manufacture of MOX fuel pins for the Prototype Fast Breeder Reactor (BHAVINI) coming up at Kalpakkam. The core consists of 181 sub assemblies containing 217 MOX fuel pins each. It is required to fabricate nearly 40,000 MOX fuel pins (3 meter long) for the first core. The Prototype Fast Breeder Reactor is designed with two different fissile enrichment zones to be loaded with MOX subassemblies with a nominal composition of 21% and 28% of PuO2. The fuel pellets of required composition are made using conventional powder metallurgy processes. The pellets are annular with an inner hole of 1.8mm diameter and outside diameter of 5.5mm. AFFF has developed the technology of making annular MOX fuel pellets for PFBR and optimized conditions of fabrication. Multistation rotary presses have been used for compaction of the pellets. The fuel pin consists of a MOX stack of 1000mm and axial blanket of deeply depleted uranium dioxide of length 300mm on either side. New techniques have been used at different stages of fabrication of the fuel pins namely pelletisation, welding and wire wrapping. Studies have been made to use laser welding technique for welding of endplugs. Automation has been introduced in a number of process steps in the fabrication line. A detailed quality control plan is prepared

  8. Fabrication and quality control of MOX fuel for Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Uranium-Plutonium mixed oxide (MOX) fuel for both thermal and fast reactors have been fabricated by Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, India. MOX fuel bundles fabricated by AFFF have been loaded in Boiling Water Reactors (BWRs) and Pressurised Heavy Water Reactors (PHWRs) and have been discharged after successful irradiation. An experimental fuel subassemby containing 37 MOX pins is being irradiated in Fast Breeder Test Reactor (FBTR) at Kalpakkam near Chennai and has seen a burn up of more than 92000 MWd/t. MOX fuel pins containing 44% PuO2 have been recently loaded as a part of the hybrid core of FBTR. AFFF has now taken up the manufacture of MOX fuel pins for the Prototype Fast Breeder Reactor (PFBR) coming up at Kalpakkam . The core consists of 181 sub assemblies containing 217 MOX fuel pins each. Prototype Fast Breeder Reactor is designed with two different fissile enrichment zones to be loaded with MOX subassemblies with a nominal composition of 21% and 28% of PuO2.The fuel pellets of required composition are made using conventional powder metallurgy processes. The pellets are annular with an inner hole of 1.8 mm diameter and outside diameter of 5.5 mm. AFFF has developed the technology of making annular MOX fuel pellets for PFBR and optimized conditions of fabrication. Multistaion rotary presses have been used for compaction of the pellets. The fuel pin consists of a MOX stack of 1000 mm and axial blanket of deeply depleted uranium dioxide of length 300 mm on either side. New techniques have been used at different stages of fabrication of the fuel pins namely pelletisation, welding and wire wrapping. Studies have been made to use laser welding technique for welding of endplugs. Automation has been introduced in a number of process steps in the fabrication line. A detailed quality control plan is prepared based on the specifications and advanced process and quality control procedures have been incorporated to

  9. The nuclear question at the start of the '80s: the breeder reactor

    International Nuclear Information System (INIS)

    The four newspaper articles and the letter cover the following matters: general introduction about breeder reactors and the situation in Swedish politics; visit to Dounreay to discuss breeder reactors (breeding, safety, plutonium production, radiation protection); PuO2-UO2 mixed fuel; description of breeder reactors; efficiency in use of U-235; DFR and PFR; breeder reactors in Swedish politics (arguments for and against nuclear power in general, breeder reactors in particular); discussion of the future of nuclear power in Sweden. (U.K.)

  10. Dental ceramics: An update

    OpenAIRE

    Shenoy Arvind; Shenoy Nina

    2010-01-01

    In the last few decades, there have been tremendous advances in the mechanical properties and methods of fabrication of ceramic materials. While porcelain-based materials are still a major component of the market, there have been moves to replace metal ceramics systems with all ceramic systems. Advances in bonding techniques have increased the range and scope for use of ceramics in dentistry. In this brief review, we will discuss advances in ceramic materials and fabrication techniques. Examp...

  11. Measurement of properties of ceramic breeder materials for fusion reactor. Towards work function measurement under irradiation

    International Nuclear Information System (INIS)

    The 'High Temperature Kelvin Probe' was employed to study the effect of hydrogen on the surface of Li2O. The results revealed that the work function change was insensitive to the abrupt change of oxygen potential induced by H2 addition, in contrast to the results of Li4SiO4 as were previously obtained by the present authors. An attempt is now being made to modify the current system so as to be able to measure 'in-situ' the work function change caused by irradiation. As a preliminary step, an Au foil was irradiated by a proton beam by use of an accelerator, in order to evaluate the work function change of a reference electrode if the probe were exposed to irradiation. (author)

  12. Measurement of properties of ceramic breeder materials for fusion reactor. Towards work function measurement under irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Yamawaki, Michio; Yamaguchi, Kenji; Suzuki, Atsushi [Tokyo Univ. (Japan). Faculty of Engineering; Hayashi, Kimio

    2000-01-01

    The 'High Temperature Kelvin Probe' was employed to study the effect of hydrogen on the surface of Li{sub 2}O. The results revealed that the work function change was insensitive to the abrupt change of oxygen potential induced by H{sub 2} addition, in contrast to the results of Li{sub 4}SiO{sub 4} as were previously obtained by the present authors. An attempt is now being made to modify the current system so as to be able to measure 'in-situ' the work function change caused by irradiation. As a preliminary step, an Au foil was irradiated by a proton beam by use of an accelerator, in order to evaluate the work function change of a reference electrode if the probe were exposed to irradiation. (author)

  13. Maximizing fluence rate and field uniformity of light blanket for intraoperative PDT

    OpenAIRE

    LIANG, XING; Kundu, Palak; Finlay, Jarod; Goodwin, Michael; Zhu, Timothy C.

    2012-01-01

    A light blanket is designed with a system of cylindrically diffusing optical fibers, which are spirally oriented. This 25×30 cm rectangular light blanket is capable of providing uniform illumination during intraoperative photodynamic therapy. The flexibility of the blanket proves to be extremely beneficial when conforming to the anatomical structures of the patient being treated. Previous tests of light distribution from the blanket have shown significant loss of intensity with the length of ...

  14. Neutronic analysis of a dual He/LiPb coolant breeding blanket for DEMO

    OpenAIRE

    Catalán, J.P.; Ogando Serrano, Francisco; Sanz Gonzalo, Javier; Palermo, I.; Veredas, G.; Gómez Ros, J. M.; Sedano, L

    2010-01-01

    A conceptual design of a DEMO fusion reactor is being developed under the Spanish Breeding Blanket Technology Programme: TECNO_FUS based on a He/LiPb dual coolant blanket as reference design option. The following issues have been analyzed to address the demonstration of the neutronic reliability of this conceptual blanket design: power amplification capacity of the blanket, tritium breeding capability for fuel self-sufficiency, power deposition due to nuclear heating in superconducting coils ...

  15. APT Blanket Safety Analysis: Counter Current Flow Limitation for Cavity Spaces

    International Nuclear Information System (INIS)

    The thermal-hydraulic modeling aspects for the APT blanket system have been broken up into two basic modeling components: (1) the blanket system and (2) the cavity flood system. In most cases these systems are modeled separately. This separate study for the coolability of the blanket modules can also be used to establish/evaluate a functional design requirement on gap size between the blanket modules

  16. 75 FR 60095 - Sempra LNG Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-09-29

    ... LNG Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY..., by Sempra LNG Marketing, LLC (Sempra), requesting blanket authorization to export up to a total of... Order No. 2795, which granted Cheniere Marketing, LLC (Cheniere) blanket authorization to...

  17. Charge breeder for the SPIRAL1 upgrade: Preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Maunoury, L., E-mail: maunoury@ganil.fr; Delahaye, P.; Dubois, M.; Bajeat, O.; Frigot, R.; Jeanne, A.; Jardin, P.; Kamalou, O.; Lecomte, P.; Osmond, B.; Peschard, G.; Savalle, A. [GANIL, Bd H. Becquerel BP 55027, F-14076 Caen Cedex 05 (France); Angot, J.; Sole, P.; Lamy, T. [LPSC - Université Grenoble Alpes - CNRS/IN2P3, 53 rue des Martyrs, F-38026 Grenoble Cedex (France); Barton, C. [Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom)

    2016-02-15

    In the framework of the SPIRAL1 upgrade under progress at the GANIL lab, the charge breeder based on a LPSC Phoenix ECRIS, first tested at ISOLDE has been modified to benefit of the last enhancements of this device from the 1+/n+ community. The modifications mainly concern the 1 + optics, vacuum techniques, and the RF—buffer gas injection into the charge breeder. Prior to its installation in the midst of the low energy beam line of the SPIRAL1 facility, it has been decided to qualify its performances and several operation modes at the test bench of LPSC lab. This contribution shall present preliminary results of experiments conducted at LPSC concerning the 1 + to n+ conversion efficiencies for noble gases as well as for alkali elements and the corresponding transformation times.

  18. Feasibility study on the thorium fueled boiling water breeder reactor

    International Nuclear Information System (INIS)

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  19. Role of the breeder in long-term energy economics

    International Nuclear Information System (INIS)

    Private and public decisions affecting the use of nuclear and other energy technologies over a long-run time horizon were studied using the ETA-MACRO model which provides for economic- and energy-sector interactions. The impact on the use of competing energy technologies of a public decision to apply benefit-cost analysis to the production of carbon dioxide that enters the atmosphere is considered. Assuming the public choice is to impose an appropriate penalty tax on those technologies which generate CO2 and to allow decentralized private decisions to choose the optimal mix of energy technologies that maximize a nonlinear objective function subject to constraints, the study showed that breeder technology provides a much-larger share of domestically consumed energy. Having the breeder technology available as a substitute permits control of CO2 without significant reductions in consumption or gross national product growth paths

  20. Fast breeder reactors: Experience and trends. V. 2

    International Nuclear Information System (INIS)

    The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium was attended by almost 400 participants (340 participants, 35 observers and 20 journalists) from 25 countries and five international organizations. More than 80 papers were presented and discussed during six regular sessions and four poster sessions. A separate abstract was prepared for each of these papers

  1. 75 FR 11557 - Woven Electric Blankets From China

    Science.gov (United States)

    2010-03-11

    ... permitted by section 201.8 of the Commission's rules, as amended, 67 FR 68036 (November 8, 2002). Even where... specified in II (C) of the Commission's Handbook on Electronic Filing Procedures, 67 FR 68168, 68173... COMMISSION Woven Electric Blankets From China AGENCY: United States International Trade Commission....

  2. 18 CFR 33.1 - Applicability, definitions, and blanket authorizations.

    Science.gov (United States)

    2010-04-01

    ... the outstanding voting securities; or (iii) Any security of a subsidiary company within the holding... company subsidiary in connection with such acquisition. (4) A holding company granted blanket... subsidiaries, or associate companies within the holding company system has captive customers in the...

  3. Greater flamingos Phoenicopterus roseus are partial capital breeders

    OpenAIRE

    Rendón-Martos, Manuel; Rendón, Miguel A.; Garrido, Araceli; Amat, Juan A.

    2011-01-01

    Capital breeding refers to a strategy in which birds use body stores for egg formation, whereas income breeders obtain all resources for egg formation at breeding sites. Capital breeding should occur more in large-bodied species because the relative cost of carrying stores for egg formation becomes smaller with increasing body size. Based on a comparison between stable isotopes of carbon and nitrogen in potential prey at wintering sites and eggs, we examined whether greater flamingos use nutr...

  4. Longitudinal course of extrinsic allergic alveolitis in pigeon breeders.

    OpenAIRE

    Bourke, S. J.; Banham, S W; Carter, R; P. Lynch; Boyd, G.

    1989-01-01

    The purpose of this study was to assess the longitudinal course of pigeon breeders' disease by evaluating 24 patients with the acute form of the disease 10 years after their original diagnosis. Twenty one patients attended for clinical assessment, pulmonary function studies, chest radiography, and antibody measurement. Eighteen had continued to keep pigeons, emphasising their commitment to the hobby. Despite continued antigen exposure pigeon related symptoms had improved in most patients and ...

  5. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    The twenty-second Annual Meeting of the International Working Group on Fast Reactors took place in Vienna, 18-21 April 1989. Nineteen representatives from twelve Member States and International Organizations attended the Meeting. This publication is a collection of presentations in which the participants reported the status of their national programmes on fast breeder reactors. A separate abstract was prepared for each of the twelve papers from this collections. Refs, figs, tabs and 1 graph

  6. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    The present document contains information on the status of fast breeder reactor development and on worldwide activities in this advanced nuclear power technology during 1989 as reported at the 23rd meeting of the IWGFR in Vienna, April 1990. The publication is intended to provide information regarding the current status of LMFBR development in IAEA Member States. A separate abstract was prepared for each of the 11 papers presented by the participants of this meeting. Refs, figs and tabs

  7. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    The present document contains information on the status of fast breeder reactor development and on worldwide activities in this advanced nuclear power technology during 1990 as reported at the 24th meeting of the IWGFR in Tsuruga, Japan, 15-18 April 1991. The publication is intended to provide information regarding the current status of LMFBR development in IAEA Member States and CEC. Figs and tabs

  8. Modeling of tritium transport in lithium aluminate fusion solid breeders

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C.; Clemmer, R.G.

    1985-02-01

    Lithium aluminate is a candidate tritium-breeding material for fusion reactor blankets. One of the concerns with using LiAlO/sub 2/ is tritium recovery from this material, particularly at low operating temperatures and high fluences. The data from various tritium release experiments with ..gamma..-LiAlO/sub 2/ and related materials are reviewed and analyzed to determine under what conditions bulk diffusion is the rate-limiting mechanism for tritium transport and what the effective bulk diffusion coefficient should be. Steady-state and transient models based on bulk diffusion are developed and used to interpret the data. Design calculations are then performed with the verified models to determine the steady-state inventory and time to reach equilibrium for a full-scale fusion blanket.

  9. Integrated-blanket-coil (IBC) concept applied to the poloidal field and blanket systems of a tokamak reactor

    International Nuclear Information System (INIS)

    A novel concept is proposed for combining the blanket and coil functions of a fusion reactor into a single component. This concept, designated the integrated-blanket-coil (IBC) concept, is applied to the poloidal field and blanket systems of a Tokamak reactor. An examination of resistive power losses in the IBC suggests that these losses can be limited to less than or equal to 10% of the fusion thermal power. By assuming a sandwich construction for the IBC walls, MHD-induced pressure drops and associated pressure stresses are shown to be modest and well below design limits. For the stainless steel reference case examined in this paper, the MHD-induced pressure drop was estimated to be approx. 1/3 MPa and the associated primary membrane stress was estimated to be approx. 47 MPa. The preliminary analyses presented in this paper indicate that the IBC concept offers promise as a means for making fusion reactors more compact by combining blanket and coils functions in a single component

  10. Activation analyses for the Korea helium cooled ceramic reflector test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Cheol Woo, E-mail: cwl@kaeri.re.kr [Korea Atomic Energy Research Institute, 989 Daeduck-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Lee, Young-Ouk [Korea Atomic Energy Research Institute, 989 Daeduck-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Ahn, Mu-Young; Cho, Seungyon [National Fusion Research Institute, Gwahangno, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, 989 Daeduck-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2013-10-15

    The activation analyses were performed to obtain activities, inventories and dose rates after shutdown for the Korea HCCR TBM. MonteBurns code coupled with MCNP and CINDER codes was used in the activation calculation and the irradiation history based on the SA2 scenario was applied. The total activity in TBM after shutdown was evaluated as a value of 4.12 × 10{sup 16} Bq. Dose rates after shutdown from the activated HCCR TBM was estimated. The decay gamma-ray spectra in each region inside TBM were estimated based on the calculated activities. The dose rate at 0 cm–3 m from the FW surface of HCCR TBM was evaluated according to the cooling time of 1 day, 1 week, 1 month, 6 months and 1 year. The dose rate at 0 cm from the FW surface was evaluated to be 575.84 Sv/h at 1 day after shutdown.

  11. Current status of technology development for fabrication of Indian Test Blanket Module (TBM) of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Jayakumar, T., E-mail: tjk@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam 603102 (India); Rajendra Kumar, E. [TBM Division, Institute for Plasma Research (IPR), Bhat, Gandhinagar 382428 (India)

    2014-10-15

    Highlights: • Status of technology developments for Indian TBM to be installed in ITER is presented. • Procedure development for EB, laser and laser-hybrid welding of RAFM steel presented. • Filler wires for RAFM steel for TIG, NG-TIG and laser-hybrid welding have been developed. • Feasibility of production of channel plate by HIP technology has been demonstrated. - Abstract: Ever since India decided to install its Lead-Lithium Ceramic Breeder (LLCB) TBM in ITER, various technologies for fabrication of Indian TBM are being pursued by IPR and IGCAR, in collaboration with various research laboratories in India. Welding consumables for joining India specific RAFM steels (IN-RAFMS), procedures for hot isostatic pressing, electron beam welding, laser and laser-hybrid welding have been developed. Considering the complex nature and limited access available for inspection, innovative inspection procedures that involved use of phased array ultrasonic and C-scan imaging are also being pursued. This paper presents the current status of these developments and provides a roadmap for the future activities planned in realizing Indian TBM for testing in ITER.

  12. Ceramic Laser Materials

    Directory of Open Access Journals (Sweden)

    Guillermo Villalobos

    2012-02-01

    Full Text Available Ceramic laser materials have come a long way since the first demonstration of lasing in 1964. Improvements in powder synthesis and ceramic sintering as well as novel ideas have led to notable achievements. These include the first Nd:yttrium aluminum garnet (YAG ceramic laser in 1995, breaking the 1 KW mark in 2002 and then the remarkable demonstration of more than 100 KW output power from a YAG ceramic laser system in 2009. Additional developments have included highly doped microchip lasers, ultrashort pulse lasers, novel materials such as sesquioxides, fluoride ceramic lasers, selenide ceramic lasers in the 2 to 3 μm region, composite ceramic lasers for better thermal management, and single crystal lasers derived from polycrystalline ceramics. This paper highlights some of these notable achievements.

  13. Fatigue of dental ceramics

    OpenAIRE

    Zhang, Yu; Sailer, Irena; lawn, brian

    2013-01-01

    Clinical data on survival rates reveal that all-ceramic dental prostheses are susceptible to fracture from repetitive occlusal loading. The objective of this review is to examine the underlying mechanisms of fatigue in current and future dental ceramics

  14. Ceramic art in sculpture

    OpenAIRE

    Rokavec, Eva

    2014-01-01

    Diploma seminar speaks of ceramics as a field of artistic expression and not just as pottery craft. I presented short overview of developing ceramic sculpture and its changing role. Clay inspires design and touch more than other sculpture media. It starts as early as in prehistory. Although it sometimes seems that was sculptural ceramics neglected in art history overview, it was not so in actual praxis. There is a rich tradition of ceramics in the East and also in Europe during the renaissanc...

  15. Ceramic Laser Materials

    OpenAIRE

    Guillermo Villalobos; Jasbinder Sanghera; Ishwar Aggarwal; Bryan Sadowski; Jesse Frantz; Colin Baker; Brandon Shaw; Woohong Kim

    2012-01-01

    Ceramic laser materials have come a long way since the first demonstration of lasing in 1964. Improvements in powder synthesis and ceramic sintering as well as novel ideas have led to notable achievements. These include the first Nd:yttrium aluminum garnet (YAG) ceramic laser in 1995, breaking the 1 KW mark in 2002 and then the remarkable demonstration of more than 100 KW output power from a YAG ceramic laser system in 2009. Additional developments have included highly doped microchip lasers,...

  16. Experiences with fast breeder reactor education in laboratory and short course settings

    International Nuclear Information System (INIS)

    The breeder reactor industry throughout the world has grown impressively over the last two decades. Despite the uncertainties in some national programs, breeder reactor technology is well established on a global scale. Given the magnitude of this technological undertaking, there has been surprisingly little emphasis on general breeder reactor education - either at the university or laboratory level. Many universities assume the topic too specialized for including appropriate courses in their curriculum - thus leaving students entering the breeder reactor industry to learn almost exclusively from on-the-job experience. The evaluation of four course presentations utilizing visual aids is presented

  17. Development of hi-tech ceramics fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Park, Ji Yeon; Kim, Sun Jai; Jung Choong Hwan; Oh, Seok Jin

    1997-07-01

    There are some ceramic materials being used in the nuclear energy such as nuclear fuel, coolant pump seals, tritium breeder materials, a high temperature absorber, and the solid electrolyte for recovering tritium. In addition, lots of researches recently have been conducted on the development of highly functional ceramics such as highly efficient shielding materials, functional graded materials and radioactive isotopes-separating materials. Therefore, one of the objectives of this project is to develop ultra-fine and pure powder manufacturing technology. Tritium breeder materials, LiAlO{sub 2}, Li{sub 2}ZrO{sub 3} and Li{sub 2}TiO{sub 3} were made with a combustion process of mixed fuels that is developed indigenously in this project. Additionally, this study also focused on the development of promising low temperature electrolytes of ceria. By using the ceria powder made by the combustion process of GNP was investigated their sinterability and the electrolytic characteristics. (author). 167 refs., 74 tabs., 91 figs

  18. Thin film ceramic thermocouples

    Science.gov (United States)

    Gregory, Otto (Inventor); Fralick, Gustave (Inventor); Wrbanek, John (Inventor); You, Tao (Inventor)

    2011-01-01

    A thin film ceramic thermocouple (10) having two ceramic thermocouple (12, 14) that are in contact with each other in at least on point to form a junction, and wherein each element was prepared in a different oxygen/nitrogen/argon plasma. Since each element is prepared under different plasma conditions, they have different electrical conductivity and different charge carrier concentration. The thin film thermocouple (10) can be transparent. A versatile ceramic sensor system having an RTD heat flux sensor can be combined with a thermocouple and a strain sensor to yield a multifunctional ceramic sensor array. The transparent ceramic temperature sensor that could ultimately be used for calibration of optical sensors.

  19. Engineering test station for TFTR blanket module experiments

    International Nuclear Information System (INIS)

    A conceptual design has been carried out for an Engineering Test Station (ETS) which will provide structural support and utilities/instrumentation services for blanket modules positioned adjacent to the vacuum vessel of the TFTR (Tokamak Fusion Test Reactor). The ETS is supported independently from the Test Cell floor. The ETS module support platform is constructed of fiberglass to eliminate electromagnetic interaction with the pulsed tokamak fields. The ETS can hold blanket modules with dimensions up to 78 cm in width, 85 cm in height, and 105 cm in depth, and with a weight up to 4000 kg. Interfaces for all utility and instrumentation requirements are made via a shield plug in the TFTR igloo shielding. The modules are readily installed or removed by means of TFTR remote handling equipment

  20. Ochratoxicosis in White Leghorn breeder hens: Production and breeding performance

    Directory of Open Access Journals (Sweden)

    Zahoor Ul Hassan*, Muhammad Zargham Khan, Ahrar Khan, Ijaz Javed1, Umer Sadique2 and Aisha Khatoon

    2012-10-01

    Full Text Available This study was designed to evaluate the effect of Ochratoxin A (OTA upon production and breeding parameters in White Leghorn (WL breeder hens. For this purpose, 84 WL breeder hens were divided into seven groups (A-G. The hens in these groups were maintained on feed contaminated with OTA @ 0.0 (control, 0.1, 0.5, 1.0, 3.0, 5.0 and 10.0 mg/Kg, respectively for 21 days. These hens were artificially inseminated with semen obtained from healthy roosters kept on OTA free feed. Egg production and their quality parameters were recorded. Fertile eggs obtained from each group were set for incubation on weekly basis. At the end of the experiment, hens in each group were killed to determined gross and microscopic lesions in different organs. OTA residue concentrations were determined in extracts of liver, kidneys and breast muscles by immunoaffinity column elution and HPLC-Fluorescent detection techniques. Feeing OTA contaminated diet resulted in a significant decrease in egg mass and egg quality parameters. Liver and kidneys showed characteristic lesions of ochratoxicosis. Residue concentration (ng/g of OTA in the hens fed 10 mg/kg OTA, was the highest in liver (26.336±1.16 followed by kidney (8.223±0.85 and were least in breast muscles (1.235±0.21. Embryonic mortalites were higher, while hatachabilites of the chicks were lower in the groups fed higher doses of OTA. Feeding OTA contaminated diets to breeder hen resulted in residues accumulation in their tissues along with significantly reduced production and breeding performance.

  1. Campylobacter epidemiology from breeders to their progeny in Eastern Spain.

    Science.gov (United States)

    Ingresa-Capaccioni, S; Jiménez-Trigos, E; Marco-Jiménez, F; Catalá, P; Vega, S; Marin, C

    2016-03-01

    While horizontal transmission is a route clearly linked to the spread of Campylobacter at the farm level, few studies support the transmission of Campylobacter spp. from breeder flocks to their offspring. Thus, the present study was carried out to investigate the possibility of vertical transmission. Breeders were monitored from the time of housing day-old chicks, then throughout the laying period (0 to 60 wk) and throughout their progeny (broiler fattening, 1 to 42 d) until slaughter. All samples were analyzed according with official method ISO 10272:2006. Results revealed that on breeder farms, Campylobacter isolation started from wk 16 and reached its peak at wk 26, with 57.0% and 93.2% of positive birds, respectively. After this point, the rate of positive birds decreased slightly to 86.0% at 60 wk. However, in broiler production all day-old chicks were found negative for Campylobacter spp, and the bacteria was first isolated at d 14 of age (5.0%), with a significant increase in detection during the fattening period with 62% of Campylobacter positive animals at the end of the production cycle. Moreover, non-positive sample was determined from environmental sources. These results could be explained because Campylobacter may be in a low concentration or in a non-culturable form, as there were several studies that successfully detected Campylobacter DNA, but failed to culture. This form can survive in the environment and infect successive flocks; consequently, further studies are needed to develop more modern, practical, cost-effective and suitable techniques for routine diagnosis.

  2. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    Baldev Raj; S L Mannan; P R Vasudeva Rao; M D Mathew

    2002-10-01

    Fast breeder reactors (FBRs) are destined to play a crucial role inthe Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of fissile material than in thermal reactors, with a matching increase in burn-up. The design of the fuel is an important aspect which has to be optimised for efficient, economic and safe production of power. FBR components operate under hostile and demanding environment of high neutron flux, liquid sodium coolant and elevated temperatures. Resistance to void swelling, irradiation creep, and irradiation embrittlement are therefore major considerations in the choice of materials for the core components. Structural and steam generator materials should have good resistance to creep, low cycle fatigue, creep-fatigue interaction and sodium corrosion. The development of carbide fuel and structural materials for the Fast Breeder Test Reactor at Kalpakkam was a great technological challenge. At the Indira Gandhi Centre for Atomic Research (IGCAR), advanced research facilities have been established, and extensive studies have been carried out in the areas of fuel and materials development. This has laid the foundation for the design and development of a 500 MWe Prototype Fast Breeder Reactor. Highlights of some of these studies are discussed in this paper in the context of our mission to develop and deploy FBR technology for the energy security of India in the 21st century.

  3. Fuel Cycle Economics of Fast Breeders with Plutonium

    International Nuclear Information System (INIS)

    Pu-fuelled fast breeder systems are characterized by their attractive fuel cycle economics. Basically, the economics are influenced by a number of reactor parameters like fissile material rating, fuel bum-up, breeding ratio and thermal efficiency, on the one hand, and by a number of economic parameters like the plutonium price, the interest rate and the fabrication and reprocessing costs on the other. To a certain extent, the two sets of parameters are interdependent and the cost parameters are influenced by the existing nuclear industry as well. In the present paper it is shown, with the help of a number of specific examples, that the fuel cycle of Pu fast breeders is not a static and isolated property of the reactor but is dynamic in nature. Depending on the cost situation and other conditions, the fuel cycle can always be optimized anew to fit into the existing overall economics. A high Pu price, for example, requires a high fissile rating or a high breeding ratio, whereas, if the Pu price falls, neither a high rating nor a high breeding ratio is necessary to keep the fuel cycle costs low. The influence of fabrication costs may be regulated to some extent by varying the burn-up. The effect of reprocessing costs may be made comparatively insignificant provided reprocessing can be carried out in large centrally located multi-purpose plants for converter elements. Because of the high flexibility of the fuel cycle of Pu fast breeders, the attractiveness of their fuel cycle economics can be retained under a wide range of competitive conditions. (author)

  4. MFTF-B Upgrade for blanket-technology testing

    International Nuclear Information System (INIS)

    Based on preliminary studies at Lawrence Livermore National Laboratory (LLNL), we believe the Mirror Fusion Test Facility (MFTF-B) could be upgraded for operation in a hot-ion Kelley mode in a portion of the central cell to provide fusion nuclear engineering data, particularly blanket technology information, by the end of the decade. Cost of this mode of operation would be modest compared with that of the other fusion devices considered in the last few years for such purposes

  5. Helium-3 blankets for tritium breeding in fusion reactors

    Science.gov (United States)

    Steiner, Don; Embrechts, Mark; Varsamis, Georgios; Vesey, Roger; Gierszewski, Paul

    1988-01-01

    It is concluded that He-3 blankets offers considerable promise for tritium breeding in fusion reactors: good breeding potential, low operational risk, and attractive safety features. The availability of He-3 resources is the key issue for this concept. There is sufficient He-3 from decay of military stockpiles to meet the International Thermonuclear Experimental Reactor needs. Extraterrestrial sources of He-3 would be required for a fusion power economy.

  6. Model problem of MHD flow in a lithium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Cherepanov, V.Y.

    1978-01-01

    A model problem is considered for a feasibility study concerning controlled MHD flow in the blanket of a Tokamak nuclear reactor. The fundamental equations for the steady flow of an incompressible viscous fluid in a uniform transverse magnetic field are solved in rectangular coordinates, in the zero-induction approximation and with negligible induced currents. A numerical solution obtained for a set of appropriate boundary constraints establishes the conditions under which no stagnation zones will be formed.

  7. Instrumentation and control improvements at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I C systems of the next generation of liquid metal reactor (LMR) plants.

  8. Elements for evaluation of fast breeder reactor's potential in Argentina

    International Nuclear Information System (INIS)

    Fast Breeder Reactors (FBR) main features are presented in a general form, including their physical principles, the history of their evolution, their relevant technological aspects and the basis for their comparison to other energy sources. This is completed with descriptions of typical reactors and a model of FBR penetration in the Argentine electrical network. It is recommended to form a multidisciplinary board to study which position should be taken with respect to this type of reactors. In the author's opinion a Research activity should be started and gradually increased for passing to Development activities after a short while. (Author)

  9. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    G Pandikumar; V Gopalakrishnan; P Mohanakrishnan

    2009-05-01

    In a thermal neutron reactor, multiple recycle of U–Pu fuel is not possible due to degradation of fissile content of Pu in just one recycle. In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder reactor under construction at Kalpakkam. After about five recycles, the cycle-to-cycle variation in the above parameters is below 1%.

  10. Reliability modeling of Clinch River breeder reactor electrical shutdown systems

    International Nuclear Information System (INIS)

    The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

  11. Application of an LP model to breeder strategy studies

    International Nuclear Information System (INIS)

    The paper discusses the relationships between the capital cost differential (FBR--LWR) allowable for economic breeder introduction and energy demand, resource availability (through price--quantity schedule), and economic environment for a range of future projections. The ALPS linear programming reactor systems analysis code, developed by Hanford Engineering Development Laboratory, was used for economic optimizations where they were done, and where they were not it provided a useful tool to compute the discounted total system power cost over the planning horizon for a given set of reactor mix and cost parameters

  12. Instrumentation and control improvements at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, L.J.; Planchon, H.P.

    1993-03-01

    The purpose of this paper is to describe instrumentation and control (I&C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I&C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I&C systems of the next generation of liquid metal reactor (LMR) plants.

  13. Binary breeder reactor: an option for Brazilian energy future

    International Nuclear Information System (INIS)

    To assure a continued supply of electric energy beyond a few decades from now, developmemnt of fast breeder reactors is a necessity. Binary fueled LMFBRs combine an improvement in the inherent safety of fast reactors and an efficient use of the abundant thorium. A nuclear system that starts with PWRs and gradually shifts to a FBR system or to a FBR-PWR symbiotic system appears to be the most reasonable one for Brazil. Natural uranium requirements and possible sequences of reactor introductions are discussed for some postulated Brazilian situations. A permanent system of approx. 100 GWe capacity can be established based on the estimated resource of natural uranium. (Author)

  14. Binary breeder reactor an option for Brazilian energy future

    International Nuclear Information System (INIS)

    To assure a continued supply of electric energy beyond a few decades from now, development of fast breeder reactors is a necessity. Binary fueled LMFBRs combine an improvement in the inherent safety of fast reactors and an efficient use of the abundant thorium. A nuclear system that starts with PWRs and gradually shifts to a FBR system or to a FBR-PWR symbiotic system appears to be the most resonable one for Brazil. Natural uranium requirements and possible sequences of reactor introductions are discussed for some postulated Brazilian situations. A permanent system of approximatelly 100 GWe capacity can be established based on the estimated resource of natural uranium. (Author)

  15. Investigation of aqueous slurries as fusion reactor blankets

    International Nuclear Information System (INIS)

    Numerical and experimental studies were carried out to assess the feasibility of using an aqueous slurry, with lithium in its solid component, to meet the tritium breeding, cooling, and shielding requirements of a controlled thermonuclear reactor (CTR). The numerical studies were designed to demonstrate the theoretical ability of a conceptual slurry blanket to breed adequate tritium to sustain the CTR. The experimental studies were designed to show that the tritium retention characteristics of likely solid components for the slurry were conducive to adequate tritium recovery without the need for isotopic separation. The numerical portion of this work consisted in part of using ANISN, a one-dimensional finite difference neutron transport code, to model the neutronic performance of the slurry blanket concept. The parameters governing tritium production and retention in a slurry were computed and used to modify the results of the ANISN computer runs. The numerical work demonstrated that the slurry blanket was only marginally capable of breeding sufficient tritium without the aid of a neutron multiplying region. The experimental portion of this work consisted of several neutron irradiation experiments, which were designed to determine the retention abilities of LiF particles

  16. Manufacturing and characterization of porous SiC for flow channel inserts in dual-coolant blanket designs

    Energy Technology Data Exchange (ETDEWEB)

    Bereciartu, Ainhoa [CEIT and Tecnun (University of Navarra), Manuel de Lardizabal 15, 20018 San Sebastian (Spain); Ordas, Nerea, E-mail: nordas@ceit.es [CEIT and Tecnun (University of Navarra), Manuel de Lardizabal 15, 20018 San Sebastian (Spain); Garcia-Rosales, Carmen [CEIT and Tecnun (University of Navarra), Manuel de Lardizabal 15, 20018 San Sebastian (Spain); Morono, Alejandro; Malo, Marta; Hodgson, Eric R. [CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Abella, Jordi [Institut Quimic de Sarria, University Ramon Llull, Via Augusta 390, 08017 Barcelona (Spain); Sedano, Luis [CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain)

    2011-10-15

    SiC is the primary candidate for the flow channel inserts in dual-coolant blanket concepts. Porous SiC ceramics are attractive candidates for this non-structural application, since they can satisfy the required properties through a low cost manufacturing route, compared to SiC{sub f}/SiC. This work shows first results of the manufacturing of porous SiC ceramics prepared with different amounts of Y{sub 2}O{sub 3} and Al{sub 2}O{sub 3} as sintering additives. C powders were used as pore-formers by their burnout during oxidation after sintering. Comparison of microstructure, porosity, flexural strength, thermal and electrical conductivity and corrosion under Pb-15.7Li of porous SiC without and with sintering additives is presented. The addition of 2.5 wt.% of Y{sub 2}O{sub 3} and Al{sub 2}O{sub 3} improves the mechanical properties, and reduces the thermal and electrical conductivity down to reasonable values. Preliminary corrosion tests under Pb-15.7 Li at 500 deg. C show that the absence of a dense coating on porous SiC leads to poor corrosion behavior.

  17. APT Blanket Safety Analysis: Preliminary Analyses of Downflow Through a Lateral Row 1 Blanket Model Under Near RHR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    To address a concern about a potential maldistribution of coolant flow through an APT blanket module under low flow near RHR conditions, a scoping study of downflow mixed convection in parallel channels was conducted. Buoyancy will adversely effect the flow distribution in module bins with downflow and non-uniform power distributions. This study consists of two parts: a simple analytical model of flow in a two channel network, and a lumped eleven channel FLOWTRAN-TF model of a front lateral Row-1 blanket module bin. Results from both models indicate that the concern about coolant flow in a vertical model being diverted away from high power regions by buoyancy is warranted. The FLOWTRAN-TF model predicted upflow (i.e., a flow reversal) through several of the high power channels, under some low flow conditions. The transition from the regime with downflow in all channels to a regime with upflow in some channels was abrupt.

  18. Fast breeder reactors: experience and trends. V. 1

    International Nuclear Information System (INIS)

    The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium presentations were divided into sessions devoted to the following topics: Experience of LMFBR construction and operation and resultant development strategies (6 papers); LMFBR plant startup and commissioning tests and general behaviour (8 papers); Core performance experience for high burnup and core design trends (8 papers); Experience and trends in the LMFBR fuel cycle (4 papers); Core design and behaviour (3 papers); Fuels and materials (7 papers). A separate abstract was prepared for each of these papers

  19. Breeder reactors: a technique at the service of humanity

    International Nuclear Information System (INIS)

    A genuine energy policy is not conceived purely for a short term. It must on the contrary take into consideration many national and international facts in order to arrive at a balance which takes into account both the interests of the country where it is to be applied and the future interests of humanity. Growth and energy consumption make a pair. Considering the forecasts of future consumption, a rational utilization of the energy sources is a priority. The rational utilization of the energy potentialities of uranium takes a prominent place in this priority. In the fission energy of the atoms, the breeder reactors are the only types which can give their full meanings to the words economy, ecology, rationality etc. In calling for innovation, the breeder reactors are the prime movers for an advanced industry and a guarantee for the future penetration of electricity in many fields. They are thus important elements for the creation of employment. This paper also deals with questions of international cooperation, non-proliferation and the necessity for disarmament

  20. Axial blanket fuel design and demonstration. First semi-annual progress report, January-September 1980

    International Nuclear Information System (INIS)

    The axial blanket fuel design in this program, which is retrofittable in operating pressurized water reactors, involves replacing the top and bottom of the enriched fuel column with low-enriched (less than or equal to 1.0 wt % 235U) fertile uranium. This repositioning of the fissile inventory in the fuel rod leads to decreased axial leakage and increased discharge burnups in the enriched fuel. Various axial blanket fuel designs, with blanket thicknesses from 0 to 10 inches and blanket enrichments from 0.2 to 1.0 wt % 235U, were investigated to determine the relationship between uranium utilization and power peaking. Analyses were preformed to assess the nuclear, mechanical, and thermal-hydraulic effects arising from the use of axial blankets. Four axial blanket lead test assemblies are being fabricated for scheduled irradiation in cycle 5 of Sacramento Municipal Utility District's Rancho Seco pressurized water reactor. Analyses to support licensing cycle 5 are in progress

  1. Ceramic tamper-revealing seals

    Science.gov (United States)

    Kupperman, David S.; Raptis, Apostolos C.; Sheen, Shuh-Haw

    1992-01-01

    A flexible metal or ceramic cable with composite ceramic ends, or a u-shaped ceramic connecting element attached to a binding element plate or block cast from alumina or zirconium, and connected to the connecting element by shrink fitting.

  2. Analyses of fine paste ceramics

    International Nuclear Information System (INIS)

    Four chapters are included: history of Brookhaven fine paste ceramics project, chemical and mathematical procedures employed in Mayan fine paste ceramics project, and compositional and archaeological perspectives on the Mayan fine paste ceramics

  3. Analyses of fine paste ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Sabloff, J A [ed.

    1980-01-01

    Four chapters are included: history of Brookhaven fine paste ceramics project, chemical and mathematical procedures employed in Mayan fine paste ceramics project, and compositional and archaeological perspectives on the Mayan fine paste ceramics. (DLC)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  5. Conceptual study on high performance blanket in a spherical tokamak fusion-driven transmuter

    International Nuclear Information System (INIS)

    A preliminary conceptual design on high performance dual-cooled blanket of fusion-driven transmuter is presented based on neutronic calculation. The dual-cooled system has some attractive advantages when utilized in transmutation of HLW (High Level Wastes). The calculation results show that this kind of blanket could safely transmute about 6 ton minor actinides (produced by 170 GW(e) Year PWRs approximately) and 0.4 ton fission products per year, and output 12 GW thermal power. In addition, the variation of power and critical factor of this blanket is relatively little during its 1-year operation period. This blanket is also tritium self-sustainable

  6. Hazard report update. ECRI Institute revises its recommendation for temperature limits on blanket warmers.

    Science.gov (United States)

    2009-07-01

    ECRI Institute now recommends that temperature settings on blanket warming cabinets be limited to 130 degrees F (54 degrees C). We had previously recommended a limit of 110 degrees F (43 degrees C) because solutions were often being warmed in the same cabinets as blankets, and the lower temperature eliminated the serious burn risk presented by excessively heated solutions. With increasing recognition in the healthcare community that solutions should be kept at lower temperatures than--and therefore heated separately from--blankets, we believe that our recommendation for blankets can be made less stringent. We continue to recommend that solution warming cabinets be limited to 110 degrees F. PMID:20848953

  7. Continuous Fiber Ceramic Composites

    Energy Technology Data Exchange (ETDEWEB)

    None

    2002-09-01

    Fiber-reinforced ceramic composites demonstrate the high-temperature stability of ceramics--with an increased fracture toughness resulting from the fiber reinforcement of the composite. The material optimization performed under the continuous fiber ceramic composites (CFCC) included a series of systematic optimizations. The overall goals were to define the processing window, to increase the robustinous of the process, to increase process yield while reducing costs, and to define the complexity of parts that could be fabricated.

  8. Tritium recovery from helium purge stream of solid breeder blanket by cryogenic molecular sieve bed. 2. Regeneration operation of cryogenic molecular sieve bed

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Yoshinori; Enoeda, Mikio; Nishi, Masataka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Regeneration operation is a very important operation, because it is the most influential factor for deciding the net operation cycle time and the minimum dimension of Cryogenic Molecular Sieve Bed (CMSB). However, the experimental data of CMSB regeneration operation was not so sufficient that even the optimum regeneration procedure could not be decided yet. This work was focused on getting the primary information about various regeneration procedures. (author)

  9. Measuring Fracture Times Of Ceramics

    Science.gov (United States)

    Shlichta, Paul J.; Bister, Leo; Bickler, Donald G.

    1989-01-01

    Electrical measurements complement or replace fast cinematography. Electronic system measures microsecond time intervals between impacts of projectiles on ceramic tiles and fracture tiles. Used in research on ceramics and ceramic-based composite materials such as armor. Hardness and low density of ceramics enable them to disintegrate projectiles more efficiently than metals. Projectile approaches ceramic tile specimen. Penetrating foil squares of triggering device activate display and recording instruments. As ceramic and resistive film break oscilloscope plots increase in electrical resistance of film.

  10. Defect production in ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J. [Oak Ridge National Lab., TN (United States); Kinoshita, C. [Kyushu Univ. (Japan)

    1997-08-01

    A review is given of several important defect production and accumulation parameters for irradiated ceramics. Materials covered in this review include alumina, magnesia, spinel silicon carbide, silicon nitride, aluminum nitride and diamond. Whereas threshold displacement energies for many ceramics are known within a reasonable level of uncertainty (with notable exceptions being AIN and Si{sub 3}N{sub 4}), relatively little information exists on the equally important parameters of surviving defect fraction (defect production efficiency) and point defect migration energies for most ceramics. Very little fundamental displacement damage information is available for nitride ceramics. The role of subthreshold irradiation on defect migration and microstructural evolution is also briefly discussed.

  11. The impact of blanket design on activation and thermal safety

    International Nuclear Information System (INIS)

    Activation and thermal safety analyses for experimental and power reactors are presented. The effects of a strong neutron absorber, B4C, on activation and temperature response of experimental reactors to Loss-of-Cooling Accidents are investigated. Operational neutron fluxes, radioactivities of elements and thermal transients are calculated using the codes ONEDANT, REAC and THIOD, respectively. The inclusion of a small amount of B4C in the steel blanket of an experimental reactor reduces its activation and the post LOCA temperature escalation significantly. Neither the inclusion of excessive amounts of B4C nor enriched 10B in the first walls of an experimental reactor bring much advantage. The employment of a 2 cm graphite tile liner before the first wall helps to limit the post LOCA escalation of first wall temperature. The effect of replacing a 20 cm thick section of a steel shield of a fusion power reactor with B4C is also analyzed. The first wall temperature peak is reduced by 100 degree C in the modified blanket. The natural convection effect on thermal safety of a liquid lithium cooled blanket are investigated. Natural convection has no impact at all, unless the magnetic field can be reduced. If magnets can be shut off rapidly after the accident, then the temperature escalation of the first wall will be limited. Upflow of the coolant is better than the initial downflow design from a thermal safety point of view. Activities of three structural materials, OTR stainless steel, SS-316 and VCrTi are compared. Although VCrTi has higher activity for a period of two hours after the accident, it has one to two orders of magnitude less activity than those of the steels in the mid- and long-terms. 29 refs., 42 figs., 9 tabs

  12. Impact of prescribed burning on blanket peat hydrology

    Science.gov (United States)

    Holden, Joseph; Palmer, Sheila M.; Johnston, Kerrylyn; Wearing, Catherine; Irvine, Brian; Brown, Lee E.

    2015-08-01

    Fire is known to impact soil properties and hydrological flow paths. However, the impact of prescribed vegetation burning on blanket peatland hydrology is poorly understood. We studied 10 blanket peat headwater catchments. Five were subject to prescribed burning, while five were unburnt controls. Within the burnt catchments, we studied plots where the last burn occurred ˜2 (B2), 4 (B4), 7 (B7), or greater than 10 years (B10+) prior to the start of measurements. These were compared with plots at similar topographic wetness index locations in the control catchments. Plots subject to prescribed vegetation burning had significantly deeper water tables (difference in means = 5.3 cm) and greater water table variability than unburnt plots. Water table depths were significantly different between burn age classes (B2 > B4 > B7 > B10+) while B10+ water tables were not significantly different to the unburnt controls. Overland flow was less common on burnt peat than on unburnt peat, recorded in 9% and 17% of all runoff trap visits, respectively. Storm lag times and hydrograph recession limb periods were significantly greater (by ˜1 and 13 h on average, respectively) in the burnt catchments overall, but for the largest 20% of storms sampled, there was no significant difference in storm lag times between burnt and unburnt catchments. For the largest 20% of storms, the hydrograph intensity of burnt catchments was significantly greater than those of unburnt catchments (means of 4.2 × 10-5 and 3.4 × 10-5 s-1, respectively), thereby indicating a nonlinear streamflow response to prescribed burning. Together, these results from plots to whole river catchments indicate that prescribed vegetation burning has important effects on blanket peatland hydrology at a range of spatial scales.

  13. Self-Assembling, Flexible, Pre-Ceramic Composite Preforms

    Science.gov (United States)

    Jaskowiak, Martha H.; Eckel, Andrew J.; Gorican, Daniel

    2009-01-01

    In this innovation, light weight, high temperature, compact aerospace structures with increased design options are made possible by using self-assembling, flexible, pre-ceramic composite materials. These materials are comprised of either ceramic or carbon fiber performs, which are infiltrated with polymer precursors that convert to ceramics upon thermal exposure. The preform architecture can vary from chopped fibers formed into blankets or felt, to continuous fibers formed into a variety of 2D or 3D weaves or braids. The matrix material can also vary considerably. For demonstration purposes, a 2D carbon weave was infiltrated with a SiC polymer precursor. The green or unfired material is fabricated into its final shape while it is still pliable. It is then folded or rolled into a much more compact shape, which will occupy a smaller space. With this approach, the part remains as one continuous piece, rather than being fabricated as multiple sections, which would require numerous seals for eventual component use. The infiltrated preform can then be deployed in-situ. The component can be assembled into its final shape by taking advantage of the elasticity of the material, which permits the structure to unfold and spring into its final form under its own stored energy. The pre-ceramic composites are converted to ceramics and rigidized immediately after deployment. The final ceramic composite yields a high-temperature, high-strength material suitable for a variety of aerospace structures. The flexibility of the material, combined with its high-temperature structural capacity after rigidization, leads to a less complex component design with an increased temperature range. The collapsibility of these structures allows for larger components to be designed and used, and also offers the potential for increased vehicle performance. For the case of collapsible nozzle extensions, a larger nozzle, and thus a larger nozzle exit plane, is possible because interference with

  14. Recovery of tritium from a liquid lithium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Talbot, J.B.

    1981-01-01

    The sorption of tritium on yttrium from liquid lithium and the subsequent release of tritium from yttrium by thermal regeneration of the metal sorbent were investigated to study such a tritium-recovery process for a fusion reactor blanket of liquid lithium. Recent static sorption experiments have shown the effects of lithium temperature and possible impurities on the sorption of tritium. Diffusivity data, obtained from previous tritium recovery experiments, were evaluated to show the importance of the yttrium surface condition in controlling the release of tritium.

  15. Magnetohydrodynamic research in fusion blanket engineering and metallurgical processing

    International Nuclear Information System (INIS)

    A review of recent research activities in liquid metal magnetohydrodynamics (LM-MHDs) is presented in this article. Two major reserach areas are discussed. The first topic involves the thermomechanical design issues in a proposed tokamak fusion reactor. The primary concerns are in the magneto-thermal-hydraulic performance of a self-cooled liquid metal blanket. The second topic involves the application of MHD in material processing in the metallurgical and semiconductor industries. The two representative applications are electromagnetic stirring (EMS) of continuously cast steel and the Czochralski (CZ) method of crystal growth in the presence of a magnetic field. (author) 24 figs., 10 tabs., 136 refs

  16. Optimisation of safety parameters in fast breeder test reactor

    International Nuclear Information System (INIS)

    Full text: Optimisation of safety parameters is an important aspect to be considered in the design of nuclear power plant and also becomes extremely important activity to be followed up during the commissioning and operating phases of the plant taking into account the operational feed back and review of incidental situations and available diversity and reliability. Otherwise, the spurious/ superfluous trips on the reactor besides affecting the availability of the plant, initiate plant transients causing stress for the plant equipment resulting in reduction of plant life. This activity has a significant role to play in attaining the maximum availability of the plant, without compromising safety. The study and evolution of optimisation process in fast breeder test reactor (FBTR); at Kalpakkam has been an interesting and rewarding experience

  17. Design study of an upgraded charge breeder for ISOLDE

    CERN Document Server

    Shornikov, A; Wenander, F; Pikin, A

    2013-01-01

    In this work we present our progress in the design study of a new Electron Beam Ion Source (EBIS) to be installed as a charge breeder for reacceleration of rare ions at ISOLDE. The work is triggered by the HIE-ISOLDE upgrade {[}1] and the planned TSR@ISOLDE project {[}2]. To fulfill the requests of the user community the new EBIS should reach an electron beam density of 10(4) A/cm(2) at electron energies up to 150 key and, provide UHV environment and ion cooling in the breeding region to ensure confinement of the ions long enough to reach the requested charge states. We report on the established design parameters and first prototyping steps towards production and testing of suitable equipment. (C) 2013 Elsevier B.V. All rights reserved.

  18. Designing a SCADA system simulator for fast breeder reactor

    Science.gov (United States)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

    2016-04-01

    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  19. Feasibility and deployment strategy of water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    The author have studied water cooled thorium breeder reactor based on matured pressurized water reactor (PWR) plant technology for several years. Through these studies it is concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array with heavy water coolant in the primary loop by replacing original light water is appropriate for achieving sufficient breeding performance as sustainable fission system and high enough burn-up as an economical power plant. The heavy water cooled thorium reactor is feasible to be introduced by using Pu recovered from spent fuel of LWR, keeping continuity with current LWR infrastructure. This thorium reactor can be operated as sustainable energy supplier and also MA transmuter to realize future society with less long-lived nuclear waste

  20. Computational intelligent systems for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Nearly 15000 process signals are digitized by physically and functionally distributed embedded systems in Prototype Fast Breeder Reactor (PFBR). Digitized signals are processed and relevant information is displayed through Large video display systems at Control Room. It is necessary that correct and reliable information need to be provided to the plant operator. Computational intelligent systems play a major role in enhancing the safe operation of the Nuclear reactor. The paper explains the features of three such systems, one for on-line validation of neutronic power channel through on-line thermal balance calculation and another for detection of anomalous reactivity addition through on-line reactivity balance computation and third for on-line computation of Reactor power from fluctuations of core thermocouple signals. (author)

  1. Challenges for Plant Breeders from the View of Animal Nutrition

    Directory of Open Access Journals (Sweden)

    Gerhard Flachowsky

    2015-12-01

    Full Text Available The question of how to feed the growing world population is very old, but because of the increase of population and possible climate change, currently it has an explosive impact. Plant breeding can be considered as the starting point for the whole human food chain. Therefore, high, stable and highly digestible yields of phytogenic biomass with low external inputs of non-renewable resources, such as water, fuel, arable land, fertilizers, etc.; low emissions of gases with greenhouse potential during cultivation; and high resistance against biotic and abiotic stressors, including adaptation to potential climate change, and a low concentration of undesirable substances in the plants are real challenges for plant breeders in the future. Virtually unlimited resources such as sunlight, nitrogen and carbon dioxide from the air as well as the genetic pool of microbes, plants and animals can be used to breed/develop optimal plants/crops. Biofortification of plants may also be an objective of plants breeders, but it is more important for human nutrition to avoid micronutrient deficiencies. A lower concentration of undesirable substances in the plants can be considered as more important than higher concentrations of micronutrients in plants/feeds. Animal nutritionists have various possibilities for feed additive supplementation to meet animal nutrient requirements. Examples to reduce undesirable substances in feed plants are discussed and shown in the paper. In summary, plant breeding has a large and strategic potential for global feed and food security. All breeding technologies may contribute to solving important global challenges, such as sustainable use of limited global resources, improved use of unlimited resources, adaption to climate change and lowering global greenhouse gas emission. More publically supported research seems to be necessary in this field. All methods of plant breeding that contribute to a more resource-efficient production of high

  2. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  3. Ceramic Technology Project

    Energy Technology Data Exchange (ETDEWEB)

    1992-03-01

    The Ceramic Technology Project was developed by the USDOE Office of Transportation Systems (OTS) in Conservation and Renewable Energy. This project, part of the OTS's Materials Development Program, was developed to meet the ceramic technology requirements of the OTS's automotive technology programs. Significant accomplishments in fabricating ceramic components for the USDOE and NASA advanced heat engine programs have provided evidence that the operation of ceramic parts in high-temperature engine environments is feasible. These programs have also demonstrated that additional research is needed in materials and processing development, design methodology, and data base and life prediction before industry will have a sufficient technology base from which to produce reliable cost-effective ceramic engine components commercially. A five-year project plan was developed with extensive input from private industry. In July 1990 the original plan was updated through the estimated completion of development in 1993. The objective is to develop the industrial technology base required for reliable ceramics for application in advanced automotive heat engines. The project approach includes determining the mechanisms controlling reliability, improving processes for fabricating existing ceramics, developing new materials with increased reliability, and testing these materials in simulated engine environments to confirm reliability. Although this is a generic materials project, the focus is on the structural ceramics for advanced gas turbine and diesel engines, ceramic bearings and attachments, and ceramic coatings for thermal barrier and wear applications in these engines. To facilitate the rapid transfer of this technology to US industry, the major portion of the work is being done in the ceramic industry, with technological support from government laboratories, other industrial laboratories, and universities.

  4. Status of fast breeder reactor development in the United States of America - April 1984

    International Nuclear Information System (INIS)

    The Breeder Technology program continues to produce viable information on fuel performance, nuclear systems technology, and power conversion technology. The unique testing capabilities design into the FFTF have resulted in well-validated materials and fuels irradiation information that has confirmed and extended previous data bases. Current directions for the research and development program are to improve the technology for power conversion systems, components, instrumentation, and materials technology to the point where cost reduction and reliability potentials are realized. Operation of the breeder test facility complex at the Hanford Engineering Development Laboratory (HEDL), the Energy Technology Engineering Center (ETEC), and the Argonne National Laboratory (ANL) continues to provide the experience base and test capability for the breeder R and D effort. International cooperation will be even more important in the future than in the past for several reasons. Significant new investments still have to be made in breeder R and D to improve designs, achieve economic competitiveness and to develop practical breeder fuel cycle capabilities. Progress can be accelerated, redundancies avoided, and economics achieved if nations coordinate their programs, and where possible, divide up the work. In addition, there is clear mutual benefit in encouraging the countries involved in breeder development to harmonize standards and regulations related to safety. It is also important that the advanced nations work together closely in assuring that adequate international safeguards, export controls, and national physical security measures keep pace with breeder reactor and fuel cycle developments

  5. Verification of ceramic structures

    NARCIS (Netherlands)

    Behar-Lafenetre, S.; Cornillon, L.; Rancurel, M.; Graaf, D. de; Hartmann, P.; Coe, G.; Laine, B.

    2012-01-01

    In the framework of the "Mechanical Design and Verification Methodologies for Ceramic Structures" contract [1] awarded by ESA, Thales Alenia Space has investigated literature and practices in affiliated industries to propose a methodological guideline for verification of ceramic spacecraft and instr

  6. Beryllium usage in fusion blankets and beryllium data needs

    International Nuclear Information System (INIS)

    Increasing numbers of designers are choosing beryllium for fusion reactor blankets because it, among all nonfissile materials, produces the highest number (2.5 neutron in an infinite media) of neutrons per 14-MeV incident neutron. In amounts of about 20 cm of equivalent solid density, it can be used to produce fissile material, to breed all the tritium consumed in ITER from outboard blankets only, and in designs to produce Co-60. The problem is that predictions of neutron multiplication in beryllium are off by some 10 to 20% and appear to be on the high side, which means that better multiplication measurements and numerical methods are needed. The n,2n reactions result in two helium atoms, which cause radiation damage in the form of hardening at low temperatures (300/degree/C). The usual way beryllium parts are made is by hot pressing the powder. A lower cost method is to cold press and then sinter. There is no radiation damage data on this form of beryllium. The issues of corrosion, safety relative to the release of the tritium built-up inside beryllium, and recycle of used beryllium are also discussed. 10 figs

  7. Fission power flattening in hybrid blankets using mixed fuel

    International Nuclear Information System (INIS)

    In a source-driven fissionable blanket, a flat fission power density (FPD) is achieved by using a mixed fuel (ThO2 and natural UO2) with the thoriumuranium ratio changing from front to back in the ten fuel rows along the radial direction. A straightforward graphic method is used. The temporal behavior of the FPD has been observed for an operation period of 6 months and for a plant load factor of 75% by applying a fusion driver neutron flux of 1014 14-MeV neutrons(cm2 . s) at the first wall, corresponding to --2.25 MWm2. To keep the power density flat, it is necessary to replace the fuel in rows 1, 2, and 3, close to the first wall. The time intervals for this operation increase, counting from initial start-up, typically, 2 months, 6 months, etc. One result of this study is that plutonium produced in such a hybrid blanket contains very low amounts of even isotopic components even over very long operation times of --3 yr. Hence, if fusion reactors are introduced into the energy market, special regulations are needed for international safeguarding

  8. Damage Identification of Continuum Structures Using Blanketing Effect

    International Nuclear Information System (INIS)

    A damage identification method for the continuum structures with making use of the blanketing effect is proposed in this paper. The conventional damage indicator methods such as flexibility matrix method are difficult to deal with the damage identification of complex continuum structure such as multi-span beam. The presented method takes the change rates according to certain diagonal elements of the damaged structure flexibility matrix as identification indicator function. The 'blanketing effect' of the indicator function with respect to the damage factors makes it easy to identify the damage of multi-span beam and plate structures. Some useful formulas are derived and the sensitivity matrix of the identification indicator function is studied. The numerical simulation is performed and the identification results of several damage status are compared. Both beam and plate structures are used for numerical simulation. The results show that the presented indicator function method is very brevity and effective. It is especially suitable for the damage identification of the multi-span beam and plate structures such as bridges and floor

  9. Elevator mode convection in liquid metal blankets for fusion reactors

    Science.gov (United States)

    Zikanov, Oleg; Liu, Li

    2015-11-01

    The work is motivated by the design of liquid-metal blankets for nuclear fusion reactors. Mixed convection in a downward flow in a vertical duct with strong contant-rate heating of one wall (the Grashof number up to 1012) and strong transverse magnetic field (the Hartmann number up to 104) is considered. It is found that in an infinitely long duct the flow is dominated by exponentially growing elevator modes having the form of a combination of ascending and descending jets. An analytical solution approximating the growth rate of the modes is derived. Analogous flows in finite-length pipes and ducts are analyzed using the high-resolution numerical simulations. The results of the recent experiments are reproduced and explained. It is found that the flow evolves in cycles consisting of periods of exponential growth and breakdowns of the jets. The resulting high-amplitude fluctuations of temperature is a feature potentially dangerous for operation of a reactor blanket. Financial support was provided by the US NSF (Grant CBET 1232851).

  10. Annual report of the CTR Blanket Engineering research facility in 1994

    International Nuclear Information System (INIS)

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor(CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1994. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (author)

  11. Conceptual design of an electricity generating tritium breeding blanket sector for INTOR/NET

    International Nuclear Information System (INIS)

    A study is made of a fusion reactor power blanket and its associated equipment with the objective of producing a conceptual design for a blanket sector of INTOR, or one of its national variants (e.g. NET), from which electricity could be generated simultaneously with the breeding of tritium. (author)

  12. 77 FR 76013 - Sempra LNG Marketing, LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-12-26

    ... LNG Marketing, LLC; Application for Blanket Authorization To Export Previously Imported Liquefied.... 2885, which granted Sempra LNG Marketing authority to export a cumulative total of ] 250 Bcf of... Marketing requests blanket authorization to export LNG from the Cameron Terminal that has been...

  13. 75 FR 19954 - Cheniere Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-04-16

    ... Cheniere Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY... Cheniere Marketing, LLC (CMI), requesting blanket authorization to export liquefied natural gas (LNG) that... exported from the Sabine Pass LNG terminal owned by CMI's affiliate, Sabine Pass LNG, L.P., in...

  14. 78 FR 4400 - Eni USA Gas Marketing LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2013-01-22

    ... USA Gas Marketing LLC; Application for Blanket Authorization To Export Previously Imported Liquefied... and order (Order No. 2923) that granted Eni USA Gas Marketing authority to export a cumulative total... Application, Eni USA Gas Marketing requests blanket authorization to export LNG from the Cameron Terminal...

  15. Synthesis and fabrication of lithium-titanate pebbles for ITER breeding blanket by solid state reaction and spherodization

    Energy Technology Data Exchange (ETDEWEB)

    Mandal, D., E-mail: dmandal@barc.gov.i [Chemical Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Shenoi, M.R.K.; Ghosh, S.K. [Chemical Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2010-08-15

    {sup 6}Li produces tritium by (n, {alpha}) nuclear reaction, {sup 6}Li + {sup 1}n {yields} {sup 4}He + {sup 3}H. Lithium titanate (Li{sub 2}TiO{sub 3}) enriched with {sup 6}Li, is the most promising candidate for solid test blanket module (TBM) material for fusion reactors. Various processes are reported in literature for the fabrication of Li{sub 2}TiO{sub 3} pebbles for its use as TBM material. A process has been developed based on the solid state reaction of lithium-carbonate and titanium-dioxide for the synthesis of lithium titanate and pebble fabrication by extrusion, spherodization and sintering. This paper discusses the sequence of steps followed in this process and the properties obtained. Analytical grade titanium-dioxide and lithium-carbonate were taken in stoichiometric ratio and were milled to ensure thorough intimate mixing and obtain fine particles less than 45 {mu}m before its calcination at 900 {sup o}C. Following calcination, the agglomerated product was again milled to fine particles of size less than 45 {mu}m. Aqueous solution of ploy-vinyl-alcohol was added as binder prior to its feeding in the extruder. The extruded strips were spherodized and spherical pebbles were dried and sintered at 900 {sup o}C for different duration. Pebbles of desired density and porosity were obtained by suitable combination of sintering temperature and duration of sintering. Properties of the prepared pebbles were also characterized for sphericity, pore size distribution, grain size, crushing load strength, etc. The values were found to be conforming to the desired properties for use as solid breeder. The attractive feature of this process is almost no waste generation.

  16. Wash resistance and repellent properties of Africa University mosquito blankets against mosquitoes

    Directory of Open Access Journals (Sweden)

    N. Lukwa

    2013-04-01

    Full Text Available The effect of permethrin-treated Africa University (AU mosquito blankets on susceptible female Anopheles gambiae sensu lato mosquitoes was studied under laboratory conditions at Africa University Campus in Mutare, Zimbabwe. Wash resistance (ability to retain an effective dose that kills ≥80% of mosquitoes after a number of washes and repellence (ability to prevent ≥80% of mosquito bites properties were studied. The AU blankets were wash resistant when 100% mortality was recorded up to 20 washes, declining to 90% after 25 washes. Untreated AU blankets did not cause any mortality on mosquitoes. However, mosquito repellence was 96%, 94%, 97.9%, 87%, 85% and 80.7% for treated AU blankets washed 0, 5, 10, 15, 20 and 25 times, respectively. Mosquito repellence was consistently above 80% from 0-25 washes. In conclusion, AU blankets washed 25 times were effective in repelling and killing An. gambiae sl mosquitoes under laboratory conditions.

  17. Resonance self-shielding in the blanket of a hybrid reactor

    International Nuclear Information System (INIS)

    Three sets of energy group cross sections were obtained using various approximations for resonance self shielding. The three models used in obtaining the cross sections were: (a) infinitely dilute model, (b) homogeneous-medium resonance self shielding, and (c) heterogeneous-medium resonance self shielding. The effects on the blanket performance of fusion--fission hybrid reactors, and in particular, on the performance of the current reference Westinghouse Demonstration Tokamak Hybrid Reactor blanket, were compared and analyzed for a variety of fuel-coolant combinations. It has been concluded that (1) the infinitely dilute cross sections can be used to produce preliminary crude estimates for beginning-of-life (BOL) only, (2) the resonance absorber finite dilution should be considered for BOL, poorly moderated blankets and well moderated blankets with low fissile material content situations, and (3) the spacial details should be considered in high fissile content, well moderated blanket situations

  18. Blanket design and performance for the LOTUS fusion-fission hybrid test facility

    International Nuclear Information System (INIS)

    This report summarizes the results of studies performed during 1982 to design an optimized blanket for the initial series of experiments to be conducted in the LOTUS test facility at the Swiss Federal Institute of Technology in Lausanne (EPFL). The experiments are expected to begin in early 1984. An Overview of different hybrid blanket design concepts proposed to date is first given. The technological and economic implications of the different blanket design philosophies are discussed to provide the basis and rationale for the thorium fast-fission blanket design concept selected for the first series of experiments. Detailed description, dimensions, and characteristics of the selected blanket design are given. The neutronic optimization studies on which the design is based are described in detail. Instrumentation and measurement techniques to be used in LOTUS are described elsewhere

  19. Ceramics As Materials Of Construction

    OpenAIRE

    Zaki, A.; Eteiba, M. B.; Abdelmonem, N.M.

    1988-01-01

    This paper attempts to review the limitations for using the important ceramics in contact with corrosive media. Different types of ceramics are included. Corrosion properties of ceramics and their electrical properties are mentioned. Recommendations are suggested for using ceramics in different media.

  20. Transient characteristic analyses of ex-vessel coolant pipe break for Chinese helium-cooled solid breeder TBM based on RELAP5 code

    International Nuclear Information System (INIS)

    Chinese helium-cooled solid breeder (CH HCSB) test blanket module (TBM) with helium cooling system and secondary cooling water system was modeled and thermal-hydraulic behavior and safety performance of the system were assessed using the RELAP5/MOD3.4 code. According to the accident sequences of ITER accident analysis specification for TBM, the transient analysis of the design basis ex-vessel coolant pipe break accident was carried out. The influences of different break locations, leak areas and plasma shutdown processes on the first wall of TBM were compared. The results indicate that it is much more danger when the pipe break occurs at the downstream side of the helium circulator compared with that at upstream side. The results also show that the accident consequence is worse in case of smaller area break than that in case of larger area break. In case of much more severe accident that the ex-vessel break leads to the break of TBM the first wall, the results reveal that the decay heat can be removed to cool down TBM by natural circulation and radiation. The first wall melting can be avoided if the method to shutdown plasma within 3 seconds in case of ex-vessel break is adopted. (authors)

  1. High pressure ceramic joint

    Science.gov (United States)

    Ward, Michael E.; Harkins, Bruce D.

    1993-01-01

    Many recuperators have components which react to corrosive gases and are used in applications where the donor fluid includes highly corrosive gases. These recuperators have suffered reduced life, increased service or maintenance, and resulted in increased cost. The present joint when used with recuperators increases the use of ceramic components which do not react to highly corrosive gases. Thus, the present joint used with the present recuperator increases the life, reduces the service and maintenance, and reduces the increased cost associated with corrosive action of components used to manufacture recuperators. The present joint is comprised of a first ceramic member, a second ceramic member, a mechanical locking device having a groove defined in one of the first ceramic member and the second ceramic member. The joint and the mechanical locking device is further comprised of a refractory material disposed in the groove and contacting the first ceramic member and the second ceramic member. The present joint mechanically provides a high strength load bearing joint having good thermal cycling characteristics, good resistance to a corrosive environment and good steady state strength at elevated temperatures.

  2. Shutdown and Closure of the Experimental Breeder Reactor - II

    International Nuclear Information System (INIS)

    The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor - II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated lay-up plan defining the system end state, as well as instructions for achieving the lay-up condition. A goal of system-by-system lay-up is to minimize surveillance and

  3. Shutdown and closure of the experimental breeder reactor - II

    International Nuclear Information System (INIS)

    The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and

  4. Preliminary thermo-mechanical analysis of ITER breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Shigeto; Kuroda, Toshimasa; Enoeda, Mikio [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1999-01-01

    Thermo-mechanical analysis has been conducted on ITER breeding blanket taking into account thermo-mechanical characteristics peculiar to pebble beds. The features of the analysis are to adopt an elasto-plastic constitutive model for pebble beds and to take into account spatially varying thermal conductivity and heat transfer coefficient, especially in the Be pebble bed, depending on the stress. ABAQUS code and COUPLED TEMPERATURE-DISPLACEMENT procedure of the code are selected so that thermal conductivity is automatically calculated in each calculation point depending on the stress. The modified DRUCKER-PRAGER/Cap plasticity model for granular materials of the code is selected so as to deal with such mechanical features of pebble bed as shear failure flow and hydrostatic plastic compression, and capability of the model is studied. The thermal property-stress correlation used in the analysis is obtained based on the experimental results at FZK and the results of additional thermo-mechanical analysis performed here. The thermo-mechanical analysis of an ITER breeding blanket module has been performed for four conditions: case A; nominal case with spatial distribution of thermal conductivity and heat transfer coefficient in Be pebble bed depending on the stress, case B; constant thermal conductivity, case C; thermal conductivity = -20% of nominal case, and case D; thermal conductivity = +20% of nominal case. In the nominal case the temperature of breeding material (Li{sub 2}ZrO{sub 3}) ranges from 317degC to 554degC and the maximum temperature of Be pebble bed is 446degC. It is concluded that the temperature distribution is within the current design limits. Though the analyses performed here are preliminary, the results exhibit well the qualitative features of the pebble bed mechanical behaviors observed in experiments. For more detail quantitative estimates of the blanket performance, further investigation on mechanical properties of pebble beds by experiment

  5. An Evaluation of liquid metal leak detection methods for the Clinch River Breeder Reactor Plant

    Energy Technology Data Exchange (ETDEWEB)

    Morris, C.J.; Doctor, S.R.

    1977-12-01

    This report documents an independent review and evaluation of sodium leak detection methods described in the Clinch River Breeder Reactor Preliminary Safety Analysis Report. Only information in publicly available documents was used in making the assessments.

  6. The APS ceramic chambers

    Energy Technology Data Exchange (ETDEWEB)

    Milton, S.; Warner, D.

    1994-07-01

    Ceramics chambers are used in the Advanced Photon Source (APS) machines at the locations of the pulsed kicker and bumper magnets. The ceramic will be coated internally with a resistive paste. The resistance is chosen to allow the low frequency pulsed magnet field to penetrate but not the high frequency components of the circulating beam. Another design goal was to keep the power density experienced by the resistive coating to a minimum. These ceramics, their associated hardware, the coating process, and our recent experiences with them are described.

  7. Influence of age and strain on reproductive performance of the broiler breeder female

    OpenAIRE

    Alzenbarakji, Nada

    2011-01-01

    Chicken meat is an important source of high quality protein in the diet of most people in the world. Consequently, the increasing demand for this meat has made chicken meat production the most important growth sector among other meat species. This has been achieved by a half century of intensive genetic selection for growth traits; however, this was coupled with poor reproductive performance of broiler breeders. Ross 708 represents a broiler breeder strain that has been developed for breas...

  8. Poor welfare or future investment? Different growth pattern of broiler breeders

    OpenAIRE

    Calais, Andreas

    2015-01-01

    The parental stock of meat type chickens (broiler breeders) are commonly feed restricted to decrease their rapid growth and the issues associated with it. Among these birds, chronic hunger and stress are the most prominent welfare concerns and mass heterogeneity within flocks a major management challenge. The present study compared small and large broiler breeders of the same age within a flock, with the hypothesis that small birds would show signs of poorer welfare indicated by higher cortic...

  9. Detection and characterization of chicken anemia virus from commercial broiler breeder chickens

    OpenAIRE

    Omar Abdul; Hailemariam Zerihun; Hair-Bejo Mohd; Giap Tan

    2008-01-01

    Abstract Background Chicken anemia virus (CAV) is the causative agent of chicken infectious anemia (CIA). Study on the type of CAV isolates present and their genetic diversity, transmission to their progeny and level of protection afforded in the breeder farms is lacking in Malaysia. Hence, the present study was aimed to detect CAV from commercial broiler breeder farms and characterize CAV positive samples based on sequence and phylogenetic analysis of partial VP1 gene. Results A total of 12 ...

  10. Safety and core design of large liquid-metal cooled fast breeder reactors

    OpenAIRE

    Qvist, Staffan Alexander

    2013-01-01

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cyc...

  11. Neutral gas blanket effects in a gaseous divertor

    International Nuclear Information System (INIS)

    The gaseous divertor employs a neutral gas blanket to absorb the plasma heat flux in the divertor chamber. This novel method for resolving the heat loading problem in a conventional divertor system is simulated experimentally. In our operational range (nsub(e) 13 cm-3, Tsub(e) <= 5 eV) it is demonstrated that the localized plasma heat flux is scattered relatively uniformly with neutral pressures of a few microns. At large neutral pressures the plasma stream is neutralized without touching a material wall. Plasma pumping inhibits neutral backflow and can sustain a neutral pressure difference comparable to the plasma pressure. Effective divertor channel conductance is measured to be reduced by a factor of six. (orig.)

  12. Heating performances of a IC in-blanket ring array

    Science.gov (United States)

    Bosia, G.; Ragona, R.

    2015-12-01

    An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is due to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing, at the highest power level, high RF standing and DC rectified voltages at the plasma periphery, with frequent voltage breakdowns and enhanced local wall loading. In a previous work [1] the concept for an Ion Cyclotron Heating & Current Drive array (and using a different wave guide technology, a Lower Hybrid array) based on the use of periodic ring structure, integrated in the reactor blanket first wall and operating at high input power and low power density, was introduced. Based on the above concept, the heating performance of such array operating on a commercial fusion reactor is estimated.

  13. Heating performances of a IC in-blanket ring array

    Energy Technology Data Exchange (ETDEWEB)

    Bosia, G., E-mail: gbosia@to.infn.it [Department of Physics, University of Turin (Italy); Ragona, R. [Laboratory for Plasma Physics-LPP-ERM/KMS, Brussels (Belgium)

    2015-12-10

    An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is due to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing, at the highest power level, high RF standing and DC rectified voltages at the plasma periphery, with frequent voltage breakdowns and enhanced local wall loading. In a previous work [1] the concept for an Ion Cyclotron Heating & Current Drive array (and using a different wave guide technology, a Lower Hybrid array) based on the use of periodic ring structure, integrated in the reactor blanket first wall and operating at high input power and low power density, was introduced. Based on the above concept, the heating performance of such array operating on a commercial fusion reactor is estimated.

  14. Cosmetic wastewater treatment by upflow anaerobic sludge blanket reactor

    International Nuclear Information System (INIS)

    Anaerobic treatment of pre-settled cosmetic wastewater in batch and continuous experiments has been investigated. Biodegradability tests showed high COD and solid removal efficiencies (about 70%), being the hydrolysis of solids the limiting step of the process. Continuous treatment was carried out in an upflow anaerobic sludge blanket reactor. High COD and TSS removal efficiencies (up to 95% and 85%, respectively) were achieved over a wide range of organic load rate (from 1.8 to 9.2 g TCOD L-1 day-1). Methanogenesis inhibition was observed in batch assays, which can be predicted by means of a Haldane-based inhibition model. Both COD and solid removal were modelled by Monod and pseudo-first order models, respectively.

  15. Cosmetic wastewater treatment by upflow anaerobic sludge blanket reactor

    Energy Technology Data Exchange (ETDEWEB)

    Puyol, D.; Monsalvo, V.M.; Mohedano, A.F. [Seccion de Ingenieria Quimica, Facultad de Ciencias, Universidad Autonoma de Madrid, C/ Francisco Tomas y Valiente 7, 28049, Madrid (Spain); Sanz, J.L. [Departamento de Biologia Molecular, Facultad de Ciencias, Universidad Autonoma de Madrid, C/ Francisco Tomas y Valiente 7, 28049, Madrid (Spain); Rodriguez, J.J., E-mail: juanjo.rodriguez@uam.es [Seccion de Ingenieria Quimica, Facultad de Ciencias, Universidad Autonoma de Madrid, C/ Francisco Tomas y Valiente 7, 28049, Madrid (Spain)

    2011-01-30

    Anaerobic treatment of pre-settled cosmetic wastewater in batch and continuous experiments has been investigated. Biodegradability tests showed high COD and solid removal efficiencies (about 70%), being the hydrolysis of solids the limiting step of the process. Continuous treatment was carried out in an upflow anaerobic sludge blanket reactor. High COD and TSS removal efficiencies (up to 95% and 85%, respectively) were achieved over a wide range of organic load rate (from 1.8 to 9.2 g TCOD L{sup -1} day{sup -1}). Methanogenesis inhibition was observed in batch assays, which can be predicted by means of a Haldane-based inhibition model. Both COD and solid removal were modelled by Monod and pseudo-first order models, respectively.

  16. Flow induced vibrations in liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Flow induced vibrations are well known phenomena in industry. Engineers have to estimate their destructive effects on structures. In the nuclear industry, flow induced vibrations are assessed early in the design process, and the results are incorporated in the design procedures. In many cases, model testing is used to supplement the design process to ensure that detrimental behaviour due to flow induced vibrations will not occur in the component in question. While these procedures attempt to minimize the probability of adverse performance of the various components, there is a problem in the extrapolation of analytical design techniques and/or model testing to actual plant operation. Therefore, sodium tests or vibrational measurements of components in the reactor system are used to provide additional assurance. This report is a general survey of experimental and calculational methods in this area of structural mechanics. The report is addressed to specialists and institutions in industrialized and developing countries who are responsible for the design and operation of liquid metal fast breeder reactors. 92 refs, 90 figs, 8 tabs

  17. Performance of female broiler breeders submitted to different feeding schedules

    Directory of Open Access Journals (Sweden)

    VS de Avila

    2003-12-01

    Full Text Available The performance of Arbor Acres broiler breeders (1,296 females; 144 roosters was evaluated when submitted to the following treatments (T: T1 = feeding at 6:30 a.m. (control; T2 = 50% feeding at 6:30 a.m. and 50% at 3:30 p.m. (dual feeding; T3 = feeding at 11:00 a.m.; and T4 = feeding at 3:30 p.m. Treatments were randomly distributed in 48 pens. There were 27 females and 3 males in each pen and 12 repetitions per treatment. Nutrition and management were as recommended for the commercial strain. It was evaluated age at first egg (AFE, total egg production (TEP, number of days with production above 80% (DAP80, laying peak (P, female mortality (MOR, and gross profit margin (GM per hen. Data were submitted to analysis of variance and means were compared by Student's t-Test. TEP of T1 (186.3±2.3 and T2 (186.5±1.5 were higher (p0.10 among treatments. GM per hen was better (p<0.05 in T1 and T2 hens. Control and dual treatments were more efficient than other treatments. It was concluded that it is possible to change conventional feeding management's by the dual feeding system.

  18. Fast breeder reactor-block antiseismic design and verification

    International Nuclear Information System (INIS)

    The Specialists' Meeting on ''Fast Breeder Reactor-Block Antiseismic Design and Verification'' was organized by the ENEA Fast Reactor Department in co-operation with the International Working Group (IWGFR) of the International Atomic Energy Agency (IAEA), according to the recommendations of the 19th IAEA/IWGFR Meeting. It was held in Bologna, at the Headquarters of the ENEA Fast Reactor Department, on October 12-15, 1987, in the framework of the Celebrations for the Ninth Centenary of the Bologna University. The proceedings of the meeting consists of three parts. Part 1 contains the introduction and general comments, the agenda of the meeting, session summaries, conclusions and recommendations and the list of participants. Part 2 contains 8 status reports of Member States participating in the Working Group. Contributed papers were published in Part 3 and were further subdivided into 5 sessions as follows: whole reactor-block analysis (4 papers); whole reactor-block analysis (sloshing and buckling, seismic isolation effects) (8 papers); detailed core analysis (6 papers); shutdown systems and core structural and functional verifications (6 papers); component and piping analysis (7 papers). A separate abstract was prepared for each of the 8 status reports and 31 contributed papers. Refs, figs and tabs

  19. Capillariosis in breeder discus (Symphysodon aequifasciatus in Iran

    Directory of Open Access Journals (Sweden)

    Rahmati-Holasoo Hooman

    2010-01-01

    Full Text Available The global ornamental fish trade is a rapidly growing industry. Cultivation and propagation of ornamental fishes have been increasing in the last 20 years in Iran. Discus (Symphysodon aequifasciatus from Cichlidae is one of the most popular and expensive aquarium fish. In the past few years farming of this fish has been well developed in Iran. Two breeder discus fish (Symphysodon aequifasciatus from two different propagation centres (with high mortality with signs of anorexia, loss of balance, moribundity and darkness in skin colour were referred to Laboratory of Aquatic Diseases of Veterinary Faculty, University of Tehran. After the survey of ectoparasites, necropsy was performed under aseptic conditions; bacterial culture on standard media was done and the alimentary canal was extruded. In both fish no ectoparasite was detected and no bacteria from these cases grew on the standard media. In internal survey 5 and 25 nematodes were detected in each fish. A high number of free eggs were observed in intestine of fish. Regarding morphological characteristics of the nematodes and their eggs, they were identified as Capillaria sp. Treatment of other fish with levamisole was effective and the loss was terminated. Some helminthes like Capillaria pterophylli Heinze, 1933, can cause a high mortality in cichlid aquarium fishes. This study showed that infection with some species of Capillaria could cause a heavy loss in ornamental fish from Cichlidae. Diagnosis of parasites of these fishes can help us to prevent high mortalities.

  20. Molten Salt Breeder Reactor Analysis Based on Unit Cell Model

    International Nuclear Information System (INIS)

    Contemporary computer codes like the MCNP6 or SCALE are only good for solving a fixed solid fuel reactor. However, due to the molten-salt fuel, MSR analysis needs some functions such as online reprocessing and refueling, and circulating fuel. J. J. Power of Oak Ridge National Laboratory (ORNL) suggested in 2013 a method for simulating the Molten Salt Breeder Reactor (MSBR) with SCALE, which does not support continuous material processing. In order to simulate MSR characteristics, the method proposes dividing a depletion time into short time intervals and batchwise reprocessing and refueling at each step. We are applying this method by using the MCNP6 and PYTHON and NEWT-TRITON-PYTHON and PYTHON code systems to MSBR. This paper contains various parameters to analyze the MSBR unit cell model such as the multiplication factor, breeding ratio, change of amount of fuel, amount of fuel feeding, and neutron flux distribution. The result of MCNP6 and NEWT module in SCALE show some difference in depletion analysis, but it still seems that they can be used to analyze MSBR. Using these two computer code system, it is possible to analyze various parameters for the MSBR unit cells such as the multiplication factor, breeding ratio, amount of material, total feeding, and neutron flux distribution. Furthermore, the two code systems will be able to be used for analyzing other MSR model or whole core models of MSR

  1. Feeding broiler breeders to improve their welfare whilst maintaining productivity

    DEFF Research Database (Denmark)

    Steenfeldt, Sanna; Nielsen, Birte Lindstrøm

    ; and SOF, high soluble fibre content) were each fed to 10 groups of 12 broiler breeder chickens (age: 2 to 15 weeks). Similar growth rates were obtained on different quantities of food (e.g. food allocation in week 14: approx. 80, 100, and 130 g/d for CON, INF, and SOF, respectively) with all birds...... reaching commercial target weight at 15 weeks of age. Birds fed CON ate significantly more in a hunger test than birds on diets INF and SOF, indicating that these two high-fibre diets did reduce the level of hunger experienced by the birds. Behavioural observations carried out at 14 weeks of age showed...... high levels of tail pecking in birds fed CON and almost none in birds fed SOF, whereas birds fed INF were intermediate. Stereotypic pecking was most frequently seen in birds fed CON and never observed in birds fed INF. Birds on diet SOF appeared scruffier in their plumage, and the higher water content...

  2. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    P K Vijayan; A Basak; I V Dulera; K K Vaze; S Basu; R K Sinha

    2015-09-01

    The third stage of Indian nuclear power programme envisages the use of thorium as the fertile material with 233U, which would be obtained from the operation of Pu/Th-based fast reactors in the later part of the second stage. Thorium-based reactors have been designed in many configurations, from light water-cooled designs to high-temperature liquid metal-cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian molten salt breeder reactor (IMSBR). Presently, various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel, fundamental studies on natural circulation and corrosion behaviour of various molten salts have also been initiated.

  3. Innovations in Equipment Erection of Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) is sodium cooled, pool type reactor with generating capacity of 1250 MWt/500 MWe. Reactor assembly consists of large dimensional vessels like Safety vessel (13.54 m diameter, 12.8 m height and weight approximately 155 MT) and Main vessel (12.9 m diameter, 12.94 m height and weight approximately 202 MT including core catcher, core support structure and cooling pipes) and Steam generator (26 m length, 1.5 m diameter, and weight approximately 35 MT). PFBR reactor equipment erection was a challenging task where thin walled vessels had transported and handled with utmost precaution to avoid radial forces on the vessels which could buckle the vessels. There was a real challenge in lifting the vessels without swing, placement of large size and heavy vessel at a distance of 57 m where the crane operator had no line of site to the equipment being erected. To handle such over dimensional reactor components many mock-up tests had been carried out before erection and gained lot of confidence. Lot of care had been taken during lifting, handling and erection of thin walled over dimensional reactor components with innovative methods used for lifting fixtures, guiding arrangements, alignment fixtures and achieved the stringent erection tolerances. This paper discusses the first ever experiences gained during the handling and erection of such thin walled, over dimensional reactor components at PFBR site. (author)

  4. Method of advancing research and development of fast breeder reactors

    International Nuclear Information System (INIS)

    In the long term plan of atomic energy development and utilization, fast breeder reactors are to be developed as the main of the future nuclear power generation in Japan, and when their development is advanced, it has been decided to positively aim at building up the plutonium utilization system using FBRs superior to the uranium utilization system using LWRs. Also it has been decided that the development of FBRs requires to exert incessant efforts for a considerable long period under the proper cooperation system of government and people, and as for its concrete development, hereafter the deliberation is to be carried out in succession by the expert subcommittee on FBR development projects of the Atomic Energy Commission. The subcommittee was founded in May, 1986, to deliberate on the long term promotion measures for FBR development, the measures for promoting the research and development, the examination of the basic specification of a demonstration FBR, the measures for promoting international cooperation, and other important matters. As the results of investigation, the situation around the development of FBRs, the fundamentals at the time of promoting the research and development, the subjects of the research and development and so on are reported. (Kako, I.)

  5. Molten Salt Breeder Reactor Analysis Based on Unit Cell Model

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yongjin; Choi, Sooyoung; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    Contemporary computer codes like the MCNP6 or SCALE are only good for solving a fixed solid fuel reactor. However, due to the molten-salt fuel, MSR analysis needs some functions such as online reprocessing and refueling, and circulating fuel. J. J. Power of Oak Ridge National Laboratory (ORNL) suggested in 2013 a method for simulating the Molten Salt Breeder Reactor (MSBR) with SCALE, which does not support continuous material processing. In order to simulate MSR characteristics, the method proposes dividing a depletion time into short time intervals and batchwise reprocessing and refueling at each step. We are applying this method by using the MCNP6 and PYTHON and NEWT-TRITON-PYTHON and PYTHON code systems to MSBR. This paper contains various parameters to analyze the MSBR unit cell model such as the multiplication factor, breeding ratio, change of amount of fuel, amount of fuel feeding, and neutron flux distribution. The result of MCNP6 and NEWT module in SCALE show some difference in depletion analysis, but it still seems that they can be used to analyze MSBR. Using these two computer code system, it is possible to analyze various parameters for the MSBR unit cells such as the multiplication factor, breeding ratio, amount of material, total feeding, and neutron flux distribution. Furthermore, the two code systems will be able to be used for analyzing other MSR model or whole core models of MSR.

  6. JAERI/U.S. collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Phase IIa and IIb experiments of JAERI/U.S. Collaborative Program on Fusion Blanket Neutronics have been performed using the FNS facility at JAERI. The phase IIa experimental systems consist of the Li2O test region, the rotating neutron target and the Li2CO3 container. In phase IIb, a beryllium layer is added to the inner wall to investigate a multiplier effect. Measured parameters are source characteristics by a foil activation method and spectrum measurements using both NE-213 and proton recoil counters. The measurements inside the Li2O region included tritium production rates, reaction rate by foil activation and neutron spectrum measurements. Analysis for these parameters was performed by using two dimensional discrete ordinate codes DOT3.5 and DOT-DD, and a Monte Carlo code MORSE-DD. The nuclear data used were based on JENDL3/PR1 and PR2. ENDF/B-IV, V and the FNS file were used as activation cross sections. The configurations analysed for the test region were a reference, a beryllium front and a beryllium sandwiched systems in phase IIa, and a reference and a beryllium front with first wall systems in phase IIb. This document describes the results of analysis and comparison between the calculations and the measurements. The prediction accuracy of key parameters in a fusion reactor blanket are examined. The tritium production rates can be well predicted in the reference systems but are fairly underestimated in the system with a beryllium multiplier. Details of experiments and the experimental techniques are described separately in the another report. (author)

  7. The prevalence of subclinical endometritis and intrauterine infections in repeat breeder cows.

    Science.gov (United States)

    Pothmann, H; Prunner, I; Wagener, K; Jaureguiberry, M; de la Sota, R L; Erber, R; Aurich, C; Ehling-Schulz, M; Drillich, M

    2015-05-01

    The objectives of this study were to assess the prevalence of subclinical endometritis and the presence of common uterine pathogens in repeat breeder cows. A total of 121 cows with three or more consecutive artificial inseminations without conception and no clinical signs of disease were defined as repeat breeder cows and were enrolled in this trial. Intrauterine samples were collected with the cytobrush technique to determine the prevalence of subclinical endometritis and bacteriologic infections. Blood samples were analyzed for concentrations of progesterone and estradiol in plasma to assess ovarian activity. Furthermore, breed, parity, history of calving and postpartum uterine infection, clinical findings of transrectal palpation, and backfat thickness were analyzed as potential factors for the prevalence of subclinical endometritis in repeat breeder cows. The prevalence of subclinical endometritis in repeat breeder cows was 12.7%; but common uterine pathogens, Escherichia coli and Trueperella pyogenes, were found in only one and three cows, respectively. Ovarian activity was determined in 95.0% of all cows. Recorded variables had no effect on the prevalence of subclinical endometritis in repeat breeder cows. In conclusion, subclinical endometritis and uterine infections linked to common pathogens were playing a minor role as a cause for repeat breeder cows in this study. Alternative reasons for failure to conceive in these cows are discussed. PMID:25670153

  8. Lithium-based oxide ceramics for tritium-breeding applications

    International Nuclear Information System (INIS)

    Material preparation techniques, crystallographic data, phase diagrams, metal compatibility, and thermal properties have been assembled for the lithium-based oxide ceramics designated as potential solid tritium breeders for fusion devices. The materials discussed in this report include: Li2O, β-Li5AlO4, γ-LiAlO2, Li4SiO4, Li2SiO3, Li4TiO4, Li2TiO3, Li8ZrO6, Li4ZrO4, and Li2ZrO3. The thermal properties covered were vaporization, thermal conductivity, specific heat, and linear thermal expansion. There has been no attempt to rank the above mentioned candidates, but rather to merely indicate points that must be considered when using the various materials as solid breeders. These encompass low lithium atom densities, destructive phase transformations, a higher thermal expansion, low thermal conductivity, excessive vaporization at low temperatures, corrosive nature toward metals and difficulty in sample preparation

  9. Conceptual design of a pool type molten salt breeder reactor

    International Nuclear Information System (INIS)

    The renewed interest in molten salt coolant technology is backed by the 50 years history of molten salt nuclear technology development, mainly in Oak Ridge National Laboratory (ORNL). In Indian context MSBR is found to be one of the options for sustainable nuclear energy generation, especially in the third stage of the nuclear programme. The system can be operated at high temperature which makes high efficiency power conversion and efficient hydrogen generation through thermo-chemical reactions possible. At present development is in progress in BARC on two molten salt reactor concepts, one is pool type and the other is loop type. Here the design of pool type concept with 850MWe power is described. The core is designed to operate in the fast spectrum region so the conversion of 233U breeding is possible from thorium. Preliminary thermal hydraulic analysis is carried out with LiF-ThF4-UF4 as the primary fuel and coolant. The blanket material is also a molten salt, LiF-ThF4. Reactor physics calculations are also carried out for the feasibility studies of the core design of the reactor. FLiNaK is used as the secondary coolant for the calculations. Both forced circulation and natural circulation options are evaluated. (author)

  10. Eddy current induced electromagnetic loads on shield blankets during plasma disruptions in ITER: A benchmark exercise

    International Nuclear Information System (INIS)

    According to recent updates of ITER shield blanket design, electromagnetic loads during the plasma disruption are being evaluated to verify the mechanical confidence and reliability. As a course of such evaluations, a benchmark activity for the electromagnetic analysis, coordinated by ITER Organization, is underway between ITER parties to compare the calculation results for disruption loads on the blankets. In this paper, we present calculation results for the electromagnetic loads on the simplified but practical model of ITER shield blankets with respect to six representative disruption scenarios of which ITER distributes simulation results based on the DINA code as a reference of the design and analysis. Commercial finite element method software, ANSYS/EmagTM, was employed to evaluate the eddy current on the blanket modules with the 40o sector model for major conducting structure of the tokamak including double-walled vacuum vessel, triangular support, and vertical targets of divertors. An interface between ANSYS/EmagTM and plasma simulator was implemented with a conversion tool assigning the plasma current density on the ANSYS elements corresponding to the current filaments in DINA outputs. Discussions are made of the possible improvement of the blanket model taking more realistic blanket configuration into account at the cost of the moderate increase in computational time. A final remark is given of the possibility of incorporating halo currents into ANSYS disruption simulations, which are major sources of electromagnetic loads on in-vessel components including blankets.

  11. Synthesis and Characterization of Fibre Reinforced Silica Aerogel Blankets for Thermal Protection

    Directory of Open Access Journals (Sweden)

    S. Chakraborty

    2016-01-01

    Full Text Available Using tetraethoxysilane (TEOS as the source of silica, fibre reinforced silica aerogels were synthesized via fast ambient pressure drying using methanol (MeOH, trimethylchlorosilane (TMCS, ammonium fluoride (NH4F, and hexane. The molar ratio of TEOS/MeOH/(COOH2/NH4F was kept constant at 1 : 38 : 3.73 × 10−5 : 0.023 and the gel was allowed to form inside the highly porous meta-aramid fibrous batting. The wet gel surface was chemically modified (silylation process using various concentrations of TMCS in hexane in the range of 1 to 20% by volume. The fibre reinforced silica aerogel blanket was obtained subsequently through atmospheric pressure drying. The aerogel blanket samples were characterized by density, thermal conductivity, hydrophobicity (contact angle, and Scanning Electron Microscopy. The radiant heat resistance of the aerogel blankets was examined and compared with nonaerogel blankets. It has been observed that, compared to the ordinary nonaerogel blankets, the aerogel blankets showed a 58% increase in the estimated burn injury time and thus ensure a much better protection from heat and fire hazards. The effect of varying the concentration of TMCS on the estimated protection time has been examined. The improved thermal stability and the superior thermal insulation of the flexible aerogel blankets lead to applications being used for occupations that involve exposure to hazards of thermal radiation.

  12. A passively-safe fusion reactor blanket with helium coolant and steel structure

    International Nuclear Information System (INIS)

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m2. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ''beryllium-joint'' concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket

  13. Attachment system for helium-cooled blanket of RF DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leshukov, A. E-mail: leshu@entek.ru; Blinov, Y.; Kovalenko, V.; Shatalov, G.; Strebkov, Y.; Strizhov, A

    2002-11-01

    The development of DEMO thermonuclear reactor is a part of Russian national program on the fusion process mastering. The DEMO-S (stationary thermonuclear reactor) should be the logic continuation of the ITER-type projects (pulse thermonuclear reactors) and the prototype for commercial power plants. DEMO reactor layout suggests to use the segmented blanket with mounting/dismounting procedure through the vacuum vessel vertical ports. Taking into account this layout the blanket attachment system has been developed and the present paper is devoted to this subject. The considered attachment system includes the lower and upper toroidal support assemblies which connect all the blanket segments in the enclosed ring. In it's turn the lower support assemblies attached to the vacuum vessel through the system of hinged support pillars. The heights of support pillars for inboard and outboard blankets are selected so that to indemnify the blanket massif thermal expansions in vertical and radial directions. The support pillars have been calculated on strength taking into account the electromagnetic loads from the plasma disruptions and blanket mass. The selection of high-strength chromium steel as a structural material for the support pillars could be considered as the results of strength analysis. The conclusions on the possibility to apply this attachment system for fusion reactor blanket and the critical issues are contained in this paper too.

  14. Selective breeding of Arabian and Thoroughbred racehorses in Algeria: perceptions, objectives and practices of owners-breeders

    Directory of Open Access Journals (Sweden)

    Safia Tennah

    2014-04-01

    Full Text Available This survey, conducted with 461 racehorse owners-breeders in Algeria between 2009 and 2011, investigates their perceptions, objectives and practices regarding selective breeding. Racehorse breeding is a full-time professional activity for a third of interviewees. The holdings are small-sized with 77% owning one or two mares. The regular practice of mating is here used to categorize breeders according to their degree of professionalization (38.4% professional vs. 61.6% occasional breeders. Experience in the sector was also used to classify breeders, considering as "junior" the breeders under 10 years experience (38.8% and as "senior" those above 10 years (61.2%. More than professionalization, experience shows a significant impact on practices and objectives. Thus, experience influences breed choice (junior breeders tend to specialize while senior own both Arabian and Thoroughbreds, age at first foaling (sooner among senior breeders, information sources considered for selecting stallions (senior use more diversified sources, the importance granted to the price of mating (greater for junior breeders, the importance granted to the ranking compared to earnings (the ranking being more important to junior breeders, and the priority given to breeding (junior breeders give higher priority to a buy-race-resell activity. Finally, racehorse breeding is poorly professionalized, the only financial goal being cost coverage. Despite inappropriate practices, an interest for selection is noticed.

  15. Ceramic Solar Receiver

    Science.gov (United States)

    Robertson, C., Jr.

    1984-01-01

    Solar receiver uses ceramic honeycomb matrix to absorb heat from Sun and transfer it to working fluid at temperatures of 1,095 degrees and 1,650 degrees C. Drives gas turbine engine or provides heat for industrial processes.

  16. Ceramic fiber filter technology

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, B.L.; Janney, M.A.

    1996-06-01

    Fibrous filters have been used for centuries to protect individuals from dust, disease, smoke, and other gases or particulates. In the 1970s and 1980s ceramic filters were developed for filtration of hot exhaust gases from diesel engines. Tubular, or candle, filters have been made to remove particles from gases in pressurized fluidized-bed combustion and gasification-combined-cycle power plants. Very efficient filtration is necessary in power plants to protect the turbine blades. The limited lifespan of ceramic candle filters has been a major obstacle in their development. The present work is focused on forming fibrous ceramic filters using a papermaking technique. These filters are highly porous and therefore very lightweight. The papermaking process consists of filtering a slurry of ceramic fibers through a steel screen to form paper. Papermaking and the selection of materials will be discussed, as well as preliminary results describing the geometry of papers and relative strengths.

  17. Advanced Ceramics Property Measurements

    Science.gov (United States)

    Salem, Jonathan; Helfinstine, John; Quinn, George; Gonczy, Stephen

    2013-01-01

    Mechanical and physical properties of ceramic bodies can be difficult to measure correctly unless the proper techniques are used. The Advanced Ceramics Committee of ASTM, C-28, has developed dozens of consensus test standards and practices to measure various properties of a ceramic monolith, composite, or coating. The standards give the "what, how, how not, and why" for measurement of many mechanical, physical, thermal, and performance properties. Using these standards will provide accurate, reliable, and complete data for rigorous comparisons with other test results from your test lab, or another. The C-28 Committee has involved academics, producers, and users of ceramics to write and continually update more than 45 standards since the committee's inception in 1986. Included in this poster is a pictogram of the C-28 standards and information on how to obtain individual copies with full details or the complete collection of standards in one volume.

  18. Making Ceramic Cameras

    Science.gov (United States)

    Squibb, Matt

    2009-01-01

    This article describes how to make a clay camera. This idea of creating functional cameras from clay allows students to experience ceramics, photography, and painting all in one unit. (Contains 1 resource and 3 online resources.)

  19. Development of the breeding blanket and shield model for the fusion power reactors system SYCOMORE

    Energy Technology Data Exchange (ETDEWEB)

    Li-Puma, Antonella, E-mail: antonella.lipuma@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Jaboulay, Jean-Charles, E-mail: Jean-Charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Martin, Brunella, E-mail: brunella.martin@gmail.com [Incka, 19-21 Rue du 8 mai 1945, F-94110 Arcueil (France)

    2014-10-15

    SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory. Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper. For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available. Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE. In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006–2007 studies) is also presented.

  20. Selecting Ceramics - Introduction

    OpenAIRE

    Cassidy, M.

    2002-01-01

    AIM OF PRESENTATION: To compare a number of materials for extracoronal restoration of teeth with particular reference to CAD-CAM ceramics. CASE DESCRIPTION AND TREATMENT CARRIED OUT: This paper will be illustrated using clinical examples of patients treated using different ceramic restorations to present the advantages and disadvantages and each technique. The different requirements of tooth preparation, impression taking and technical procedures of each system will be presented and compar...

  1. Preliminary Analysis of Liquid Metal MHD Pressure Drop in the Blanket for the FDS

    Institute of Scientific and Technical Information of China (English)

    王红艳; 吴宜灿; 何晓雄

    2002-01-01

    Preliminary analysis and calculation of liquid metal Li17Pb83 magnetohydrodynamic (MHD) pressure drop in the blanket for the FDS have been presented to evaluate the significance of MHD effects on the thermal-hydraulic design of the blanket. To decrease the liquid metal MHD pressure drop, Al2O3 is applied as an electronically insulated coating onto the inner surface of the ducts. The requirement for the insulated coating to reduce the additional leakage pressure drop caused by coating imperfections has been analyzed. Finally, the total liquid metal MHD pressure drop and magnetic pump power in the FDS blanket have been given.

  2. Realization of a flat fission power density in a hybrid blanket over long operation periods

    International Nuclear Information System (INIS)

    A straightforward numerical graphical method is applied to achieve a flat fission power density (FPD) in a hybrid blanket by using a mixed fuel (ThO2 and natural UO2) with variable fractions of the fuel components in the radial direction. The neutronic analysis is carried out on a blanket with a hard neutron spectrum in the fissionable zone by simply omitting the moderating beryllium neutron multiplier. Mainly due to this precaution in the blanket design, the FPD could be kept quasi-constant over a relatively long plant lifetime

  3. Clinical application of bio ceramics

    Science.gov (United States)

    Anu, Sharma; Gayatri, Sharma

    2016-05-01

    Ceramics are the inorganic crystalline material. These are used in various field such as biomedical, electrical, electronics, aerospace, automotive and optical etc. Bio ceramics are the one of the most active areas of research. Bio ceramics are the ceramics which are biocompatible. The unique properties of bio ceramics make them an attractive option for medical applications and offer some potential advantages over other materials. During the past three decades, a number of major advances have been made in the field of bio ceramics. This review focuses on the use of these materials in variety of clinical scenarios.

  4. Ceramic electrolyte coating and methods

    Science.gov (United States)

    Seabaugh, Matthew M.; Swartz, Scott L.; Dawson, William J.; McCormick, Buddy E.

    2007-08-28

    Aqueous coating slurries useful in depositing a dense coating of a ceramic electrolyte material (e.g., yttrium-stabilized zirconia) onto a porous substrate of a ceramic electrode material (e.g., lanthanum strontium manganite or nickel/zirconia) and processes for preparing an aqueous suspension of a ceramic electrolyte material and an aqueous spray coating slurry including a ceramic electrolyte material. The invention also includes processes for depositing an aqueous spray coating slurry including a ceramic electrolyte material onto pre-sintered, partially sintered, and unsintered ceramic substrates and products made by this process.

  5. Comparison of serum leptin, glucose, total cholesterol and total protein levels in fertile and repeat breeder cows

    Directory of Open Access Journals (Sweden)

    Saime Guzel

    2014-12-01

    Full Text Available In the present study we measured serum glucose, leptin, total cholesterol and total protein concentrations in repeat breeder cows and compared them with fertile cows. For this aim, 20 repeat breeder cows and 20 fertile cows were used as material. Repeat breeder cows were found to have lower levels of leptin and glucose as compared with fertile ones. No significant differences in total cholesterol and total protein levels were observed between the two groups. No significant correlation of leptin with glucose, total cholesterol and total protein was observed in fertile and repeat breeder cows. Low concentrations of glucose and leptin can have some effects on reproductive problems as repeat breeder and help to understand potential mechanisms impairing fertility in repeat breeder cows.

  6. Manufacturing of prototype fast breeder reactor components: challenges and achievements

    International Nuclear Information System (INIS)

    In the presentation, three components of 500 MWe Prototype Fast Breeder Reactor (PFBR), viz. grid plate, roof slab and fuel handling systems, are focused, which have been responsible for the considerable delay of the project schedule. The manufacturing challenges of grid plate mainly originated from large number of sleeves resulting in higher self weight and hard facing of large diameter sleeves. Machining of large diameter plates and shell assembly to the required tight tolerances on dimensions, hard facing with nickel based cobalt free hard facing material on continuous, large diameter (6.7 m) annular tracks, heat treatment of large austenitic stainless steel parts at 1050℃ with controlled rates of cooling and heating together with control on temperature gradient across the parts, complex assembly of a large number of parts (∼14900) meeting the important requirements on verticality of sleeve assemblies (Ø0.1 mm) and delicate handling and transportation are truly challenging activities in the manufacturing technology. In case of roof slab, complex manufacturing process, especially welding between the shell and stiffeners caused lamellar tearing problems and extensive testing time. Inclined fuel transfer machine, multiple repairs, heavy weight and testing strategy resulted in long manufacturing and testing time. Some general lessons learnt are also brought out in this presentation. Technology development prior to start of construction is essential for long delivery components. Judicious choice of tolerances, number and location of welds and inspections has to be made. Robust criteria need to be applied for the acceptance of manufacturing deviations and material compositions. Indigenous materials should be used after qualifications of manufacturing process of direct relevance apart from routine standards. From the rich experience gained through the manufacture and erection of reactor assembly components of PFBR, important guidelines and approaches were derived

  7. Defect assessment procedure: A french approach for fast breeder reactors

    International Nuclear Information System (INIS)

    As a result of a collaborative effort between Commissariat a l'Energie Atomique, Electricite de France, and NOVATOME to produce and improve rules for fast breeder reactors, RCC-MR, an interim defect assessment procedure is now available in the first draft version (appendix A16). This procedure addresses defects detected during in-service inspection for reactor components operating at moderate or high temperature conditions. Three stages have been considered: initiation, propagation under cyclic loading with or without holdtime and crack instability by ductile and creep rupture. For each of these topics, procedures and rules based on fracture mechanics are proposed. Prediction of initiation is obtained by a simplified method named σd method which relies on the evaluation of the real stress-strain history on a small distance d (d = 0.05 mm for 316L(N) austenitic steel) close to the crack front and material characteristics (limiting stresses) that are available in nuclear codes. This method has been developed for fatigue, creep and creep-fatigue conditions. Defect growth assessment is performed for fatigue and creep-fatigue conditions. For creep-fatigue conditions, fatigue and creep crack growth per cycle are calculated separately and the total crack extension is taken as the sum of the two contributions. Extensive use of simplified method for estimating J (Js method) is made and developed when mechanical and thermal loadings are specified. On the final defect size, assessment may be made in order to avoid crack instability by ductile and creep rupture and collapse load on the remaining. The organization and contents of the present version of this appendix A16 is described. An overview of each specific rule is given

  8. Network Representation of Design Knowledge of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    A method of design knowledge representation was studied for the Japanese fast breeder reactor Monju, aiming at enhanced understanding of engineering considerations with mutual relations. Taking over design knowledge of Monju to next generation designers/engineers to be in charge of design of future FRs is by no means easy, in contrast with operation and maintenance knowledge which can be acquired in the real plant operation and maintenance. Specifications of the as-is Monju contains only a small part of the entire design knowledge, mainly by two reasons. Firstly, reasons for selecting the as-is specifications can not be understood until reaching proper knowledge source. Secondly, there are many rejected options on the design specifications. Design specifications are selected along with technical dependencies among a huge number and diversified specification items. Decisions design are made basically along with these dependencies which can hardly be traced in the currently available database or document libraries. Reasons for the rejections of options need to be profoundly understood, because those are not certainly due to technical inferiority. Some of rejected options can be worth reconsidering in the future, possibly by technical advances in materials, high-precision prediction software tools, rationalized standards/code, etc. The authors propose a new design knowledge representation approach based on networking of knowledge nodes along with the mutual dependencies. A prototype software has been developed and a basic performance test was made to visualize the dependency network. An additional function to enable design case studies on hypothetical adoptions of rejected options is now under consideration. (author)

  9. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    In the power increase performance test of the experimental fast reactor ''Joyo'', which was in progress since April, the first stage of the rated thermal output of 50 MW has been accomplished on July 5. Thereafter, the continuous opeation test at 50 MW for 100 hours was performed for the verification of its overall operational performance from August 13 to 16. The safety evaluation for power increase up to 75 MW and 100 MW, which was under way since September, last year, was completed, and the power increase was licensed on September 20. Concerning the design of the prototype fast breeder reactor ''Monju'', the studies on the specifications of the Construction Preliminary Design (2) have been finished. In respect of the analysis and preparation of materials for the Safety Licensing by the Committee, the developments of the analytical codes for rupture propagation in the heat transfer tubes of steam generators and for decay heat have been conducted. In the construction site surveys, the third geological structure survey and beach deformation survey have all ended, while the meteorological and seismic observations, the prediction of the diffusion of drained warm water, the survey of river flow, etc. are now under way. A report on the survey conducted on the construction site in Shiraki was received by the Fukui prefectural government in July, and the copies of a report on the assessment of environmental effect were submitted in August to both the national government and the Fukui prefectural government. The situations of progress of the research and development works on reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structural materials, safety and steam generators are reported. (Nakai, Y.)

  10. THE EFFECT OF BROILER BREEDER AGES ON THE QUALITIVE AND QUANTITIVE PROPERTIES OF THE EGG

    Directory of Open Access Journals (Sweden)

    Shahabodin Gharahveysi

    2012-01-01

    Full Text Available Broiler breeder age is one of the most important factors that affects on egg properties. However by increasing the age of broiler breeder flock, the quality of eggs and consequently the quality of their chick products would be changed. In order to study the effect of broiler breeder flock age on the various aspects of chick products, 300 fertilizable eggs were selected randomly, from 3 broiler breeder farms. Selected eggs were collected from broilers that were 30 (young, 36 (peak, 43(after peak, 53 (old, 60 (very old and 82 (molted weeks old. Collected eggs were stored during 3 to 4 days in the ambient temperature. Qualitive and quantitive aspects of eggs including egg weight, albumen height yolk, height thickness of egg shell, yolk color, Albumen and yolk PH and Hugh unit were studied. Trait Analysis was done by ANOVA procedure of SAS statistical software. To compare the means, Duncan test was used. The effects of age and breeder farms on the egg weight, yolk color, yolk and albumen PH, yolk and albumen height, shell egg thickness and Hugh unit were significant (p<0.05. The lowest difference was seen between the age of 53, 60 and 82 week old (p<0.05. Albumen PH and alkaline is increased by increase of age, but yolk PH is the variance. Influence of age on the traits including yolk and albumen height and Hugh unit was decreased and the color of yolk was faded by increasing age. According to obtained results from these research performances of ages of 53, 60 and 82 weeks are close together. We could conclude that older broiler breeder flocks are produce the better qualitive and quantitive properties of egg products. Since most poultry enough information about the quality of breeder chickens and the best age to have chickens, using the results of this study can be answered many questions.

  11. Selective breeding of Arabian and Thoroughbred racehorses in Algeria: perceptions, objectives and practices of owners-breeders

    OpenAIRE

    Safia Tennah; Frédéric Farnir; Nacerredine Kafidi; Ibrahim Njikam Nsangou; Pascal Leroy; Nicolas Antoine-Moussiaux

    2014-01-01

    This survey, conducted with 461 racehorse owners-breeders in Algeria between 2009 and 2011, investigates their perceptions, objectives and practices regarding selective breeding. Racehorse breeding is a full-time professional activity for a third of interviewees. The holdings are small-sized with 77% owning one or two mares. The regular practice of mating is here used to categorize breeders according to their degree of professionalization (38.4% professional vs. 61.6% occasional breeders). Ex...

  12. The Effect of Dietary Supplementation of Prebiotic and Probiotic on Performance, Humoral Immunity Responses and Egg Hatchability in Broiler Breeders

    OpenAIRE

    Hajati H; Hassanabadi A; Teimouri Yansari A

    2014-01-01

    In this experiment, the influence of prebiotic and probiotic supplementation in the broiler breeder diets on body weight, mortality, feed intake, egg production, hatchability and humoral immunity response was investigated. A total number of 13140 female and 1260 male breeders (Cobb 500) with 26 wks of age were allocated to three treatments with six replicates (800 birds each replicate). Breeders were fed control basal diet, basal diet supplemented with prebiotic (mannan oligosaccharide) or pr...

  13. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    International Nuclear Information System (INIS)

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core

  14. Development of ITER shielding blanket prototype mockup by HIP bonding

    International Nuclear Information System (INIS)

    A prototype (∼900H x 1700W x 350T mm) of the ITER shielding blanket module has been fabricated following the previous successful fabrication of a small-scale (∼500H x 400W x 150T mm) and mid-scale (∼800H x 500W x 350T mm) mock-ups. This prototype incorporates most of key design features essential to the fabrication of the ITER shielding blanket module such as 1) the first wall heat sink made of Al2O3 dispersion strengthened Cu (DSCu) with built-in SS316L coolant tubes bonded to a massive SS316LN shield block, 2) toroidally curved first wall with a radius of 5106 mm while straight in poloidal direction, 3) coolant channels oriented in poloidal direction in the first wall and in toroidal direction in the shield block, 4) the first wall coolant channel routing to avoid the interference with the front access holes, 5) coolant channels drilled through the forged SS316LN-IG shield block, and 6) four front access holes of 30 mm in diameter penetrated through the first wall and the shield block. For the joining method, especially for the first wall/side wall parts and the shield block, the solid HIP (Hot Isostatic Pressing) process was applied. It is difficult to apply conventional joining methods such as field welding, brazing, explosion bonding and mechanical one-axial diffusion bonding to a wide area bonding because sufficient mechanical strengths can not be obtained and excessive deformations occurs. In order to solve these fabrication issues, HIP bonding was applied. The first wall stainless steel (SS) coolant tubes of 10 mm in inner diameter and l mm in thickness were sandwiched by semi-circular grooved DSCu plates at the first wall and the front region of the side wall, and by semi-circular grooved SS plates at the back region of the side wall. After assembling of these first wall/side wall parts with the shield block, they were simultaneously bonded by single step HIP in order to minimize thermal effects on the mechanical properties and to reduce the number

  15. Integrated-blanket-coil (IBC) concept applied to the OH-coil for spherical tori

    International Nuclear Information System (INIS)

    This concept combines blanket and coil functions into a single component. The objectives of the concept are to: (1) provide design options, (2) simplify overall configuration, (3) enhance compactness, and (4) reduce costs. Some drawings of the system are given

  16. Main maintenance operations for Test Blanket Systems in ITER TBM port cells

    International Nuclear Information System (INIS)

    Highlights: • The Test Blanket System components layout in Port Cell room is described. • The maintenance of the two Test Blanket Systems in ITER port cell is addressed. • The overall replacement/maintenance strategy is defined. • The main maintenance tasks of the systems are discussed. • The maintenance strategy and required tools are presented. -- Abstract: Each Test Blanket System in ITER is formed by an in-vessel component, the Test Blanket Module, and several associated ancillary systems (coolant and Tritium systems, instrumentation and control systems). The paper describes the overall replacement/maintenance strategy and the main maintenance tasks that have to be considered in the design of the systems. It shows that there are no critical issues

  17. Acoustic contributions of a sound absorbing blanket placed in a double panel structure: Absorption Versus Transmission

    CERN Document Server

    Doutres, Olivier; 10.1121/1.3458845

    2010-01-01

    The objective of this paper is to propose a simple tool to estimate the absorption vs. transmission loss contributions of a multilayered blanket unbounded in a double panel structure and thus guide its optimization. The normal incidence airborne sound transmission loss of the double panel structure, without structure-borne connections, is written in terms of three main contributions; (i) sound transmission loss of the panels, (ii) sound transmission loss of the blanket and (iii) sound absorption due to multiple reflections inside the cavity. The method is applied to four different blankets frequently used in automotive and aeronautic applications: a non-symmetric multilayer made of a screen in sandwich between two porous layers and three symmetric porous layers having different pore geometries. It is shown that the absorption behavior of the blanket controls the acoustic behavior of the treatment at low and medium frequencies and its transmission loss at high frequencies. Acoustic treatment having poor sound ...

  18. Normal Operation (NO) of APT Blanket System and its Components Based on Initial Conceptual Design

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report (PSAR) for the APT.

  19. Li ceramic pebbles chemical compatibility with Eurofer samples in fusion relevant conditions

    International Nuclear Information System (INIS)

    Information on the chemical compatibility between Li ceramic breeders and reactor structural materials is an important issue for fusion reactor technology. In this work, Eurofer samples were placed inside a Li ceramic pebble bed and kept at 600 deg. C under a reducing atmosphere obtained by the flow of a purging gas (He + 0.1vol.%H2). Titanate and orthosilicate Li pebble beds were used in the experiments and exposure time ranged from 50 to 2000 h. Surface chemical reactions were investigated with nuclear microprobe techniques. The orthosilicate pebbles present chemical reactions even with the gas mixture, whereas for the samples in close contact with Eurofer there is evidence of Eurofer elemental diffusion into the pebbles and the formation of different types of compounds. Although the titanate pebbles used in the chemical compatibility experiments present surface alterations with increasing surface irregularities along the annealing time, there is no clear indication of Eurofer constituents diffusion

  20. Resolution of proliferation issues for a SFR blanket with a specific application

    Energy Technology Data Exchange (ETDEWEB)

    Stauff, N.E. [31 rue baudelaire, voisins le bretonneux, 78960 (France); Massachusetts Institute of Technology, 77 Massachusetts Ave, Cambridge, MA 02139 (United States); Forget, B.; Driscoll, M.J. [Massachusetts Institute of Technology, 77 Massachusetts Ave, Cambridge, MA 02139 (United States)

    2009-06-15

    The Sodium Fast Reactor is seen as the most realistic Gen-IV reactor to be built in the near future. France and the US are still developing their designs; these will require improved safety, competitive economics, and also proliferation resistance. To meet this last requirement, both French and American designers show some concerns with the use of breeding blankets. France and the USA won't need breeding blankets to produce plutonium because they already have large amounts of plutonium bred from their LWR fleet to start a new SFR fleet, thus breeding blankets are mainly of interest for minor actinide burning. On the contrary, India and China express great interest in blankets for their SFR designs, to reach a positive breeding gain. For example, the Indian PFBR, a 500 MWe oxide-fueled SFR has a breeding ratio of 1.05. Blankets are used in a Fast Reactor to increase the breeding ratio of the core, by breeding a significant amount of plutonium. The Plutonium bred within these blankets, if these are loaded with Uranium only, is generally of a very high quality, which makes it easily used in a nuclear explosive device. Our research has shown that the plutonium in breeding blankets can be made less attractive to make a nuclear explosive device than LWR-bred plutonium with a burnup of 50 MWd/Kg. Minor actinide doping and moderator addition were the two options studied, as they increase Pu{sup 238} and Pu{sup 240} production. In the work reported here, the methodology developed for securing a breeding blanket was successfully applied to the Indian PFBR. The full paper will describe a design of the PFBR breeding proliferation resistant plutonium within its blankets. The blankets were rendered secure by adding a zirconium hydride moderator and a small volume of MAs. It was demonstrated that reducing the attractiveness of the blanket plutonium would require no external MA dependency by choosing an adequate fuel cycle. The characteristics and performance of this design

  1. Climate-driven expansion of blanket bogs in Britain during the Holocene

    Directory of Open Access Journals (Sweden)

    A. V. Gallego-Sala

    2015-10-01

    Full Text Available Blanket bog occupies approximately 6 % of the area of the UK today. The Holocene expansion of this hyperoceanic biome has previously been explained as a consequence of Neolithic forest clearance. However, the present distribution of blanket bog in Great Britain can be predicted accurately with a simple model (PeatStash based on summer temperature and moisture index thresholds, and the same model correctly predicts the highly disjunct distribution of blanket bog worldwide. This finding suggests that climate, rather than land-use history, controls blanket-bog distribution in the UK and everywhere else. We set out to test this hypothesis for blanket bogs in the UK using bioclimate envelope modelling compared with a database of peat initiation age estimates. We used both pollen-based reconstructions and climate model simulations of climate changes between the mid-Holocene (6000 yr BP, 6 ka and modern climate to drive PeatStash and predict areas of blanket bog. We compiled data on the timing of blanket-bog initiation, based on 228 age determinations at sites where peat directly overlies mineral soil. The model predicts large areas of northern Britain would have had blanket bog by 6000 yr BP, and the area suitable for peat growth extended to the south after this time. A similar pattern is shown by the basal peat ages and new blanket bog appeared over a larger area during the late Holocene, the greatest expansion being in Ireland, Wales and southwest England, as the model predicts. The expansion was driven by a summer cooling of about 2 °C, shown by both pollen-based reconstructions and climate models. The data show early Holocene (pre-Neolithic blanket-bog initiation at over half of the sites in the core areas of Scotland, and northern England. The temporal patterns and concurrence of the bioclimate model predictions and initiation data suggest that climate change provides a parsimonious explanation for the early Holocene distribution and later

  2. Piezoelectric Ceramics and Their Applications

    Science.gov (United States)

    Flinn, I.

    1975-01-01

    Describes the piezoelectric effect in ceramics and presents a quantitative representation of this effect. Explains the processes involved in the manufacture of piezoelectric ceramics, the materials used, and the situations in which they are applied. (GS)

  3. Cooled Ceramic Turbine Vane Project

    Data.gov (United States)

    National Aeronautics and Space Administration — N&R Engineering will investigate the feasibility of cooled ceramics, such as ceramic matrix composite (CMC) turbine blade concepts that can decrease specific...

  4. Wash resistance and repellent properties of Africa University mosquito blankets against mosquitoes

    OpenAIRE

    Lukwa, N; A. Makuwaza; T. Chiwade; S.L. Mutambu; M. Zimba; P. Munosiyei

    2013-01-01

    The effect of permethrin-treated Africa University (AU) mosquito blankets on susceptible female Anopheles gambiae sensu lato mosquitoes was studied under laboratory conditions at Africa University Campus in Mutare, Zimbabwe. Wash resistance (ability to retain an effective dose that kills ≥80% of mosquitoes after a number of washes) and repellence (ability to prevent ≥80% of mosquito bites) properties were studied. The AU blankets were wash resistant when 100% mortality was recorded up t...

  5. Application Effect’s Research of Vetiver Eco Blanket in Pubugou Reservoir Fluctuating Zone

    OpenAIRE

    Lan Huijuan; Zheng Kaiyuan; Ni Fuquan; Deng Yu; Liu Dengyu; Yang Xinwei; Wang Tao

    2015-01-01

    To solve the ecological disasters in Pubugou Reservoir Fluctuating Zone, ecological blanket governance model is proposed in this paper, which may provide good early environment for plants’ survival in fluctuation zone, and then play the function of greening and sustainable development to ensure the slopes’ stability. Meanwhile, based on the result of vetiver ecological blanket in Hanyuan experimental zone, we find that three kinds of typical Fluctuating Zone slope’s greening effect is good, w...

  6. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2010-06-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  7. Implementation of two-phase tritium models for helium bubbles in HCLL breeding blanket modules

    OpenAIRE

    Fradera, Jordi; Sedano, L.A.; Mas de les Valls Ortiz, Elisabet; Batet Miracle, Lluís

    2011-01-01

    Tritium self-sufficiency requirement of future DT fusion reactors involves large helium production rates in the breeding blankets; this might impact on the conceptual design of diverse fusion power reactor units, such as Liquid Metal (LM) blankets. Low solubility, long residence-times and high production rates create the conditions for Helium nucleation, which could mean effective T sinks in LM channels. A model for helium nano-bubble formation and tritium conjugate transport phen...

  8. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2007-08-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  9. Numerical-experimental analyses by Hot-Wire method of an alumina cylinder for future studies on thermal conductivity of the fusion breeder materials

    International Nuclear Information System (INIS)

    The determination of the thermal conductivity of breeder materials is one of the main goal in order to find the best candidate material for the fusion reactor technology. Experimental tests have been and will be carried out with a dedicated experimental devices, built at the Department of Civil and Industrial Engineering of the University of Pisa. The methodological approach used in doing that is characterized by two main phases strictly interrelated each other: the first one focused on the experimental evaluation of thermal conductivity of a ceramic material, by means of hot wire method, to be subsequently used in the second phase, based on the test rig method, to determine the thermal conductivity of pebble bed material. To the purpose, two different experimental devices have been designed and built. This paper deals with the first phase of the methodology. In this framework, the equipment set up and built to perform Hot wire tests, the ceramic material (a cylinder of alumina), the experimental procedure and the measured results obtained varying the temperature, are presented and discussed. The experimental campaign has been lead from 50°C up to 400°C. The thermal conductivity of the ceramic material at different bulk temperatures has been obtained in stationary conditions (detected on the basis of the temperature values measured during the experiment). Numerical analyses have been also performed by means of FEM code Ansys©. The numerical results were in quite good agreement with the experimental one, confirming also the reliability of code in reproducing heat transfer phenomena

  10. Mechanical properties of ceramics

    CERN Document Server

    Pelleg, Joshua

    2014-01-01

    This book discusses the mechanical properties of ceramics and aims to provide both a solid background for undergraduate students, as well as serving as a text to bring practicing engineers up to date with the latest developments in this topic so they can use and apply these to their actual engineering work.  Generally, ceramics are made by moistening a mixture of clays, casting it into desired shapes and then firing it to a high temperature, a process known as 'vitrification'. The relatively late development of metallurgy was contingent on the availability of ceramics and the know-how to mold them into the appropriate forms. Because of the characteristics of ceramics, they offer great advantages over metals in specific applications in which hardness, wear resistance and chemical stability at high temperatures are essential. Clearly, modern ceramics manufacturing has come a long way from the early clay-processing fabrication method, and the last two decades have seen the development of sophisticated technique...

  11. OXYGEN TRANSPORT CERAMIC MEMBRANES

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Sukumar Bandopadhyay; Dr. Nagendra Nagabhushana

    2000-10-01

    This is the third quarterly report on oxygen Transport Ceramic Membranes. In the following, the report describes the progress made by our university partners in Tasks 1 through 6, experimental apparatus that was designed and built for various tasks of this project, thermodynamic calculations, where applicable and work planned for the future. (Task 1) Design, fabricate and evaluate ceramic to metal seals based on graded ceramic powder/metal braze joints. (Task 2) Evaluate the effect of defect configuration on ceramic membrane conductivity and long term chemical and structural stability. (Task 3) Determine materials mechanical properties under conditions of high temperatures and reactive atmospheres. (Task 4) Evaluate phase stability and thermal expansion of candidate perovskite membranes and develop techniques to support these materials on porous metal structures. (Task 5) Assess the microstructure of membrane materials to evaluate the effects of vacancy-impurity association, defect clusters, and vacancy-dopant association on the membrane performance and stability. (Task 6) Measure kinetics of oxygen uptake and transport in ceramic membrane materials under commercially relevant conditions using isotope labeling techniques.

  12. Diffusion in ceramics

    CERN Document Server

    Pelleg, Joshua

    2016-01-01

    This textbook provides an introduction to changes that occur in solids such as ceramics, mainly at high temperatures, which are diffusion controlled, as well as presenting research data. Such changes are related to the kinetics of various reactions such as precipitation, oxidation and phase transformations, but are also related to some mechanical changes, such as creep. The book is composed of two parts, beginning with a look at the basics of diffusion according to Fick's Laws. Solutions of Fick’s second law for constant D, diffusion in grain boundaries and dislocations are presented along with a look at the atomistic approach for the random motion of atoms. In the second part, the author discusses diffusion in several technologically important ceramics. The ceramics selected are monolithic single phase ones, including: A12O3, SiC, MgO, ZrO2 and Si3N4. Of these, three refer to oxide ceramics (alumina, magnesia and zirconia). Carbide based ceramics are represented by the technologically very important Si-ca...

  13. Activation Characteristics of Fuel Breeding Blanket Module in Fusion Driven Subcritical System

    Institute of Scientific and Technical Information of China (English)

    HUANG Qun-Ying; LI Jian-Gang; CHEN Yi-Xue

    2004-01-01

    @@ Shortage of energy resources and production of long-lived radioactivity wastes from fission reactors are among the main problems which will be faced in the world in the near future. The conceptual design of a fusion driven subcritical system (FDS) is underway in Institute of Plasma Physics, Chinese Academy of Sciences. There are alternative designs for multi-functional blanket modules of the FDS, such as fuel breeding blanket module (FBB)to produce fuels for fission reactors, tritium breeding blanket module to produce the fuel, i.e. tritium, for fusion reactor and waste transmutation blanket module to try to permanently dispose of long-lived radioactivity wastes from fission reactors, etc. Activation of the fuel breeding blanket of the fusion driven subcritical system (FDS-FBB) by D-T fusion neutrons from the plasma and fission neutrons from the hybrid blanket are calculated and analysed under the neutron wall loading 0.5 MW/m2 and neutron fluence 15 MW. yr/m2. The neutron spectrum is calculated with the worldwide-used transport code MCNP/4C and activation calculations are carried out with the well known European inventory code FISPACT/99 with the latest released IAEA Fusion Evaluated Nuclear Data Library FENDL-2.0 and the ENDF/B-V uranium evaluated data. Induced radioactivities, dose rates and afterheats, etc, for different components of the FDS-FBB are compared and analysed.

  14. Assessment of ceramic membrane filters

    Energy Technology Data Exchange (ETDEWEB)

    Ahluwalia, R.K.; Geyer, H.K.; Im, K.H. [and others

    1995-08-01

    The objectives of this project include the development of analytical models for evaluating the fluid mechanics of membrane coated, dead-end ceramic filters, and to determine the effects of thermal and thermo-chemical aging on the material properties of emerging ceramic hot gas filters. A honeycomb cordierite monolith with a thin ceramic coating and a rigid candle filter were evaluated.

  15. Detailed 3-D nuclear analysis of ITER blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, T.D., E-mail: tdbohm@wisc.edu [University of Wisconsin-Madison, Madison, WI (United States); Sawan, M.E.; Marriott, E.P.; Wilson, P.P.H. [University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, M.; Bullock, J. [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-10-15

    In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded.

  16. Vacuum Permeator Analysis for Extraction of Tritium from DCLL Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Humrickhouse, Paul Weston [Idaho National Laboratory; Merrill, Brad Johnson [Idaho National Laboratory

    2014-11-01

    It is envisioned that tritium will be extracted from DCLL blankets using a vacuum permeator. We derive here an analytical solution for the extraction efficiency of a permeator tube, which is a function of only two dimensionless numbers: one that indicates whether radial transport is limited in the PbLi or in the solid membrane, and another that is the ratio of axial and radial transport times in the PbLi. The permeator efficiency is maximized by decreasing the velocity and tube diameter, and increasing the tube length. This is true regardless of the mass transport correlation used; we review several here and find that they differ little, and the choice of correlation is not a source of significant uncertainty here. The PbLi solubility, on the other hand, is a large source of uncertainty, and we identify upper and lower bounds from the literature data. Under the most optimistic assumptions, we find that a ferritic steel permeator operating at 550 °C will need to be at least an order of magnitude larger in volume than previous conceptual designs using niobium and operating at higher temperatures.

  17. Surface property variations in Venusian fluidized ejecta blanket craters

    Science.gov (United States)

    Johnson, Jeffrey R.; Baker, Victor R.

    1994-01-01

    A comprehensive study of Magellan Cycles 1 and 2 radar data from Venus reveals surface roughness and dielectric variations associated with fluidized ejecta blanket (FEB) craters that help illuminate styles of flow ejecta emplacement. This study develops new procedures of digital unit mapping and polygon-filling algorithms using Magellan synthetic aperture radar (SAR), altimetry, and radiometry data. These techniques allow the extraction of radiophysical information for FEB crater materials, nearby plains, and lava flows. Backscatter curve slopes of the FEBs studied here are consistent with surface textures that are transitional between a'a and pahoehoe-like. Average surface property values of ejecta units are relatively similar for a given crater, but are discernibly different from other craters. Individual crater ejecta reflectivity and emissivity values are relatively similar to those for the surrounding plains, which may suggest a link between plains material and ejecta dielectric properties. Increasing FEB roughness downflow are interpreted to be associated with more lava-like flows, while decreasing roughness are more similar to trends typical of gravity (pyroclastic-like or debris-like) flows. Most commonly, FEB crater flow materials exhibit transitions from proximal, lava/melt-like flow styles to distal, gravity flow-like styles. Some FEBs show more complicated behavior, however, or appear to be more dominated by dielectric differences downflow, as inferred from correlations between the data sets. Such transitions may result from changes in local topography or from overlapping of flow lobes during FEB emplacement.

  18. Ceramic vane drive joint

    Science.gov (United States)

    Smale, Charles H. (Inventor)

    1981-01-01

    A variable geometry gas turbine has an array of ceramic composition vanes positioned by an actuating ring coupled through a plurality of circumferentially spaced turbine vane levers to the outer end of a metallic vane drive shaft at each of the ceramic vanes. Each of the ceramic vanes has an end slot of bow tie configuration including flared end segments and a center slot therebetween. Each of the vane drive shafts has a cross head with ends thereof spaced with respect to the sides of the end slot to define clearance for free expansion of the cross head with respect to the vane and the cross head being configured to uniformly distribute drive loads across bearing surfaces of the vane slot.

  19. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition

  20. Development of high nitrogen electrodes for fast breeder reactor applications

    International Nuclear Information System (INIS)

    Austenitic stainless steels of AISI type 316 (316 SS) and its variants are used extensively as structural material for the components of fast reactors operating at temperature up to 823 K. SS 316LN has been chosen as the major structural material for the construction of Prototype Fast Breeder Reactor (PFBR) with a targeted service life of 40 years. To reduce the risk of sensitization in SS 316LN, the carbon content has been reduced to less than 0.03 wt%, and the nitrogen content has been specified as 0.08 wt% to compensate the loss in strength due to the reduced carbon content. An improved version of this alloy with nitrogen content of 0.14 wt% in a frilly austenite matrix has been developed for the future FBRs, to enhance the service life of the structural components up to 60 years. Indigenously developed modified E3 16-1 5 electrodes were used for the fabrication of PFBR components to enhance the structural reliability of the components. The modifications from AWS/ASME SFA 5.4 include stringent composition limits, narrow range of ferrite content, and impact toughness after aging at 1023K for 100h, tensile properties at elevated (service) temperatures and intergranular corrosion (IGC) test as per ASTM A262 Practice E. Since the improved version alloy is rich in nitrogen content than the existing alloy, it has become necessary to develop a welding consumable with reasonably good weldability that is suitable for the fabrication of future FBR components. At present there are no commercially available welding consumables to weld these steels and the development is under way. In this work, a matching consumable methodology was adopted to develop the welding consumable. However, as per specification targeting the chemistry, solidification mode and delta ferrite was challenging, since the solidification mode of the weld metal shifts to fully austenitic region due to dilution of nitrogen from the base metal, which may increase the risk of hot cracking susceptibility