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Sample records for ceramic breeder pebble

  1. Numerical simulation of ceramic breeder pebble bed thermal creep behavior

    International Nuclear Information System (INIS)

    The evolution of ceramic breeder pebble bed thermal creep deformation subjected to an external load and a differential thermal stress was studied using a modified discrete numerical code previously developed for the pebble bed thermomechanical evaluation. The rate change of creep deformation was modeled at the particle contact based on a diffusion creep mechanism. Numerical results of strain histories have compared reasonably well with those of experimentally observed data at 740 C using activation energy of 180 KJ/mole. Calculations also show that, at this activation energy level, a particle bed at an elevated temperature of 800 C may cause undesired local sintering at a later time when it is subjected to an external load of 6.3 MPa. Thus, by tracking the stress histories inside a breeder pebble bed the numerical simulation provides an indication of whether the bed may encounter an undesired condition under a typical operating condition. (orig.)

  2. Numerical simulation of ceramic breeder pebble bed thermal creep behavior

    International Nuclear Information System (INIS)

    The evolution of ceramic breeder pebble bed thermal creep deformation subjected to an external load and a differential thermal stress was studied using a modified discrete numerical code previously developed for the pebble bed thermomechanical evaluation. The rate change of creep deformation was modeled at the particle contact based on a diffusion creep mechanism. Numerical results of strain histories have shown lower values as compared to those of experimentally observed data at 740 deg. C using an activation energy of 180 kJ/mol. Calculations also show that, at this activation energy level, a particle bed at an elevated temperature of 800 deg. C may cause too much particle overlapping with a contact radius growth beyond 0.65 radius at a later time, when it is subjected to an external load of 6.3 MPa. Thus, by tracking the stress histories inside a breeder pebble bed the numerical simulation provides an indication of whether the bed may encounter an undesired condition under a typical operating condition

  3. Particle flow of ceramic breeder pebble beds in bi-axial compression experiments

    International Nuclear Information System (INIS)

    Pebble beds of Tritium breeding ceramic material are investigated within the framework of developing solid breeder blankets for future nuclear fusion power plants. For the thermo-mechanical characterisation of such pebble beds, bed compression experiments are the standard tools. New bi-axial compression experiments on 20 and 30 mm high pebble beds show pebble flow effects much more pronounced than in previous 10 mm beds. Owing to the greater bed height, conditions are reached where the bed fails in cross direction and unhindered flow of the pebbles occurs. The paper presents measurements for the orthosilicate and metatitanate breeder materials that are envisaged to be used in a solid breeder blanket. The data are compared with calculations made with a Drucker-Prager soil model within the finite-element code ABAQUS, calibrated with data from other experiments. It is investigated empirically whether internal bed friction angles can be determined from pebble beds of the considered heights, which would simplify, and broaden the data base for, the calibration of the Drucker-Prager pebble bed models

  4. 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.)

  5. 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.

  6. 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

  7. Preparation and characterization of Li4SiO4 ceramic pebbles by graphite bed method

    International Nuclear Information System (INIS)

    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 (Li4SiO4) 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 Li4SiO4 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, Li4SiO4 pebbles with desired size and shape were prepared. The obtained Li4SiO4 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

  8. Status and perspective of the R and D on ceramic breeder materials for testing in ITER

    International Nuclear Information System (INIS)

    The main line of ceramic breeder materials research and development is based on the use of the breeder material in the form of pebble beds. At present, there are three candidate pebble materials (Li4SiO4, and two forms of Li2TiO3) for DEMO reactors that will be used for testing in ITER. This paper reviews the R and D of as-fabricated pebble materials against the blanket performance requirements and makes recommendations on necessary steps toward the qualification of these materials for testing in ITER

  9. Development of fabrication technologies for advanced tritium breeder pebbles by the sol–gel method

    International Nuclear Information System (INIS)

    Highlights: • Li2TiO3 with excess Li (Li2+xTiO3+y) was developed as an advanced tritium breeder. • Pebble fabrication by the sol–gel method is a promising technique for the mass production of advanced tritium breeder pebbles. • To increase the density of the sintered Li2+xTiO3+y pebbles, the sintering temperature was changed. • At 1353 K, the density of the pebbles increased to approximately 85% T.D. without any increase in the grain size. -- Abstract: Demonstration power plant (DEMO) reactors require advanced tritium breeders with high thermal stability. Li2TiO3 with excess Li (Li2+xTiO3+y) was developed as an advanced tritium breeder. Pebble fabrication by the sol–gel method is a promising technique for the mass production of advanced tritium breeder pebbles. I have been developing a sol–gel technique for fabricating Li2+xTiO3+y pebbles, and the next step is to optimize the granulation conditions to reach the target value. In a previous study, the average grain size on the surfaces and cross sections of sintered Li2+xTiO3+y pebbles was large whereas the theoretical density (T.D.) of these pebbles was small. To increase the density of the sintered Li2+xTiO3+y pebbles, the sintering temperature was changed, and at 1353 K, the density of the pebbles increased to approximately 85% T.D. without any increase in the grain size. This suggests that the pore size in the sintered Li2+xTiO3+y pebbles decreased because of sintering in vacuum and argon

  10. 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.)

  11. Trial examination of direct pebble fabrication for advanced tritium breeders by the emulsion method

    Energy Technology Data Exchange (ETDEWEB)

    Hoshino, Tsuyoshi, E-mail: hoshino.tsuyoshi@jaea.go.jp

    2014-10-15

    Highlights: • The integration of raw material preparation and granulation is proposed as a new direct pebble fabrication process. • The emulsion method granulates gel spheres of Li{sub 2}CO{sub 3} and TiO{sub 2} or SiO{sub 2}. • The gel spheres are calcined and sintered in air. • The crush load of the sintered Li{sub 2}TiO{sub 3} or Li{sub 4}SiO{sub 4} pebbles obtained is 37.2 or 59.3 N, respectively. - Abstract: Demonstration power plant reactors require advanced tritium breeders with high thermal stability. For the mass production of advanced tritium breeder pebbles, pebble fabrication by the emulsion method is a promising technique. To develop the most efficient pebble fabrication method, a new direct pebble fabrication process utilizing the emulsion method was implemented. A prior pebble fabrication process consisted of the preparation of raw materials followed by granulation. The new process integrates the preparation and granulation of raw materials. The slurry for the emulsion granulation of Li{sub 2}TiO{sub 3} or Li{sub 4}SiO{sub 4} as a tritium breeder consists of mixtures of Li{sub 2}CO{sub 3} and TiO{sub 2} or SiO{sub 2} at specific ratios. Subsequently, gel spheres of tritium breeders are fabricated by controlling the relative flow speeds of slurry and oil. The average diameter and crush load of the obtained sintered Li{sub 2}TiO{sub 3} or Li{sub 4}SiO{sub 4} pebbles were 1.0 or 1.5 mm and 37.2 or 59.3 N, respectively. The trial fabrication results suggest that the new process has the potential to increase the fabrication efficiency of advanced tritium breeder pebbles.

  12. Trial examination of direct pebble fabrication for advanced tritium breeders by the emulsion method

    International Nuclear Information System (INIS)

    Highlights: • The integration of raw material preparation and granulation is proposed as a new direct pebble fabrication process. • The emulsion method granulates gel spheres of Li2CO3 and TiO2 or SiO2. • The gel spheres are calcined and sintered in air. • The crush load of the sintered Li2TiO3 or Li4SiO4 pebbles obtained is 37.2 or 59.3 N, respectively. - Abstract: Demonstration power plant reactors require advanced tritium breeders with high thermal stability. For the mass production of advanced tritium breeder pebbles, pebble fabrication by the emulsion method is a promising technique. To develop the most efficient pebble fabrication method, a new direct pebble fabrication process utilizing the emulsion method was implemented. A prior pebble fabrication process consisted of the preparation of raw materials followed by granulation. The new process integrates the preparation and granulation of raw materials. The slurry for the emulsion granulation of Li2TiO3 or Li4SiO4 as a tritium breeder consists of mixtures of Li2CO3 and TiO2 or SiO2 at specific ratios. Subsequently, gel spheres of tritium breeders are fabricated by controlling the relative flow speeds of slurry and oil. The average diameter and crush load of the obtained sintered Li2TiO3 or Li4SiO4 pebbles were 1.0 or 1.5 mm and 37.2 or 59.3 N, respectively. The trial fabrication results suggest that the new process has the potential to increase the fabrication efficiency of advanced tritium breeder pebbles

  13. 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.)

  14. 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.

  15. Plutonium destruction with pebble bed type HTGRs using Pu burner balls and breeder balls

    International Nuclear Information System (INIS)

    It was made clear that pebble bed type HTGRs using Pu burner balls (pu balls) and breeder balls (Th balls) possesses a potential to burn weapons-grade Pu to 740 Gwd/TPu. The total amounts of Pu and 239Pu of can reduced to about 20 and 1%, respectively. (author). 10 refs, 4 figs, 2 tabs

  16. EXOTIC-7: irradiation of ceramic breeder materials to high lithium burnup

    International Nuclear Information System (INIS)

    The EXOTIC-7 irradiation experiment in the high flux reactor (HFR) has been completed. Its aim has been to investigate the effects of high lithium-burnup on the mechanical stability and tritium release characteristics of candidate ceramic breeder materials, originating from the fusion programmes of CEA, FZK, ENEA, AECL and ECN. The tested ceramic breeder materials were pellets of Li2ZrO3, LiAlO2 and Li8ZrO6 and pebbles of Li4SiO4 and Li2ZrO3, with a variety of characteristics, like grain size and porosity. The test matrix provided the simultaneous irradiation of eight independent capsules with on-line tritium monitoring. Two capsules contained a mixture of Li4SiO4 and beryllium pebbles. The experimental design, sample loading and main irradiation parameters are described. Some PIE results and analysis of in-situ tritium release behaviour are presented. (orig.)

  17. 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.)

  18. Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Highlights: • This work proposes three possible designs for a thorium Pebble Bed Reactor. • A high-conversion PBR (CR > 0.96), passively safe and within practical constraints. • A thorium breeder PBR (220 cm core) in practical regime, but not passively safe. • A passively safe breeder, requiring higher fuel reprocessing and recycling rates. - Abstract: More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3 g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30 g thorium pebbles, for enhanced conversion. The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600 °C. Results show that high conversion ratios (CR > 0.96) and passive safety can be combined in

  19. Tritium recovery from ceramic breeder blanket

    International Nuclear Information System (INIS)

    It is known that chemical forms of tritium released from ceramic breeders are T2O and T2. Among issues relevant to the tritium chemical form, tritium inventory is one of the major criteria in the selection of breeder material. The primary purpose of this report is to study the dependence of tritium inventory in a blanket with ceramic solid breeder on the tritium chemical form. In this light, tritium inventory in a Li2O blanket has been evaluated as a function of tritium chemical form under the conditions of the Japanese Fusion Experimental Reactor (FER). It was shown that in a blanket with Li2O as a breeder, which has a strong affinity to water vapor, the inventory due to T2O adsorption becomes quite large. In order to reduce the T2O adsorption inventory, conversion of the tritium chemical form through an isotope exchange reaction with hydrogen added to the sweep gas (T2O + 2 H2 → H2O + 2 HT) has been proposed, and its advantages and problems have been examined. Lithium hydroxide formation and mass transfer, which are considered to be inherent in the Li2O-T2O system and to be critical issues for the feasibility of a Li2O blanket, have been also discussed. (author)

  20. Lithium reprocessing technology for ceramic breeders

    Science.gov (United States)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Saito, Minoru; Tatenuma, Katuyashi; Kainose, Mitsuru

    1995-03-01

    Lithium ceramics have been receiving considerable attention as tritium breeding materials for fusion reactors. Reprocessing technology development for these materials is proposed to recover lithium, as an effective use of resources and to remove radioactive isotopes. Four potential ceramic breeders (Li 2O, LiAlO 2, Li 2ZrO 3 and Li 4SiO 4) were prepared in order to estimate their dissolution properties in water and various acids (HCl, HNO 3, H 2SO 4, HF and aqua regia). The dissolution rates were determined by comparing the weight of the residue with that of the starting powder (the weight method). Recovery properties of lithium were examined by the precipitation method.

  1. Corrosion susceptibility of EUROFER97 in lithium ceramics breeders

    International Nuclear Information System (INIS)

    EUROFER97 specimens were exposed in vacuum to lithium silicate pebbles at 550 °C for up to 2880 h, to evaluate its corrosion susceptibility in a simulated breeder blanket environment. The specimens and pebble bed were then analyzed and characterized by SEM-EDX, XRD, and HR-TEM. The results revealed the formation of a double chromium/iron oxide corrosion layer. HR-TEM also showed that the inner layer was amorphous, while the outer was crystalline. The amorphous layer was brittle, broke easily, and became detached from the steel

  2. 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

  3. 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.

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

    International Nuclear Information System (INIS)

    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 Li2TiO3 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

  5. EXOTIC-7: Irradiation of ceramic breeder materials to high lithium burnup

    International Nuclear Information System (INIS)

    The EXOTIC-7 irradiation experiment in the High Flux Reactor (HFR) at Petten has been completed. Its aim has been to investigate the effects of high lithium-burnup on the mechanical stability and tritium release characterisitcs of candidate ceramic breeder materials, originating from the Fusion Programmes of CEA, FZK, ENEA, AECL and ECN. The tested ceramic breeder materials were pellets of Li2ZrO3, LiAlO2 and Li8ZrO6 and pebbles of Li4SiO4 and Li2ZrO3, with a variety of characteristics, like grain size and porostiy. The test matrix provided the simultaneous irradiation of eight independent capsules with on-line tritium monitoring. Two capsules containd a mixture of Li4SiO4 and beryllium pebbles. The experimental design, sample loading and main riiadiation parameters are described. Some PIE results and analysis of in-situ tritium release behaviour are presented. (orig.)

  6. Reactivity control system of a passively safe thorium breeder pebble bed reactor

    International Nuclear Information System (INIS)

    Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF3 gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MWth passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel, between 100 cm and

  7. Disposition of weapon-grade plutonium with pebble bed type HTGRs using Pu burner balls and Th breeder balls

    International Nuclear Information System (INIS)

    A concept of reactor system was developed with which weapons-grade plutonium could be made perfectly worthless in use for weapons. It is a pebble bed type HTGR using Pu burner ball fuels and Th breeder ball fuels. The residual amounts of 239Pu in spent Pu balls become less than 1% of the initial loading. Furthermore, a method was found that the power coefficient could be made negative by heavy Pu loading in the Pu burner ball fuels

  8. Neutron irradiation of candidate ceramic breeder materials of fusion reactors

    International Nuclear Information System (INIS)

    In the context of the European programs for the future fusion reactors, the Process Chemistry Department of ENEA, Casaccia Center (Rome), has been involved in preparing ceramic blanket materials as tritium breeders; a special consideration has been addressed to the nuclear characterization of LiAlO2 and Li2ZrO3. In this paper are reported neutron irradiation of ceramic specimens in TRIGA reactor and γ-spectrometric measurements for INAA purposes; and isothermal annealing of the irradiated samples and tritium extraction, by using an 'out of pile' system. (author) 3 refs.; 4 figs.; 4 tabs

  9. Analysis of the running-in phase of a Passively Safe Thorium Breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Highlights: • This work analyzes important trends of the running-in phase of a thorium breeder PBR. • Depletion equations are solved for important actinides and a fission product pair. • Breeding U-233 is achieved in 7 years by cleverly adjusting the feed fuel enrichment. • A safety analysis shows the thorium PBR is passively safe during the running-in phase. - Abstract: The present work investigates the running-in phase of a 100 MWth Passively Safe Thorium Breeder Pebble Bed Reactor (PBR), a conceptual design introduced in previous equilibrium core design studies by the authors. Since U-233 is not available in nature, an alternative fuel, e.g. U-235/U-238, is required to start such a reactor. This work investigates how long it takes to converge to the equilibrium core composition and to achieve a net production of U-233, and how this can be accelerated. For this purpose, a fast and flexible calculation scheme was developed to analyze these aspects of the running-in phase. Depletion equations with an axial fuel movement term are solved in MATLAB for the most relevant actinides (Th-232, Pa-233, U-233, U-234, U-235, U-236 and U-238) and the fission products are lumped into a fission product pair. A finite difference discretization is used for the axial coordinate in combination with an implicit Euler time discretization scheme. Results show that a time dependent adjustment scheme for the enrichment (in case of U-235/U-238 start-up fuel) or U-233 weight fraction of the feed driver fuel helps to restrict excess reactivity, to improve the fuel economy and to achieve a net production of U-233 faster. After using U-235/U-238 startup fuel for 1300 days, the system starts to work as a breeder, i.e. the U-233 (and Pa-233) extraction rate exceeds the U-233 feed rate, within 7 years after start of reactor operation. The final part of the work presents a basic safety analysis, which shows that the thorium PBR fulfills the same passive safety requirements as the

  10. 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

  11. 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.)

  12. Radiolysis of Slightly Overstoichiometric Lithium Orthosilicate Pebbles

    OpenAIRE

    Zarins, A.; Supe, A; Kizane, G; Knitter, R.; Reinholds, I; Vitins, A; Tilika, V; Actins, A; Baumane, L

    2010-01-01

    : One of the technological problems of a fusion reactor is the change in composition and structure of ceramic breeder (Li4SiO4 or Li2TiO3 pebbles) during long-term operation. Changes in the composition and structure of the Li4SiO4 ceramic pebbles at fast electron irradiation (E = 5 MeV, dose rate up to 88 MGy•h-1, absorbed dose up to 10.6 GGy) at 543-573 K were investigated in this study. Overstoichiometric (2.5 weight % of additional SiO2) lithium orthosilicate pebbles were fabricated by...

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

    International Nuclear Information System (INIS)

    Highlights: • Breeding benchmark experiment on LLCB TBM in ITER was performed. • Nuclear responses measured are TPR and reaction rate of 115In(n, n′)115mIn 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 115In(n, n′)115mIn 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 6Li and 7Li in breeder layers were measured with the help of two different compositions of Li isotopes (60.69% 6Li and 7.54% 6Li) in Li2CO3. 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 6Li at one location in the breeder layer was also measured by direct online measurement of tritons from 6Li(n, t)4He reaction using silicon surface barrier detector and 6Li to triton converter. Additional verification of neutron spectra (En > 0.35 MeV) in the mock-up zones were obtained by measuring 115In(n, n′)115mIn 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 Li2CO3 pellets was 1.11 in first breeder zone and 1.09 in second breeder zone with uncertainty 8.3% at 1σ level. The experimental details

  14. 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

  15. Tritium transport and release from lithium ceramic breeder materials

    International Nuclear Information System (INIS)

    In an operating fusion reactor,, the tritium breeding blanket will reach a condition in which the tritium release rate equals the production rate. The tritium release rate must be fast enough that the tritium inventory in the blanket does not become excessive. Slow tritium release will result in a large tritium inventory, which is unacceptable from both economic and safety viewpoints As a consequence, considerable effort has been devoted to understanding the tritium release mechanism from ceramic breeders and beryllium neutron multipliers through theoretical, laboratory, and in-reactor studies. This information is being applied to the development of models for predicting tritium release for various blanket operating conditions

  16. 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

  17. Progress of R and D on the technology of In-pile irradiation and tritium In-situ extraction experiment of solid breeder pebble bed for CN HCCB in CARR

    International Nuclear Information System (INIS)

    The progress of the key technology of the In-Pile Irradiation and Tritium In-Situ Extraction (IPITISE) experiment was introduced. According to the design and requirements of the Helium cooled ceramic breeder (HCCB) tritium system, the scheme of the IPITISE experiment was established. The primary components of this apparatus included pebble bed assembly (PBA), tritium extraction system (TES), and tritium measurement system (TMS). The primary design and calculation of the structure, nuclear physics, and thermo-hydraulics of the PBA were carried out. It can be concluded that the max weight load of the PBA was 150 g above. The effective thermal neutron flux rate of the PBA would be high as 1.73 E +14 n/cm2 · s. The power of irradiation heat could be reached 18.47 W/cm3. After optimization of the design and parameter of PBA structure, the temperature of reduced activation ferrite/martensite (RAFM) steel with tritium permeation barrier (TPB) coating and Li4SiO4 could be respective controlled under 150∼550℃ and 200∼ 920℃. The in-situ tritium release behaviors would be studied by this experiment as well as the tritium permeation through the structure materials under irradiation condition or the reactor was shut down. Consequently, the irradiation performance of the key materials, the retention characteristics and release behaviors of Li4SiO4 ceramic breeder pebbles, the tritium permeation data of RAFM with TPB coating and credible evaluation of in-situ tritium extraction technology would be provided for China tritium breeder test blanket module in the CIPITISE experiment. (authors)

  18. Destruction of weapons-grade plutonium with pebble bed type HTGRs using burner balls and breeder balls

    International Nuclear Information System (INIS)

    As the method of disposing the plutonium coming from disassembled weapons, the method of burning the fuel in which the plutonium is mixed with a parent material in LWRs or the disposal by glass solidification is proposed. In the former method, it is desirable to do the reprocessing of spent fuel for effectively utilizing fission products. The latter method needs watch against the diversion of the plutonium. The authors devised the method of effectively annihilating plutonium by separating into the burner balls of plutonium and the breeder balls of a parent material, and burning those by mixing in a pebble bed type high temperature gas-cooled reactor, while continuously exchanging them. It was clarified from the aspect of nuclear characteristics that by using this method, 239Pu can be annihilated to the state of enabling the direct abandonment without reprocessing. The flow of burner balls and breeder balls in the reactor is shown, and multi-pass fuel exchange method was adopted to burn Pu in burner balls up. The rate of Pu annihilation was determined by the change of the amount of Pu for the burnup evaluated by lattice burning calculation. The maximum amount of Pu charge in one burner ball is limited by the maximum allowable power output of burner balls. (K.I.)

  19. Modeling of tritium behavior in ceramic breeder materials

    International Nuclear Information System (INIS)

    The model described in this paper considers diffusion and desorption as the rate-controlling mechanisms for tritium release from a ceramic breeder material. This model was used to predict the tritium release from samples of Li2SiO3 and LiAlO2, given the temperature history of the samples. The diffusion-desorption model did a better job of predicting the tritium release for these samples under pure helium purge gas than did a pure diffusion model using the best values for the diffusivity of these materials available. The activation energies of desorption found from the best fit of the predicted tritium release to the observed release were 105-108 kJ/mol for Li2SiO3 and 85.7 kJ/mol for LiAlO2. These values are in fair agreement with activation energies reported in the literature. 13 refs., 6 figs

  20. Progress in the development of Li{sub 2}ZrO{sub 3} and Li{sub 2}TiO{sub 3} pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Lulewicz, J.D.; Roux, N. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France)

    1998-03-01

    Li{sub 2}ZrO{sub 3} and Li{sub 2}TiO{sub 3} pebbles are being developed as ceramic breeder for the European Helium-cooled pebble bed DEMO blanket concept. Status is given of the fabrication work, and of the properties characteristics determination. (author)

  1. R and D activities on helium cooled solid breeder TBM in Korea

    International Nuclear Information System (INIS)

    R and D activities currently being undertaken for HCSB TBM include joining technologies of structural material, breeder and reflector pebble material development, the effect of TBM ferritic-martensitic steel on the ripple of toroidal magnetic field, and ceramic coating on graphite pebble. The HIP joining performance of FM steel is evaluated. Lithium ceramic breeder and graphite reflector pebble fabrication methods are under development using special fabrication process, and the initial characteristics of the pebbles are assessed. Silicon carbide coating on graphite pebble is also investigated and its preliminary results are mentioned. Finally, an accurate evaluation of the effect of TBM and ferromagnetic inserts on magnetic field are implemented. The current results of these R and D issues are addressed in this paper.

  2. Pebble bed packing in prismatic containers

    International Nuclear Information System (INIS)

    Highlights: • The essential part of ceramic breeder blankets is pebble beds. • The packing factor for blanket relevant cavities must be known. • Tomography experiments revealed details of packing arrangements. • Packing experiments confirm that reference packing factors will be achieved. -- Abstract: New analyses of previous tomography investigations show in detail void fraction fluctuations close to walls generated by regular pebble arrangements. Local packing factors within the pebble bed were determined for characteristic zones. These results are very helpful for the interpretation of the packing experiments performed with spherical pebbles in different kinds of Plexiglas containers dominated by flat walls. The packing factors for single-size pebbles in the containers with a piston on top are fairly independent of bed height unless the height to diameter ratio becomes less than 10. For the closed rectangular containers, the development of structured packings is rendered more difficult. However, for blanket relevant bed heights, both for orthosilicate and beryllium pebbles, the packing factors obtained which agree well with previously obtained reference values

  3. 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)

  4. Thermodynamics of ceramic breeder materials for fusion reactors

    International Nuclear Information System (INIS)

    Based on known or deduced phase relationships in ternary lithium oxygen systems such as Li-Al-O, Li-Si-O and Li-Zr-O, the unknown free enthalpy of formation values of ternary compounds are calculated starting from the known data of the compounds of the binary border systems. Criterion for the data assessment is interconsistency of the data of all the compounds within a given multi-component system. With the help of these data the development of partial pressures during the breeding process can be calculated for all the compounds of interest. In order to facilitate a compatibility assessment the quaternary systems Cr-Li-Si-O, Fe-Li-Si-O and Be-Li-Si-O were also investigated and thermodynamic data of pertinent ternary and quaternary compounds determined. Both the partial pressure development and the compatibility behaviour of a lithium containing compound are criteria for its qualification as a breeder material for a fusion reactor. (orig.)

  5. 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 breeder-inside-tube (BIT) blanket concept, the tritium leakage into the coolant was evaluated and the consequent tritium losses into the steam circuit were determined. The 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. (author) 8 refs.; 2 figs

  6. Pressurizing Behavior on Ingress of Coolant into Pebble Bed of Blanket of Fusion DEMO Reactor

    International Nuclear Information System (INIS)

    Solid breeder blankets are being developed as candidate blankets for the Fusion DEMO reactor in Japan. JAEA is performing the development of the water cooled and helium cooled solid breeder blankets. The blanket utilizes ceramic breeder pebbles and multiplier pebbles beds cooled by high pressure water or high pressure helium in the cooling tubes placed in the blanket box structure. In the development of the blanket, it is very important to incorporate the safety technology as well as the performance improvement on tritium production and energy conversion. In the safety design and technology, coolant ingress in the blanket box structure is one of the most important events as the initiators. Especially the thermal hydraulics in the pebble bed in the case of the high pressure coolant ingress is very important to evaluate the pressure propagation and coolant flow behavior. This paper presents the preliminary results of the pressure loss characteristics by the coolant ingress in the pebble bed. Experiments have been performed by using alumina pebble bed (4 litter maximum volume of the pebble bed) and nitrogen gas to simulate the helium coolant ingress into breeder and multiplier pebble beds. Reservoir tank of 10 liter is filled with 1.0 MPa nitrogen. The nitrogen gas is released at the bottom part of the alumina pebble bed whose upper part is open to the atmosphere. The pressure change in the pebble bed is measured to identify the pressure loss. The measured values are compared with the predicted values by Ergun's equation, which is the correlation equation on pressure loss of the flow through porous medium. By the results of the experiments with no constraint on the alumina pebble bed, it was clarified that the measured value agreed in the lower flow rate. However, in the higher flow rate where the pressure loss is high, the measured value is about half of the predicted value. The differences between the measured values and the predicted values will be discussed from

  7. Reprocessing of lithium titanate pebbles by graphite bed method

    International Nuclear Information System (INIS)

    Lithium titanate enriched by 6Li isotope is considered as a candidate of tritium breeding materials for fusion reactors due to its excellent performance. The reuse of burned Li2TiO3 pebbles is an important issue because of the high costs of 6Li-enriched materials and waste considerations. For this purpose, reprocessing of Li2TiO3 pebbles by graphite bed method was developed. Simulative Li2TiO3 pebbles with low-lithium content according to the expected lithium burn-up were fabricated. After that, Li2TiO3 pebbles were re-fabricated with lithium carbonate as lithium additives, in order to gain the composition of lithium titanate with a Li/Ti ratio of 2. The process was optimized to obtain reprocessed Li2TiO3 pebbles that were suitable for reuse as ceramic breeder. Density, porosity, grain size and crushing load of the reprocessed pebbles were characterized. This process did not deteriorate the properties of the reprocessed pebbles and was almost no waste generation

  8. Status of advanced tritium breeder development for DEMO in the broader approach activities in Japan

    International Nuclear Information System (INIS)

    DEMO reactors require '6Li-enriched ceramic tritium breeders' which have high tritium breeding ratios (TBRs) in the blanket designs of both EU and JA. Both parties have been promoting the development of fabrication technologies of Li2TiO3 pebbles and of Li4SiO4 pebbles including the reprocessing. However, the fabrication techniques of tritium breeders pebbles have not been established for large quantities. Therefore, these parties launch a collaborative project on scaleable and reliable production routes of advanced tritium breeders. In addition, this project aims to develop fabrication techniques allowing effective reprocessing of 6Li. The development of the production and 6Li reprocessing techniques includes preliminary fabrication tests of breeder pebbles, reprocessing of lithium, and suitable out-of-pile characterizations. The R and D on the fabrication technologies of the advanced tritium breeders and the characterization of developed materials has been started between the EU and Japan in the DEMO R and D of the International Fusion Energy Research Centre (IFERC) project as a part of the Broader Approach activities from 2007 to 2016. The equipment for production of advanced breeder pebbles is planned will be installed in the DEMO R and D building at Rokkasho, Japan. The design work in this facility was carried out. The specifications of the pebble production apparatuses and related equipment in this facility were fixed, and the basic data of these apparatuses was obtained. In this design work, the preliminary investigations of the dissolution and purification process of tritium breeders were carried out. From the results of the preliminary investigations, lithium resources of 90% above were recovered by the aqueous dissolving methods using HNO3 and H2O2. The removal efficiency of 60Co by the addition in the dissolved solutions of lithium ceramics were 97-99.9% above using activated carbon impregnated with 8-hydroxyquinolinol. In this report, preparation status

  9. Fabrication of porous LiAlO2 ceramic breeder material

    International Nuclear Information System (INIS)

    The gamma-LiAlO2 ceramic material is the reference candidate for the solid breeder option of the Next European Torus Program. The experiments and methodologies developed in Italy to produce high surface area gamma-LiAlO2 powders to be compacted by cold pressing and sintering at 70 to 90% of the theoretical density, keeping a near fully open porosity is presented. The lithiating step was assessed for the Li2CO3 and Li2O2 precursors reacting with Al2O3 having submicron grain size. Sol-gel methodologies were also developed for the gamma-LiAlO2 preparation by which very high surface area ceramic grade powders were obtained

  10. 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.)

  11. 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.)

  12. 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)

  13. 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)

  14. 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.

  15. Modelling of thermal and mechanical behaviour of pebble beds

    International Nuclear Information System (INIS)

    FZK (Forshungzentrum Karlsruhe) is developing a Helium Cooled Pebble Bed (HCPB) Blanket Concept for fusion power reactors based on the use of ceramic breeder materials and beryllium multiplier in the form of pebble beds. The design of such a blanket requires models and computer codes describing the thermal-mechanical behavior of pebble beds to evaluate the temperatures, stresses, deformations and mechanical interactions between pebble beds and the structure with required accuracy and reliability. The objective to describe the beginning of life condition for the HCPB blanket seems near to be reached. Mechanical models that describe the thermo-mechanical behavior of granular materials used in form of pebble beds are implemented in a commercial structure code. These models have been calibrated using the results of a large series of dedicated experiments. The modeling work is practically concluded for ceramic breeder; it will be carried on in the next year for beryllium to obtain the required correlations for creep and the thermal conductivity. The difficulties for application in large components (such as the HCPB blanket) are the limitations of the present commercial codes to manage such a set of constitutive equations under complex load conditions and large mesh number. The further objective is to model the thermal cycles during operation; the present correlations have to be adapted for the release phase. A complete description of the blanket behavior during irradiation is at the present out of our capability; this objective requires an extensive R and D program that at the present is only at the beginning. (Y.Tanaka)

  16. 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)

  17. Analysis of the in-pile operation and preliminary results of the post irradiation dismantling of the pebble bed assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Magielsen, A.J.; Peeters, M.M.W.; Hegeman, J.B.J.; Stijkel, M.P.; Laan, J.G. van der [NRG-Nuclear Research and Consultancy Group (Netherlands)

    2007-07-01

    The Pebble Bed Assemblies (PBA) are four tritium breeding sub scale modules, representing a segment of the European Helium Cooled Pebble Bed Test Blanket. The objective of these experiments is the study the thermomechanical behaviour of the pebble bed assemblies during irradiation. This objective will be full- filled by the analysis of changes in the in-pile temperature profiles during irradiation and the post irradiation examination of the pebble beds in the Hot Cells. The PBA has been irradiated in the HFR in Petten for 294 Full Power Days (FPD), to a dose of 2-3 dpa in Eurofer, and estimated lithium burnup of 2-3 %. Changes in the temperature profile during in-pile operation are indication for pebble bed creep compaction during first start up and the possible formation gas gaps between the pebble beds and the structure. During progressive irradiation the radial and axial differential temperatures within the breeder and beryllium pebble beds are evaluated. During start up of the sub sequent irradiation cycles (each 26 FPD) the temperature differences within the beryllium pebble beds show a slight increase suggesting changes in the structure of the pebble beds. The PBA are transported from the HFR to the Hot Cell Laboratory in upright position to maintain the gas gaps between the pebble beds and Eurofer. Various microscopy preparation techniques are used to study the deformation state of the pebble beds (signs of creep compaction and sintering), formation of gas gaps between the pebble beds and structural materials and the interaction layers between eurofer-ceramic and eurofer-beryllium. In this paper first results on the Post Irradiation examination are given. (orig.)

  18. Tritium percolation through porous ceramic breeders - a random-lattice approach

    International Nuclear Information System (INIS)

    Among the major processes leading to tritium transport through Li ceramic breeders the percolation of gaseous tritium species through the connected porosity remains the least amenable to a satisfactory treatment. The combination of diffusion and reaction through the convoluted transport pathways prescribed by the system of pores poses a formidable challenge. The key issue is to make the fundamental connection between the tortuousity of the medium with the transport processes in terms of only basic parameters (e.g., molecular diffusion coefficient and porosity distribution) that are amenable to fundamental understanding and experimental determinations. This fundamental challenge is met within the following approaches. On the microscale the short range transport is modeled via a connection-diffusion-reaction approach. On a macro scale the long range transport is described within a matrix formalism. The convoluted microstructure of the pore system as prescribed from experimental measurements is synthesized into the present approach via Monte Carlo simulation techniques. In this way the approach requires as inputs only physical-chemical parameters that are amenable to clear basic understanding and experimental determination. In this sense it provides predictive capability. Using this approach the concept of residence time has been analyzed in a critical manner. Implication for tritium release experiments was discussed. (orig.)

  19. Tritium permeation through helium-heated steam generators of ceramic breeder blankets for DEMO

    International Nuclear Information System (INIS)

    The potential sources of tritium contamination of the helium coolant of ceramic breeder blankets have previously been evaluated for the specific case of the European BIT DEMO blanket. This confirmed that the control of tritium losses to the steam circuit is a critical issue which demands development concerning (a) permeation barriers, (b) tritium recovery processes maintaining a very low tritium activity in the coolant, and (c) control of the coolant chemistry. The specifications of these developments required the evaluation of the tritium losses through the steam generators, and includes the definition of their operating conditions by thermodynamic cycle calculations, and their thermal-hydraulic design. For both tasks, specific computer tools were developed. The geometry obtained, the surface area and the temperature profiles along the heat-exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam-oxidized Incoloy 800 austenitic stainless steel was identified as the best-suited existing material. Our results indicate that in nominal steady-state operation the tritium escape into the steam cycle could be restricted to less than 10 Ci per day. The conditions for this are specified, but their feasibility demands, in particular, the resolution of certain gas chemistry problems, and their validation in the more stringent environment of an operating blanket. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) was identified as bearing a large tritium release potential. The problems associated with such transients are discussed and possible solutions are proposed. (orig.)

  20. Experimental study and analysis of the purge gas pressure drop across the pebble beds for the fusion HCPB blanket

    International Nuclear Information System (INIS)

    Highlights: ► The pressure drop significantly increases with decreasing the pebbles diameter. ► The pressure drop slightly increases with increasing the packing factor. ► The pressure drop is directly proportional to pebble bed length and inlet pressure. ► Predictions of Ergun equation agree well with the measured values of pressure drop. ► The filters resistance has a small contribution to the total pressure drop. -- Abstract: The lithium ceramic and beryllium pebble beds of the breeder units (BU), in the fusion breeding blanket, are purged by helium to extract the bred tritium. Therefore, the objective of this study is to support the design of the BU purge gas system by studying the effect of pebbles diameter, packing factor, pebble bed length, and flow inlet pressure on the purge gas pressure drop. The pebble bed was formed by packing glass pebbles in a rectangular container (56 mm × 206 mm × 396 mm) and was integrated into a gas loop to be purged by helium at BU-relevant pressures (1.1–3.8 bar). To determine the pressure drop across the pebble bed, the static pressure was measured at four locations along the pebble bed as well as at the inlet and outlet locations. The results show: (i) the pressure drop significantly increases with decreasing the pebbles diameter and slightly increases with increasing the packing factor, (ii) for a constant inlet flow velocity, the pressure drop is directly proportional to the pebble bed length and inlet pressure, and (iii) predictions of Ergun's equation agree well with the experimental values of the pressure drop

  1. Production of nuclear fusion reactor fuel by ceramic tritium breeder material

    International Nuclear Information System (INIS)

    Fuel tritium is generated from the nuclear reaction between the fusion neutron and the lithium of the breeder material arranged in the blanket that encloses the fusion plasma in the fusion reactor. However, the release process of the generated tritium has not been completely clarified. Recently, Japan Atomic Energy Agency started the tritium generation and recovery experiment in using nuclear fusion neutron source (FNS). In this report, the recent results of study on breeder material and its manufacturing technology is presented. (author)

  2. Drucker-Prager-Cap creep modelling of pebble beds in fusion blankets

    International Nuclear Information System (INIS)

    Modelling of thermal and mechanical behaviour of pebble beds for fusion blankets is an important issue to understand the interaction of solid breeder and beryllium pebble beds with the surrounding structural material. Especially the differing coefficients of thermal expansion of these materials cause high stresses and strains during irradiation induced volumetric heating. To describe this process, the coupled thermomechanical behaviour of both pebble bed materials has to be modelled. Additionally, creep has to be considered contributing to bed deformations and stress relaxation. Motivated by experiments, we use a continuum mechanical approach called Drucker-Prager/Cap theory to model the macroscopic pebble bed behaviour. The model accounts for pressure dependent shear failure, inelastic hardening, and volumetric creep. The elastic part is described by a nonlinear elasticity law. The model has been implemented by user-defined routines in the commercial finite-element code ABAQUS. To check the numerics, the implementation is compared to an analytical solution. Furthermore, the Drucker-Prager/Cap tool is applied to a single ceramic breeder bed subject to creep under volumetric heating

  3. Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

    Science.gov (United States)

    Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

    2015-09-01

    The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  4. Crystal chemistry of immobilization of Fast Breeder Reactor (FBR) simulated waste in Sodium Zirconium Phosphate (NZP) based ceramic matrix

    International Nuclear Information System (INIS)

    Full text: Sodium zirconium phosphate (hereafter NZP) is a potential material for immobilization of long lived heat generating radio nuclides. Possibility for the incorporation of simulated waste of fast breeder reactor origin in NZP was examined. It was found that most of the elements could be immobilized in this ceramic matrix without significant changes of the three-dimensional framework of the host material. All simulated 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% waste loaded NZP waste forms have been investigated using General Structure Analysis System (GSAS) programming of the step analysis powder diffraction data of the waste forms. Rietveld refinement of crystal data on the WOx loaded waste forms (NZPI-NZPVII) gives a satisfactory convergence of R-factors. The particle size along prominent reflecting planes calculated by Scherrer's formula varies between 68-141nm. The polyhedral distortions and effective valence calculations from bond strength data are also reported. Morphological examination by SEM reveals that the size of almost rectangular parallelepiped shaped grains varies between 0.2 and 5 μm. The EDX analysis provides analytical evidence of immobilization of effluent cations in the matrix

  5. Thermal behaviour and tritium management for in-pile testing of the pebble bed assemblies in the HFR in Petten

    International Nuclear Information System (INIS)

    Four pebble-bed assemblies are to be irradiated in the HFR in Petten with the objective to study the thermo-mechanical behaviour of the breeder ceramic pebble beds during irradiation. The thermo-mechanical behaviour of the pebble bed assemblies was calculated in a 2D axi-symmetric model in MARC. In this approach there could not be accounted for the influence of thermocouple tubes on the temperature distribution in the assembly, because these are distributed in the assembly in a non axi-symmetric manner. The solution for this problem was to expand the model to a 3D model used for thermal computations only. For safety reasons the tritium production in the breeder and permeation through the first and second containment must be estimated before the in-pile experimentation begins. In order to do so, the calculated thermal distribution is used as input for the enhanced two-dimensional finite element model in MARC. Adaptations are made in the 2D model by adding the capability of performing mass flux calculations. This paper describes the finite element models used for computation of the temperature distribution and the tritium flux through the pebble bed assembly. The results of these calculations are critical for a safety assessment of the in-pile operation of the experiment and will give a better understanding of the in-pile behaviour on temperature and tritium management in advance. (orig.)

  6. Current status of safety design and safety analysis for China ITER helium coolant ceramic breeder test blanket system long

    International Nuclear Information System (INIS)

    Helium Coolant Ceramic Breeder (HCCB) Test Blanket System (TBS) designed by China are planned to be tested in ITER to validate key technologies, including demonstration of nuclear safety, for future fusion reactor breeding blankets. Furthermore, in order to be operated in ITER, a nuclear facility (INB) recognized by French nuclear safety authority, safety design and safety analysis of the TBS are mandatory for the licensing procedures. This paper summarizes the status at current design phase with following main elements: The main radiological source terms in the system are tritium and activation products. Nuclear and tritium analysis are performed to identify their inventories and distributions in system. Multiple confinement barriers are considered to be the most essential safety feature. French regulation for pressure equipment and nuclear equipment (ESP/ESPN regulations) will be followed to ensure the system integrities. ALARA principle is kept in mind during the whole safety design phases. Protective actions including choice of advanced materials, improvement of shielding, optimization of operation and maintenance activities, usage of remote handling operations, zoning and access control have been considered. Passive safety is emphasized in the system design, only minimal active safety functions including call for fusion plasma shutdown and isolation of TBM from ex-vessel ancillary systems. High reliability and redundancies are required for components related to these functions. Several accidents have been identified and analyzed. Consider the limited inventories in the system and the intrinsic safety of fusion device, positive conclusions have been obtained. (author)

  7. 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.

  8. Behavior of beryllium pebbles under irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Dalle-Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik; Baldwin, D.L.; Gelles, D.S.; Greenwood, L.R.; Kawamura, H.; Oliver, B.M.

    1998-01-01

    Beryllium pebbles are being considered in fusion reactor blanket designs as neutron multiplier. An example is the European `Helium Cooled Pebble Bed Blanket.` Several forms of beryllium pebbles are commercially available but little is known about these forms in response to fast neutron irradiation. Commercially available beryllium pebbles have been irradiated to approximately 1.3 x 10{sup 22} n/cm{sup 2} (E>1 MeV) at 390degC. Pebbles 1-mm in diameter manufactured by Brush Wellman, USA and by Nippon Gaishi Company, Japan, and 3-mm pebbles manufactured by Brush Wellman were included. All were irradiated in the below-core area of the Experimental Breeder Reactor-II in Idaho Falls, USA, in molybdenum alloy capsules containing helium. Post-irradiation results are presented on density change measurements, tritium release by assay, stepped-temperature anneal, and thermal ramp desorption tests, and helium release by assay and stepped-temperature anneal measurements, for Be pebbles from two manufacturing methods, and with two specimen diameters. The experimental results on density change and tritium and helium release are compared with the predictions of the code ANFIBE. (author)

  9. Pebble bed pebble motion: Simulation and Applications

    CERN Document Server

    Cogliati, Joshua J

    2011-01-01

    This dissertation presents a method for simulation of motion of the pebbles in a PBR. A new mechanical motion simulator, PEBBLES, efficiently simulates the key elements of motion of the pebbles in a PBR. This model simulates gravitational force and contact forces including kinetic and true static friction. It's used for a variety of tasks including simulation of the effect of earthquakes on a PBR, calculation of packing fractions, Dancoff factors, pebble wear and the pebble force on the walls. The simulator includes a new differential static friction model for the varied geometries of PBRs. A new static friction benchmark was devised via analytically solving the mechanics equations to determine the minimum pebble-to-pebble friction and pebble-to-surface friction for a five pebble pyramid. This pyramid check as well as a comparison to the Janssen formula was used to test the new static friction equations. Because larger pebble bed simulations involve hundreds of thousands of pebbles and long periods of time, P...

  10. 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.

  11. 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

  12. 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

  13. 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

  14. Conceptual design of pebble drop divertor

    International Nuclear Information System (INIS)

    A pebble drop divertor concept is proposed for future fusion reactor. The marked feature of this system is the use of multi-layer pebbles that consists of a central kernel and some coating layers, as a divertor surface component. By using multi-layer pebbles, pebble drop divertor have the advantages such as steady state wall pumping with low bulk tritium retention. The performance of whole divertor system depends on the characteristics of the multi-layer pebble. Particularly the maximum heat load of the system is determined by the dimensions, the layer structure and the material of a kernel. A kernel also has an important role to determine surface temperature, which affects the wall pumping efficiency. This paper presents the numerical results of the maximum allowable heat load and the surface temperature of the divertor pebble. From the numerical estimation of thermal stress and surface temperature, it is found that the radius of divertor pebble with ceramic kernel should be 0.5 - 1 mm. (author)

  15. Conceptual design of pebble drop divertor

    International Nuclear Information System (INIS)

    A pebble drop divertor concept is proposed for future fusion reactor. The marked feature of this system is the use of multi-layer pebbles that consists of a central kernel and some coating layers, as a divertor surface component. By using multi-layer pebbles, pebble drop divertor have the advantages such as steady state wall pumping with low bulk tritium retention. The performance of whole divertor system depends on the characteristics of the multi-layer pebble. Particularly the maximum heat load of the system is determined by the dimensions, the layer structure and the material of a kernel. A kernel also has an important role to determine surface temperature, which affects the wall pumping efficiency. This paper presents the numerical results of the maximum allowable heat load and the surface temperature of the divertor pebble. From the numerical estimation of thermal stress and surface temperature, it is found that the radius of divertor pebble with ceramic kernel should be 0.5-1 mm. (author)

  16. Design, synthesis and characterization of the advanced tritium breeder: Li4+xSi1-xAlxO4 ceramics

    Science.gov (United States)

    Zhao, Linjie; Long, Xinggui; Chen, Xiaojun; Xiao, Chengjian; Gong, Yu; Guan, Qiushi; Li, Jiamao; Xie, Lei; Chen, Xiping; Peng, Shuming

    2015-12-01

    Li4+xSi1-xAlxO4 solid solutions which were designed as the advanced tritium breeder were obtained by solid state reactions. Samples were systematically characterized by various techniques. XRD, neutron diffraction and Raman results showed that the Aluminum substituted silicon into the Li4SiO4 lattice and Li+ interstitials formed as a result of charge compensation. Rietveld refinements of neutron diffraction showed that the crystalline structure had been expanded as Al-doped. Moreover, the lithium atom density, thermal conductivity and the mechanical property of the Li4+xSi1-xAlxO4 ceramics were improved relative to the Li4SiO4.

  17. Neutronics R and D efforts in support of the European breeder blanket development programme

    International Nuclear Information System (INIS)

    The EU fusion technology programme considers two blanket development lines, the Helium-Cooled Pebble Bed (HCPB) blanket with Lithium ceramics pebbles as breeder material and beryllium pebbles as neutron multiplier, and the Helium-Cooled Lithium-Lead (HCLL) blanket with the Pb-Li eutectic alloy acting both as breeder and neutron multiplier. The long-term strategy aims at providing validated engineering designs of breeder blankets for a fusion power demonstration reactor (DEMO). As an important intermediate step, the breeder blankets need to be tested in a real fusion environment as provided by ITER. HCPB and HCLL Test Blanket Modules (TBM) have been accordingly designed for tests in dedicated ITER blanket ports. The nuclear design and performance of the breeder blanket modules rely on the results provided by neutronics design calculations. Validated computational tools and qualified nuclear data are required for high prediction accuracies including reliable uncertainty assessments. Complementary to the application of established standard tools and data for design analysis, a dedicated neutronics R and D effort is therefore conducted in the EU. This includes the development of dedicated computational tools, the generation of high quality nuclear data and their validation through integral experiments. The recent neutronic design efforts have been devoted to the European DEMO reactor study comprising (i) Monte Carlo based pre-analysis for the dimensioning of the shielding system, (ii) the generation of a generic CAD based Monte Carlo geometry model, and (iii) performance analysis for HCLL and HCPB based DEMO variants. The recent focus of the validation effort is on neutronics TBM mock-up experiments. The first experiment of this kind was performed on a TBM mock-up of the HCPB breeder blanket. The follow-up experiment on a neutronics HCLL TBM mock-up is currently under preparation. Computational pre-analysis were performed to optimise the design of the mock

  18. PEBBLES Mechanics Simulation Speedup

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2010-05-01

    Pebble bed reactors contain large numbers of spherical fuel elements arranged randomly. Determining the motion and location of these fuel elements is required for calculating certain parameters of pebble bed reactor operation. These simulations involve hundreds of thousands of pebbles and involve determining the entire core motion as pebbles are recirculated. Single processor algorithms for this are insufficient since they would take decades to centuries of wall-clock time. This paper describes the process of parallelizing and speeding up the PEBBLES pebble mechanics simulation code. Both shared memory programming with the Open Multi-Processing API and distributed memory programming with the Message Passing Interface API are used in simultaneously in this process. A new shared memory lock-less linear time collision detection algorithm is described. This method allows faster detection of pebbles in contact than generic methods. These combine to make full recirculations on AVR sized reactors possible in months of wall clock time.

  19. Overview of Helium Cooled Ceramic Reflector Test Blanket Module development in Korea

    International Nuclear Information System (INIS)

    Korea plans to install and test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in the ITER, because the HCCR blanket concept is one of options of the DEMO blanket. Currently, many design and R and D activities have been performed to develop the Korean HCCR TBM. An integrated design tool for a fusion breeder blanket has been developed based on nuclear technologies including a safety analysis for obtaining a license for testing in the ITER. A half-scale sub-module mockup of the first wall with the manifold was fabricated, and the manufacturability and thermo-hydraulic performances were evaluated. High heat load and helium cooling test facilities have been constructed. Next, the recent status of TBM material development in Korea was introduced including Reduced Activation Ferritic Martensitic (RAFM) steel, lithium ceramic pebbles and silicon carbide (SiC) coated graphite pebbles. Several fabrication methods of RAFM steel, lithium ceramic pebbles, and silicon carbide coating on graphite pebbles were investigated. Recent design and R and D progress on these areas are introduced here

  20. Tritium release from lithium orthosilicate pebbles deposited with palladium

    International Nuclear Information System (INIS)

    Full text of publication follows: Slightly over-stoichiometric lithium orthosilicate pebbles have been selected as one optional breeder material for the European Helium Cooled Pebble Bed (HCPB) blanket. This material has been developed in collaboration of Research Center Karlsruhe and the Schott Glass, Mainz. The lithium orthosilicate pebbles are fabricated from lithium hydroxide and silica by a melting and spraying method in a semi-industrial scale facility. Lithium hydroxide was selected as the precursor since enriched lithium hydroxide is commercially available. The lithium orthosilicate pebbles produced by the process contains oxide phases besides orthosilicate, but it was also found that the oxide phases can be decomposed by annealing at high temperatures. The lithium orthosilicate pebbles produced in this way possesses satisfactory pebble characteristics. Therefore, the authors performed out-of-pile annealing tests using the lithium orthosilicate pebbles irradiated in a research reactor. Moreover, the effect of the deposition of palladium in the lithium orthosilicate pebbles on the behavior of tritium release was investigated. Palladium was deposited in the lithium orthosilicate pebbles by the incipient wet impregnation method using a solution of a palladium amino complex. The lithium orthosilicate pebbles were submitted to neutron irradiation at the Kyoto university research reactor. In the out-of-pile annealing experiments, the temperature of the breeder material placed in a tubular reactor made of quartz was raised from ambient temperature to 1173 K at a constant rate of 5 K/min under the stream of sweep gases. The tritium concentration in the outlet stream of the reactor was traced with two ionization chambers. The ionization chambers were used with a water bubbler, which enables to measure the concentrations of molecular form of tritium (HT) and tritiated water vapor (HTO) separately. In the experiments, a 0.1 % hydrogen/nitrogen sweep gas was used. The

  1. Thermal cycling tests on Li4SiO4 and beryllium pebbles

    International Nuclear Information System (INIS)

    The European B.O.T. Demo-relevant solid breeder blanket is based on the use of beds of beryllium and Li4SiO4 pebbles. Particularly dangerous for the pebble integrity are the rapid temperature changes which could occur, for instance, by a sudden blanket power shut-down. A series of thermal cycle tests have been performed for various beds of beryllium and Li4SiO4 pebbles. No breaking was observed in the beryllium pebbles, however the Li4SiO4 pebbles broke by temperature rates of change of about -50 C/sec independently on pebbles size and lithium enrichment. This value is considerably higher than the peak temperature rates of change expected in the blanket. (orig.)

  2. Lithium ceramic of blankets intend for Russian fusion reactors and an influence of the ceramic properties on parameters of reactor tritium systems

    International Nuclear Information System (INIS)

    Russian Controlled Fusion Program involves development of a DEMO design and participation in ITER Project. A solid breeder blanket in DEMO contains a ceramic orthosilicate lithium breeder and a beryllium multiplier. Test Modules of the blanket are developed in a frame of ITER activities. Experimental models of tritium breeding zones (TBZ) for the Modules, materials and technology fabrication of the TBZ, tritium reactor systems to control and treat of gases released from lithium ceramic being developed. Two models of tritium breeding and neutron multiplying elements of the TBZ were designed, manufactured and have been tested already in IVV-2M nuclear reactor. The first model consists of lithium orthosilicate ceramic sphere pebbles (1-1.5 mm diameter) and beryllium sphere (0.1 and 1.0 mm diameter). Ceramic cylindrical pellets (11 mm diameter and 10 mm height) and porous beryllium (20% porosity) are in the second model. Some properties and microstructure of the ceramic elements are performed. Initial results of some changes of ceramic structure and in-pile experimental ratio of hydrogen and oxygen form of tritium release in helium/neon purge gas are presented. These results and outcome of irradiated LiAlO2, Li4SiO4 and Li2SiO3 ceramics in a water-graphite nuclear reactor are considered to be a DATE BASE for development of the Test Modules and the DEMO blanket and influence of the kinetic tritium release parameters on DEMO tritium systems are discussed. (author)

  3. Preliminary safety analysis of a thorium high-conversion pebble bed reactor

    International Nuclear Information System (INIS)

    An inherently safe thorium High-Conversion Pebble Bed Reactor would combine the inherent safety characteristics of the Pebble Bed Reactor with the favourable waste characteristics and resource availability of the thorium fuel cycle. Previous work by the authors showed that high conversion ratio's can be achieved within a thorium Pebble Bed Reactor (PBR) at a practical operating regime. The thorium PBR core design consists of a cylindrical core with a central driver zone surrounded by a breeder zone. The breeder pebbles have a 30 g heavy metal (HM) loading to enhance conversion of Th-232 into U-233, while the driver pebbles (10 w% U-233) contain a lower metal loading to enhance fission. In previous studies, thorium PBR designs were presented for three core diameters, using a 7.5 g heavy metal (HM) loading for the driver pebbles. The current paper investigates the safety of these thorium PBR designs in terms of reactivity coefficients and possible reactivity insertion due to water ingress. Early results indicated that the values of the reactivity coefficients for the three designs with 7.5 g HM loading per driver pebble were rather small and the possible reactivity insertion due to water ingress was very large. Therefore, also a lower HM loading per driver pebble (4 g) was investigated to reduce the impact of water ingress, since the core becomes less under-moderated. For the three core diameters investigated, it is shown that reducing the metal loading in the driver pebbles to 4 g is indeed advantageous in terms of safety, water ingress leads to a smaller reactivity increase but also the reactivity coefficients become stronger negative. Secondly, the breeding performance of the cores with a 4 g driver pebble HM loading improves. On the downside, the driver pebble residence times become shorter, which could increase fuel reprocessing costs. Fuel pebbles would have to be recycled at an increased rate, which might be more challenging from a practical perspective

  4. Gas bubble network formation in irradiated beryllium pebbles monitored by X-Ray micro-tomography

    International Nuclear Information System (INIS)

    Full text of publication follows: The efficient and safe operation of helium cooled ceramic breeder blankets requires among others an efficient tritium release during operation at blanket relevant temperatures. In the past out-of-pile thermal desorption studies on low temperature neutron irradiated beryllium have shown that tritium and helium release peaks occur together. This phenomenon can be interpreted in terms of growth and coalescence of helium bubbles and tritium that either is trapped inside the helium bubbles in form of T2 molecules or in their strain field. With increasing temperature the bubble density and size at grain interfaces increase together with the probability of interconnected porosities and channel formation to the outer surface, leading to simultaneous helium and tritium release peaks in TDS. For a reliable prediction of gas release up to end-of-life conditions at blanket relevant temperatures, knowledge of the dynamics of bubble growth and coalescence as well as the 3D distribution of bubble network formation is indispensable. Such data could also be used to experimentally validate any future model predictions of tritium and helium release rates. A high resolution computer aided micro-tomography (CMT) setup has been developed at the European Synchrotron Radiation Facility which allowed reconstructing 3-D images of beryllium pebbles without damaging them. By postprocessing the data a 3D rendering of inner surfaces and of interconnected channel networks can be obtained, thus allowing the identification of open porosities in neutron irradiated and tempered beryllium pebbles. In our case Beryllium pebbles of 2 mm diameter had been neutron irradiated in the 'Beryllium' experiment at 770 K with 1.24 x 1025 nxm-2 resulting in 480 appm He and 12 appm Tritium. After annealing at 1500 K CMT was performed on the pebbles with 4.9 and 1.4 μm voxel resolution, respectively, followed by morphological and topological post-processing of the reconstructed

  5. Pebble bed modular reactor (PBMR)

    International Nuclear Information System (INIS)

    In 1993, the pebble bed modular reactor (PBMR) was identified by ESKOM, the electric utility of South Africa, as a leading option for the installation of new generating capacity to their electric grid. This innovative nuclear power plant incorporates a closed cycle primary coolant system utilizing helium to transport heat energy directly from the modular pebble bed reactor to a recuperative power conversion unit with a single-shaft turbine/compressor/generator. This replacement of the steam cycle that is common in present nuclear power plants (NPP) with a direct gas cycle provides the benefits of simplification and a substantial increase in overall system efficiency with the attendant lowering of capital and operational costs. Although the historical development of this plant is interrelated to other types of high temperature gas cooled reactors (HTGRs), the principle focus herein is on the pebble bed (spherical) fuel element type reactor. The long-term development of this reactor type began in Germany by the KFA Nuclear Research Center (now FZJ). Two pebble bed plants were constructed in Germany, the 46 MW(th)/15 MW(e) Arbeitsgemeinshaft Versuchsreaktor (AVR) and the 750 MW(th)/296 MW(e) thorium high temperature reactor (THTR-300). Basically, these steam/electric plants validated the temperature and fission product retention capabilities of the ceramic (TRISO) coated fuel particle and the safety characteristics of the HTGR. Most notable of the operational achievements was with the AVR in sustaining longterm operation at an average core outlet temperature of 950 deg. C, and in demonstration of safety such as extended loss of forced cooling on the core. More details on the AVR and THTR-300 plants are provided The next evolution of the pebble bed plant began in the early 1980s with development of the modular reactor. This small reactor added the unique characteristic of being able to cool the core entirely by passive heat transfer mechanisms following postulated

  6. Test apparatus for ITER blanket pebble packing behavior

    International Nuclear Information System (INIS)

    Current Japanese design for ITER Driver Blanket consists of three breeder layers, nine multiplier layers and five cooling panels. The breeder layers and the multiplier layers contain 1 mm diameter spheres of Li2O and Be, respectively. The heat transfer in such 'Pebble Layered Blanket' is largely affected by the packing fraction of the pebbles which can be easily changed by the vibration during the operation. The packing fraction of the pebbles are expected to be as high as possible on the view point of nuclear heat design to maintain the optimum temperature of the breeder layer. Thus, it is necessary to establish the stable packed bed of the breeder and multiplier. The present experimental apparatus was fabricated for the engineering tests with the partial model of Japanese blanket. Test apparatus consists of stainless steel test panels, transparent plastic test panels, vibrators and measurement instruments. The apparatus can examine various parameters of sphere packed beds such as packing fraction, panels deformation, loading weight at the bottom of the panels and so on under various vibrating conditions. (author)

  7. 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.

  8. Studies on tritium breeding ratio for solid breeder blanket cooled by pressurized water through nuclear and thermal analyses

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) has been performing the research, development and design of blankets with water-cooled solid breeder for fusion power plant as a leading institute in Japan, according to the long-term R and D program established by the Fusion Council in 1999. For our design, pebbles of a ceramic tritium breeder (Li2TiO3) and a beryllium neutron multiplier (Be) are packed in the constitutive layer structures of a test blanket module (TBM) for ITER. These reports are results of one-dimensional nuclear and thermal analyses on the TBM emphasizing on optimized configuration of the breeder and multiplier layers. Taking into account increment of TBR, the radial widths of the breeder and multiplier layers are optimized. The main results of our study are as follows: (1) In multilayered structures of pebble beds, existence of the peak of the TBR was revealed within the range of the volume ratio R=V(Be)/V(Li2TiO3)=4-5. (2) In the case of optimized layer structure for the single packing, a layer of Be was set to be the two layers behind a layer of Li2TiO3. The R became available for staying in the range of R=4-5. Consequently, the TBR respectively increased by 2.0%, 3.2% and 4.0% with 7.5%(nature), 40% and 90% of enrichment of 6Li compared to TBR of TBM in which the layers of Be and Li2TiO3 were interlaminated. This database of TBR for optimized layer structure contributes to the estimation of TBR at the design stage of the TBM and demonstration blanket aimed to strengthen the commercial competitiveness and technical feasibility. (author)

  9. 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

  10. Pebble-bed pebble motion: Simulation and Applications

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2011-11-01

    Pebble bed reactors (PBR) have moving graphite fuel pebbles. This unique feature provides advantages, but also means that simulation of the reactor requires understanding the typical motion and location of the granular flow of pebbles. This report presents a method for simulation of motion of the pebbles in a PBR. A new mechanical motion simulator, PEBBLES, efficiently simulates the key elements of motion of the pebbles in a PBR. This model simulates gravitational force and contact forces including kinetic and true static friction. It's used for a variety of tasks including simulation of the effect of earthquakes on a PBR, calculation of packing fractions, Dancoff factors, pebble wear and the pebble force on the walls. The simulator includes a new differential static friction model for the varied geometries of PBRs. A new static friction benchmark was devised via analytically solving the mechanics equations to determine the minimum pebble-to-pebble friction and pebble-to-surface friction for a five pebble pyramid. This pyramid check as well as a comparison to the Janssen formula was used to test the new static friction equations. Because larger pebble bed simulations involve hundreds of thousands of pebbles and long periods of time, the PEBBLES code has been parallelized. PEBBLES runs on shared memory architectures and distributed memory architectures. For the shared memory architecture, the code uses a new O(n) lock-less parallel collision detection algorithm to determine which pebbles are likely to be in contact. The new collision detection algorithm improves on the traditional non-parallel O(n log(n)) collision detection algorithm. These features combine to form a fast parallel pebble motion simulation. The PEBBLES code provides new capabilities for understanding and optimizing PBRs. The PEBBLES code has provided the pebble motion data required to calculate the motion of pebbles during a simulated earthquake. The PEBBLES code provides the ability to

  11. Pebble Puzzle Solved

    Science.gov (United States)

    2004-01-01

    [figure removed for brevity, see original site] Figure 1 In the quest to determine if a pebble was jamming the rock abrasion tool on NASA's Mars Exploration Rover Opportunity, scientists and engineers examined this up-close, approximate true-color image of the tool. The picture was taken by the rover's panoramic camera, using filters centered at 601, 535, and 482 nanometers, at 12:47 local solar time on sol 200 (August 16, 2004). Colored spots have been drawn on this image corresponding to regions where panoramic camera reflectance spectra were acquired (see chart in Figure 1). Those regions are: the grinding wheel heads (yellow); the rock abrasion tool magnets (green); the supposed pebble (red); a sunlit portion of the aluminum rock abrasion tool housing (purple); and a shadowed portion of the rock abrasion tool housing (brown). These spectra demonstrated that the composition of the supposed pebble was clearly different from that of the sunlit and shadowed portions of the rock abrasion tool, while similar to that of the dust-coated rock abrasion tool magnets and grinding heads. This led the team to conclude that the object disabling the rock abrasion tool was indeed a martian pebble.

  12. The development of breeder reactors in the US

    International Nuclear Information System (INIS)

    This article discusses the early history of breeder development in the US, the early history of the fast reactor in the US, changes during the Carter administration, and the development of LMFBR technology. Topics considered include the intermediate-energy plutonium breeder, the molten plutonium breeder, the aqueous homogeneous reactor, the molten-salt reactor, the liquid metal-fueled reactor, electronuclear breeding, the Experimental Breeder Reactor-I, the Experimental Breeder Reactor-II, the Enrico Fermi Reactor, a programmatic change to ceramic fuel, the South East Fast Oxide Reactor, the sodium void coefficient, the 1000-MWe studies of 1964, the 1000-MWe studies of 1967-1969, the FARET design, the Fast Flux Test Facility, the Clinch River Breeder Reactor (CRBR), the gas-cooled fast breeder, the light-water breeder, materials for cladding and duct walls, and reactor safety. It is pointed out that the Congress opposes the construction of the CRBR, while the Reagan administration strongly supports it

  13. Low temperature tritium release experiment from lithium titanate breeder material

    International Nuclear Information System (INIS)

    Engineering data of neutron irradiation performance are needed to design a fusion blanket. Of the engineering data, tritium release characteristic is one of the most important data. Therefore, tritium release experiments of the tritium breeding materials were carried out to evaluate the effects of various parameters, i.e. sweep-gas flow rate, irradiation temperature, hydrogen content in sweep gas and so on, on tritium release. Lithium titanate (Li2TiO3) is a candidate tritium breeding material for the blanket design of International Thermonuclear Experimental Reactor (ITER). As for the shape of the breeder material, a small spherical form is preferred to enhance tritium release from the breeder and to reduce the induced thermal stress in the breeder. Li2TiO3 pebbles with a diameter of 1mm and a total weight of ∼134g have been fabricated, and a pebble-pac assembly of the Li2TiO3 pebbles was irradiated in the Japan Materials Testing Reactor (JMTR), for 3 cycles (about 75 days). The tritium generated in breeder, and released from the breeder was swept downstream by the sweep gas for on-line analysis of tritium content. The total concentration and gaseous concentration of tritium released from the Li2TiO3 pebbles were measured, and HT/(HT+HTO) ratio was evaluated. The sweep-gas flow rate was changed from 10 to 1,000cm3/min, and hydrogen concentration in the sweep gas was changed from 100 to 10,000 ppm. The irradiation temperature of the outer region of the pebble-pac assembly was held below 450degC. The results showed that tritium release from the Li2TiO3 pebbles was started between 100 and 140degC and that the amount of released with increasing the irradiation temperature. The sweep-gas flow rate did not have an effect on tritium release from the Li2TiO3 pebble bed in the steady state. On the other hand, the hydrogen content in the sweep gas had an effect on the tritium release from the Li2TiO3 pebble bed. (author)

  14. Fabrication tests of Li2TiO3 pebbles by direct wet process

    International Nuclear Information System (INIS)

    Lithium titanate (Li2TiO3) pebbles are considered to be the candidate material of the tritium breeders for fusion reactor from a point of good tritium recovery, chemical stability, etc. The direct wet process that Li2TiO3 pebbles were fabricated from the Li2TiO3 solution directly was proposed. In this study, pebble fabrication tests by the direct wet process were performed. The results from the preliminary test were as follows: 1) 100% Li2TiO3 powder could be dissolved when the holding time at more than 60 C was longer. 2) Good gel shape was maintained by dropping the Li2TiO3 condensed solution liquid in acetone. 3) Adjustment of a solution influenced the cracking rate of the Li2TiO3 pebble surface. Additionally, the solvent exchange was effective to decrease the crack of Li2TiO3 pebble surface and to improve the density of Li2TiO3 pebbles. It was clear that Li2TiO3 pebbles could be fabricated by the direct wet process and the pebbles with 5 μm grain and uniform structure were obtained. (orig.)

  15. Evaluation of tritium release properties of advanced tritium breeders

    International Nuclear Information System (INIS)

    Demonstration power plant (DEMO) fusion reactors require advanced tritium breeders with high thermal stability. Lithium titanate (Li2TiO3) advanced tritium breeders with excess Li (Li2+xTiO3+y) are stable in a reducing atmosphere at high temperatures. Although the tritium release properties of tritium breeders are documented in databases for DEMO blanket design, no in situ examination under fusion neutron (DT neutron) irradiation has been performed. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed, and DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. Considering the tritium release characteristics, the optimum grain size after sintering is <5 μm. From the results of the optimization of granulation conditions, prototype Li2+xTiO3+y pebbles with optimum grain size (<5 μm) were successfully fabricated. The Li2+xTiO3+y pebbles exhibited good tritium release properties similar to the Li2TiO3 pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water. (authors)

  16. Fabrication, properties, and tritium recovery from solid 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 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

  17. Interim report on core physics and fuel cycle analysis of the pebble bed reactor power plant concept

    International Nuclear Information System (INIS)

    Calculations were made to predict the performance of a pebble bed reactor operated in a mode to produce fissile fuel (high conversion or breeding). Both a one pebble design and a design involving large primary feed pebbles and small fertile pebbles were considered. A relatively short residence time of the primary pebbles loaded with 233U fuel was found to be necessary to achieve a high breeding ratio, but this leads to relatively high fuel costs. A high fissile inventory is associated with a low C/Th ratio and a high thorium loading, causing the doubling time to be long, even though the breeding ratio is high, and the fuel cost of electrical product to be high. Production of 233U fuel from 235U feed was studied and performances of the converter and breeder reactor concepts were examined varying the key parameters

  18. Neutronics Experiment on A HCPB Breeder Blanket Mock-Up

    International Nuclear Information System (INIS)

    A neutronics experiment has been performed in the frame of European Fusion Technology Program on a mock-up of the EU Test Blanket Module (TBM), Helium Cooled Pebble Bed (HCPB) concept, with the objective to validate the capability of nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. The experiment has been carried out at the FNG 14-MeV neutron source in collaboration between ENEA, Technische Universitaet Dresden, Forschungszentrum Karlsruhe, J. Stefan Institute Ljubljana and with the participation of JAEA. The mock-up, designed in such a way to replicate all relevant nuclear features of the TBM-HCPB, consisted of a steel box containing beryllium block and two intermediate steel cassettes, filled with of Li2CO3 powder, replicating the breeder insert main characteristics: radial thickness, distance between ceramic layers, thickness of ceramic layers and of steel walls. In the experiment, the TPR has been measured using Li2CO3 pellets at various depths at two symmetrical positions at each depth, one in the upper and one in the lower cassette. Twelve pellets were used at each position to determine the TPR profile through the cassette. Three independent measurements were performed by ENEA, TUD/VKTA and JAEA. The neutron flux in the beryllium layer was measured as well using activation foils. The measured tritium production in the TBM (E) was compared with the same quantity (C) calculated by the MCNP.4c using a very detailed model of the experimental set up, and using neutron cross sections from the European Fusion File (EFF ver.3.1) and from the Fusion Evaluated Nuclear Data Library (FENDL ver. 2.1, ITER reference neutron library). C/E ratios were obtained with a total uncertainty on the C/E comparison less than 9% (2 s). A sensitivity and uncertainty analysis has also been performed to evaluate the calculation uncertainty due to the uncertainty on neutron cross sections. The results of such analysis

  19. Packed fluidization, enhancement of heat transfer in pebble bed and thermonuclear fusion technology

    International Nuclear Information System (INIS)

    Packed fluidization is a technique in which small particles (size: 100-800 μm) are allowed to fluidize in the interstices of stationary pebbles (size: >1.0 mm). Packed fluidization enhances the rate of heat transfer in pebble bed at low operative gas velocity as well as at low pressure drop across the bed. Experiments were conducted to study heat transfer in unary packed bed and binary packed fluidized bed using lithium titanate and alumina pebbles (size: 3-10 mm) and lithium titanate and silica particles (size: 231-780 μm). It was found that due to packed fluidization the rate of heat transfer is enhanced and arms of the effective thermal conductivity this enhancement was up to 260%. Low thermal conductivity of pebble bed of solid breeder materials is one of the adverse key issues which must be addressed properly for the successful development of the thermonuclear fusion technology. Packed fluidization enhances the effective thermal conductivity of the pebble bed of solid breeder materials in the Test Blanket Module (TBM) of ITER type fusion reactor. (author)

  20. Solid breeder blanket concepts

    International Nuclear Information System (INIS)

    An investigation is made of a mechanical concept for the blanket with solid breeders in view of the possible adaptation to power reactor. A special arrangement of the multiplier and breeder materials is developed to permit a further neutronic optimisation

  1. 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.

  2. 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.)

  3. 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.)

  4. Nuclear analyses for two 'look-alike' helium-cooled pebble bed test blanket sub-modules proposed by the US for testing in ITER

    International Nuclear Information System (INIS)

    The US is proposing two 'look-alike' sub-modules, based on helium-cooled pebble bed (HCPB) ceramic breeder, to be tested in the same test blanket module (TBM) that will occupy a quarter of a port in ITER and placed next to the Japanese TBM. The TBM has a toroidal width of 73 cm, a radial depth of 60 cm and a poloidal height of 91 cm. The ceramic breeder is made of Li4SiO4 with 75% Li-6 enrichment (60% packing factor) and beryllium is used as the multiplier. The two sub-modules are arranged in two configurations, namely a layered configuration and an edge-on configuration. In the present work, we analyze these two sub-modules using two-dimensional discrete ordinates transport codes in R-θ model that accounts for the presence of the ITER shielding blanket and the surrounding frame of the port. The objectives are: (1) to examine the profiles of heating and tritium production rates in the two sub-modules, both in the radial and toroidal direction, in order to identify locations where neutronics measurements can be best performed with least perturbation from the surroundings (2) to provide both local and integrated values for nuclear heating rates required for subsequent thermo-mechanics analysis, and (3) to compare the tritium production capabilities of two variants for the HCPB blanket concept, mainly the parallel and the edge-on configurations. We present the main findings from this study in this paper

  5. RAZVOJ APLIKACIJ ZA PAMETNO URO PEBBLE

    OpenAIRE

    Kranjc, Denis

    2015-01-01

    V diplomski nalogi predstavljamo razvoj aplikacij za pametno uro Pebble in razvoj aplikacij za pametni telefon Android, ki komunicira z uro preko Bluetooth povezave. Pri razvoju smo uporabili razvojno okolje CloudPebble, programski razvojni paket Pebble SDK, razvojno okolje Android Studio in javansko knjižnico PebbleKit. Aplikacije za pametno uro smo razvijali v programskem jeziku C, aplikacije Android pa v programskem jeziku Java. Rezultat diplomskega dela je osem razvitih različnih aplikaci...

  6. Evidence for Pebbles in Comets

    CERN Document Server

    Kretke, K A

    2015-01-01

    When the EPOXI spacecraft flew by Comet 103P/Hartley 2, it observed large particles floating around the comet nucleus. These particles are likely low-density, centimeter- to decimeter-sized clumps of ice and dust. While the origin of these objects remains somewhat mysterious, it is possible that they are giving us important information about the earliest stages of our Solar System's formation. Recent advancements in planet formation theory suggest that planetesimals (or cometestimals) may grow directly from the gravitational collapse of aerodynamically concentrated small particles, often referred to as "pebbles." Here we show that the particles observed in the coma of 103P are consistent with the sizes of pebbles expected to efficiently form planetesimals in the region that this comet likely formed, while smaller pebbles are may be expected in the majority of comets, whose chemistry is often indicative of formation in the colder, outer regions of the protoplanetary disk.

  7. Fast breeder reactor

    International Nuclear Information System (INIS)

    The fluid-cooled fast breeder reactor described includes an outer cylindrical boundary wall, a plurality of canless fuel elements and breeder material elements received within the boundary wall and being in an array therein forming a fissionable fuel zone and a breeder material zone coaxially surrounding the fissionable fuel zone, a coolant supply system for applying fluid coolant at uniform pressure to the entire cross section within the cylindrical boundary wall, and flow guide devices extending substantially horizontally and disposed at different levels one above the other within the breeder material zone which coaxially surrounds the fissionable fuel zone, means for elastically securing the flow guide devices at alternate levels within the breeder material to the boundary wall, the flow guide devices at the levels intermediate the alternate levels being spaced by an annular gap from the boundary wall. 7 claims, 7 drawing figures

  8. Embattled breeder reactor

    International Nuclear Information System (INIS)

    A commercial fuel-cloning machine, a nuclear breeder reactor, is yet to produce electricity in the United States. It is expensive in capital and fuel costs, its fuel that must be reprocessed can become a link to nuclear weapons manufacture, and its safety is no greater than conventional nuclear reactors. The breeder has had on-again/off-again administrative support from Washington. Opponents worry about escalating costs and failure to develop alternatives like solar energy. Proponents say fossil-fuel depletion will eventually force long-term renewable resources such as the breeder anyway. Some who share parts of both views oppose present policy regarding the Clinch River Breeder demonstration plant specifically. The correct choices on breeder concept development and commercialization will be known in 2050. 3 figures

  9. Conceptual design of solid breeder blanket system cooled by supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio; Akiba, Masato [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment] [and others

    2001-12-01

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li{sub 2}TiO{sub 3} or Li{sub 2}O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for

  10. Conceptual design of solid breeder blanket system cooled by supercritical water

    International Nuclear Information System (INIS)

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li2TiO3 or Li2O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for the energy

  11. Status of the database for solid breeder materials

    International Nuclear Information System (INIS)

    The databases for solid breeder ceramics (Li2O, Li4SiO4, Li2ZrO3 and LiAl02) and beryllium multiplier material were critically reviewed and evaluated as part of the ITER/CDA design effort (1988-1990). The results have been documented in a detailed technical report which includes progress made in expanding the solid breeder and beryllium databases up through September 1993. Emphasis was placed on the physical, thermal, mechanical, chemical-stability/compatibility, tritium retention/release and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Materials properties correlations were selected for use in design analysis, and ranges of input parameters (e.g., temperature, porosity, etc.) were established. The need for updating the ceramic breeder database was discussed at the Third Ceramic Breeder Blanket Interactions (CBBI-3) workshop at UCLA in June 1994. Progress made in expanding the ceramic breeder database and plans for updating the database are discussed

  12. 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

  13. 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.

  14. Pebble Bed Reactor Dust Production Model

    International Nuclear Information System (INIS)

    The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production

  15. Pebble Bed Reactor: core physics and fuel cycle analysis

    International Nuclear Information System (INIS)

    The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes

  16. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

    There is a considerable world-wide effort directed toward the production of materials for fusion reactors. Many ceramic fabrication groups are working on making lithium ceramics in a variety of forms, to be incorporated into the tritium breeding blanket which will surround the fusion reactor. Current blanket designs include ceramic in either monolithic or packed sphere bed (sphere-pac) forms. The major thrust at AECL is the production of lithium aluminate spheres to be incorporated in a sphere-pac bed. Contemporary studies on breeder blanket design offer little insight into the requirements on the sizes of the spheres. This study examined the parameters which determine the properties of pressure drop and coolant requirements. It was determined that an optimised sphere-pac bed would be composed of two diameters of spheres: 75 weight % at 3 mm and 25 weight % at 0.3 mm

  17. Reactor vessel for pebble beds

    International Nuclear Information System (INIS)

    The wall and the bottom of the vessel for the gas-cooled pebble-bed reactor consist of numerous blocks of graphite or carbon rock piled up. They are held together by an exterior cylindrical or polygonal ring and supported by a foundation. The blocks form coherent sectors resp. annular sectors with well-defined separating lines. At high temperatures or load change operation these sectors behave like monolithic blocks expanding heely and contracting again, the center of the vessel remaining fixed. The forces causing the compression result from the own weight of the sectors and the weight of the pebble bed. This motion is supported by the convex arrangement of the opposite surfaces of the sectors and the supporting walls and by roller bearings. The bottom of the vessel may be designed funnel-shaped, in this way facilitating the removal of spheres. (DG)

  18. A smoother pebble mathematical explorations

    CERN Document Server

    Benson, Donald C

    2003-01-01

    Introduction. I. BRIDGING THE GAP. 1. The Ancient Fractions. 2. Greek Gifts. 3. The Music of the Ratios. II. THE SHAPE OF THINGS. 4. Tubeland. 5. The Calculating Eye. III. THE GREAT ART. 6. Algebra Rules. 7. The Root of the Problem. 8. Symmetry Without Fear. 9. The Magic Mirror. IV. A SMOOTHER PEBBLE. 10. On the Shoulders of Giants. 11. Six-Minute Calculus. 12. Roller-Coaster Science. Glossary. References. Index

  19. Fabrication of Li2TiO3 pebbles using PVA–boric acid reaction for solid breeding materials

    International Nuclear Information System (INIS)

    Highlights: • Li2TiO3 pebbles were successfully fabricated by the slurry droplet wetting method. • Boron was used as hardening agent of PVA and completely removed during sintering. • Microstructure of fabricated Li2TiO3 pebble was exceptionally homogeneous. • Suitable process conditions for high-quality Li2TiO3 pebble were summarized. - Abstract: Lithium metatitanate (Li2TiO3) is a candidate breeding material of the Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM). The breeding material is used in pebble-bed form to reduce the uncertainty of the interface thermal conductance. In this study, Li2TiO3 pebbles were successfully fabricated by the slurry droplet wetting method using the cross-linking reaction between polyvinyl alcohol (PVA) and boric acid. The effects of fabrication parameters on the shaping of Li2TiO3 green body were investigated. In addition, the basic characteristics of the sintered pebble were also evaluated. The shape of Li2TiO3 green bodies was affected by slurry viscosity, PVA content and boric acid content. The grain size and average crush load of sintered Li2TiO3 pebble were controlled by the sintering time. The boron was completely removed during the final sintering process

  20. The Karlsruhe solid breeder blanket and the test module to be irradiated in ITER/NET

    International Nuclear Information System (INIS)

    The blanket for the DEMO reactor should operate at an average neutron flux of 2.2 MW/m2 for 20000 h. This requires the use of a structural material which can withstand high neutron fluences without swelling. The ferritic steel Manet was chosen for this purpose. The breeder material is in the form of Li4SiO4 pebbles of 0.35 to 0.6 mm diameter. The 6 mm thick beds of pebbles are placed between beryllium plates which are cooled by high pressure helium flowing inside steel tubes. Breeder material and beryllium are contained in radial canisters, placed inside boxes. The coolant helium enters the blanket at 250deg C, cools first the box walls and then the breeder and multiplier, and leaves the blanket at 450deg C. The maximum temperature in the first wall steel is 550deg C, while the minimum and maximum temperatures in the breeder are 380 and 820deg C, respectively. The resulting total tritium inventory in the breeder is only 10 g, and the real tridimensional tritium breeding ratio is 1.11. The conceptual design of the test module, of its extraction system and of the required out-of-reactor ancillary systems has allowed an estimate of the time constants of the various components and thus allowed an assessment of the requirements given by the testing of the modules on the NET/ITER machine. (orig.)

  1. 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)

  2. 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

  3. Disposition of plutonium with HTGRs using Pu burner balls and Th breeder balls

    International Nuclear Information System (INIS)

    A concept of reactor system was developed with which weapons-grade Plutonium could be made perfectly worthless in use for weapons. It is a pebble bed type HTGR using Pu burner ball fuels and Th breeder ball fuels. The residual amounts of 239Pu in spent Pu balls become less than 1% of the initial loading. The power coefficient was made negative by reducing the parasitic neutron absorption reaction rate of 135Xe. (author)

  4. Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Mr. Baron says the administration's effort to terminate the Clinch River Breeder Reactor (CRBR) project is symptomatic; they have also placed restrictions on fusion, coal, solar, and other areas of energy development in which technological advances are held back in order to force conservation. Because the breeder reactor, unlike solar and fusion energy, is both economically and technically feasible, a demonstration plant is needed. The contentions that the CRBR design is obsolete, that its proposed size is inappropriate, or that plutonium can be diverted for weapons proliferation are argued to be invalid. Failure to complete the CRBR will have both economic and national security repercussions

  5. Development of advanced blanket materials for solid breeder blanket of fusion reactor

    International Nuclear Information System (INIS)

    The design of advanced solid breeding blanket in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high fluence, and the development of such as advanced blanket materials has been carried out by the cooperation activities among JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by wet process is a reference material as a tritium breeder, but the stability on high temperature has to be improved for application to DEMO blanket. As one of such the improved materials, TiO2-doped Li2TiO3 pebbles were successfully fabricated and TiO2-doped Li2TiO3 has been studied. For the advanced neutron multiplier, the beryllides that have high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that Be12Ti had lower swelling and tritium inventory than that of beryllium metal. The pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. From these activities, the bright prospect was obtained to realize the DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides. (author)

  6. Tritium dynamics in fusion reactor solid breeder

    International Nuclear Information System (INIS)

    In the field of the NET research progrm, the chemical and diffusive processes involved in solid ceramic breeder materials have been analysed. A mathematical model describing the phenomena has been developed to obtain a quantitative evaluation for a first design approach. The data obtained by means of the above mentioned model are in good agreement with the data obtained by other research groups working in Europe and in United States. The computer codes BLANKET2, MC2, FWBC, have been developed to simulate the phenomena

  7. 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.

  8. Postirradiation examination of beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    Postirradiation examinations of COBRA-1A beryllium pebbles irradiated in the EBR-II fast reactor at neutron fluences which generated 2700--3700 appm helium have been performed. Measurements included density change, optical microscopy, scanning electron microscopy, and transmission electron microscopy. The major change in microstructure is development of unusually shaped helium bubbles forming as highly non-equiaxed thin platelet-like cavities on the basal plane. Measurement of the swelling due to cavity formation was in good agreement with density change measurements.

  9. Formation of pebble-pile planetesimals

    CERN Document Server

    Jansson, Karl Wahlberg

    2014-01-01

    The first stage of planet formation is the accumulation of dust and ice grains into mm-cm-sized pebbles. These pebbles can clump together through the streaming instability and form gravitationally bound pebble 'clouds'. Pebbles inside such a cloud will undergo mutual collisions, dissipating energy into heat. As the cloud loses energy, it gradually contracts towards solid density. We model this process and investigate two important properties of the collapse: (i) the timescale of the collapse and (ii) the temporal evolution of the pebble size distribution. Our numerical model of the pebble cloud is zero-dimensional and treats collisions with a statistical method. We find that planetesimals with radii larger than 100 km collapse on the free-fall timescale of about 25 years. Lower-mass clouds have longer pebble collision timescales and collapse much more slowly, with collapse times of a few hundred years for 10-km-scale planetesimals and a few thousand years for 1-km-scale planetesimals. The mass of the pebble c...

  10. Feasibility study of a fission-suppressed tokamak fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Neef, W.S.; Berwald, D.H.; Garner, J.K.; Whitley, R.H.; Ghoniem, N.; Wong, C.P.C.; Maya, I.; Schultz, K.R.

    1984-12-01

    The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m/sup 2/ and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of /sup 233/U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U/sub 3/O/sub 8/ depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management.

  11. Feasibility study of a fission-suppressed tokamak fusion breeder

    International Nuclear Information System (INIS)

    The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m2 and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of 233U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U3O8 depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management

  12. Mechanism analysis of quasi-static dense pebble flow in pebble bed reactor using phenomenological approach

    International Nuclear Information System (INIS)

    Highlights: ► We introduced four basic forms of phenomenological method for pebble flow. ► We discussed the physical nature of the quasi-static pebble flow. ► We verified the applicability of the discrete element method. ► We investigated the parameter effects on quasi-static pebble flow. - Abstract: By means of the four basic forms of the phenomenological method, experimental results have intuitionally disclosed the physical mechanism from various views of the quasi-static pebble flow in a pebble bed reactor and successfully verified the availability of the discrete element method, on which the parameter effects have been investigated, including different base cone angle and different friction coefficient. The flow fields under different parameters have been discussed. On the basis of these researches, a framework of the general understanding of pebble flow mechanism has been drawn; many essential problems are discussed, including the interpretation of the quasi-static pebble flow, force analysis inside the pebble packing, propagation and distribution of the voids, internal equilibrium arches, competition mechanism, internal collapse, self-organization, equivalent shear force, equivalent normal force, the physical process of stagnant zone's influence on the overall flow field, and so on. All of these are very helpful to understand the physical mechanism of the quasi-static pebble flow in a pebble bed reactor.

  13. 3D tomography analysis of the inner structure of pebbles and pebble beds

    International Nuclear Information System (INIS)

    An analytical tool to monitor the arrangement of pebbles in a pebble bed as well as the morphology of gas bubbles in as fabricated and neutron irradiated beryllium pebbles is presented. The context of this study is the Helium Cooled Pebble Bed (HPCB) blanket design for the forthcoming generation of fusion reactors. The thermal-mechanical behavior of pebble beds is a basic issue for the HPCB. It depends strongly on the configuration of the pebbles in the bed, and in particular on the number of contacts between pebbles, and between pebbles and the blanket walls. The related contact surfaces play also a major role. The knowledge on the inner structure of the pebbles is required since during the life cycle of a power reactor helium and tritium bubbles are produced inside the beryllium pebbles and the tritium build-up can be in excess of several kilograms, being thereby a key safety issue. All the non-destructive analyses are based on 3D computer aided microtomography using a very powerful synchrotron radiation x-ray source with high spatial resolution. The data analysis relies on a topological operator called filtered medial line applied to the entire data volumes and the related graph representation. By this technique the number of contacts between the pebbles in pebble packs and their angular distribution are obtained, as well as the corresponding contact surfaces. The evaluation of bubble sizes and densities in single pebbles, the assessment of the pore channel network topology, the 3D reconstruction of the fraction of interconnected bubble porosity, and the open-to-closed-porosity ratio are among the most interesting findings. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  14. Preliminary structural design and thermo-mechanical analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    International Nuclear Information System (INIS)

    Highlights: • A helium cooled solid breeder blanket module was designed for CFETR. • Multilayer U-shaped pebble beds were adopted in the blanket module. • Thermal and thermo-mechanical analyses were carried out under normal operating conditions. • The analysis results were found to be acceptable. - Abstract: With the aim to bridge the R&D gap between ITER and fusion power plant, the Chinese Fusion Engineering Test Reactor (CFETR) was proposed to be built in China. The mission of CFETR is to address the essential R&D issues for achieving practical fusion energy. Its blanket is required to be tritium self-sufficient. In this paper, a helium cooled solid breeder blanket adopting multilayer U-shaped pebble beds was designed and analyzed. Thermo-mechanical analysis of the first wall and side wall combined with breeder unit was carried out for normal operating steady state conditions. The results showed that the maximum temperatures of the structural material, neutron multiplier and tritium breeder pebble beds are 523 °C, 558 °C and 787 °C, respectively, which are below the corresponding limits of 550 °C, 650 °C and 920 °C. The maximum equivalent stress of the structure is under the allowable value with a margin about 14.5%

  15. 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

  16. Absorber rod for pebble-bed reactor

    International Nuclear Information System (INIS)

    The absorber rod that can be moved into the pebble bed from the top reflector is enclosed by a cladding tube which, if it is completely moved down, ends above the pebble bed and is open at the bottom. Through the cladding tube the absorber rod is cooled with gas. The cladding tube consists of e.g. boron steel. If the absorber rod is drawn it takes along the cladding tube which is moved into the guide tube like a telescope. The rigidity of that part of the absorber rod projecting from the pebble bed is thus guaranteed. (DG)

  17. Fabrication of Li2TiO3 pebbles with small grain size via hydrothermal and improved dry-rolling methods

    International Nuclear Information System (INIS)

    Highlights: • The fabricated pebbles can be densified (81% T.D.) at a low sintering temperature (850 °C). • The pebbles’ size can be controlled during the fabrication process. • Average grain size of the Li2TiO3 pebbles is less than 1 μm (0.82 μm). • The molar ratio of Li to Ti of the pebbles sintered at 850 °C keeps the value of 1.97 after sintering. - Abstract: Lithium titanate (Li2TiO3) ceramic pebbles were successfully fabricated by using hydrothermal and improved dry-rolling method. In the present work, ultra-fine Li2TiO3 powder of high reactivity was prepared via hydrothermal reaction, using anatase titania and lithium hydroxide as raw materials. The as-synthesized Li2TiO3 powder exhibits an average crystalline size as small as 100 nm. Improved dry-rolling method was employed to fabricate Li2TiO3 pebbles. The green pebbles can be well-sintered (81% T.D.) at a temperature as low as 850 °C for 3 h. The pebbles have good sphericity (1.08) and narrow diameter distribution (1.0–1.2 mm) with a crush load of 35 N. Scanning electron microscope (SEM) observations of pebbles showed that the ceramic grain size was below 1 μm and atomic emission spectrometer fitted with inductively coupled plasma (ICP-AES) results confirmed that atomic ratio of Li to Ti in the fabricated pebbles was 1.97

  18. Fabrication of Li{sub 2}TiO{sub 3} pebbles with small grain size via hydrothermal and improved dry-rolling methods

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Wen; Zhou, Qilai; Xue, Lihong, E-mail: xuelh@mail.hust.edu.cn; Yan, Youwei

    2015-09-15

    Highlights: • The fabricated pebbles can be densified (81% T.D.) at a low sintering temperature (850 °C). • The pebbles’ size can be controlled during the fabrication process. • Average grain size of the Li{sub 2}TiO{sub 3} pebbles is less than 1 μm (0.82 μm). • The molar ratio of Li to Ti of the pebbles sintered at 850 °C keeps the value of 1.97 after sintering. - Abstract: Lithium titanate (Li{sub 2}TiO{sub 3}) ceramic pebbles were successfully fabricated by using hydrothermal and improved dry-rolling method. In the present work, ultra-fine Li{sub 2}TiO{sub 3} powder of high reactivity was prepared via hydrothermal reaction, using anatase titania and lithium hydroxide as raw materials. The as-synthesized Li{sub 2}TiO{sub 3} powder exhibits an average crystalline size as small as 100 nm. Improved dry-rolling method was employed to fabricate Li{sub 2}TiO{sub 3} pebbles. The green pebbles can be well-sintered (81% T.D.) at a temperature as low as 850 °C for 3 h. The pebbles have good sphericity (1.08) and narrow diameter distribution (1.0–1.2 mm) with a crush load of 35 N. Scanning electron microscope (SEM) observations of pebbles showed that the ceramic grain size was below 1 μm and atomic emission spectrometer fitted with inductively coupled plasma (ICP-AES) results confirmed that atomic ratio of Li to Ti in the fabricated pebbles was 1.97.

  19. Microstructure analysis of melt-based lithium orthosilicate/metatitanate pebbles

    International Nuclear Information System (INIS)

    Lithium containing ceramics, such as lithium orthosilicate (Li4SiO4) and lithium metatitanate (Li2TiO3), are being developed to be used as tritium sources for future fusion reactors. In the current design of the Helium-Cooled Pebble Bed Blanket, pebbles with a diameter of approximately 1 mm are featured in pebble beds in so called blanket modules surrounding the plasma core. A modified single drop, melt-based process has been developed for the production of the pebbles. This method has many advantages in regard to the yield and the recycling potential of the used material. The current reference material is produced with an excess of 2.5 wt. % SiO2, which results in a 2-phase material composed of roughly 90 mol % Li4SiO4 and 10 mol % Li2SiO3, lithium metasilicate. There are generally two possible methods to increase the mechanical properties of these pebbles: (a) minimize production based defects such as cracks and pores, and (b) by adding stronger phases to the ceramic. Recent work has focused on the modification of the established melt-spraying process to develop a droplet generation technique with more control over the process. Both lithium orthosilicate and metatitanate have the advantage of being low activation materials, which allows the addition of titania, TiO2, to the melt, resulting in the formation of a strengthening secondary phase of lithium metatitanate. This paper looks at the influence of the phase content on the microstructure and the resulting mechanical properties. Phase analysis and porosity measurements were performed, and in particular the microstructure and mechanical crush loads were characterized. (orig.)

  20. Effect of bed configuration on pebble flow uniformity and stagnation in the pebble bed reactor

    International Nuclear Information System (INIS)

    Highlights: • Pebble flow uniformity and stagnation characteristics are very important for HTR-PM. • Arc- and brachistochrone-shaped configuration effects are studied by DEM simulation. • Best bed configurations with uniform flow and no stagnated pebbles are suggested. • Detailed quantified characteristics of bed configuration effects are shown for explanation. - Abstract: Pebble flow uniformity and stagnation characteristics are very important for the design of pebble bed high temperature gas-cooled reactor. Pebble flows inside some specifically designed contraction configurations of pebble bed are studied by discrete element method. The results show the characteristics of stagnation rates, recycling rates, radial distribution of pebble velocity and residence time. It is demonstrated clearly that the bed with a brachistochrone-shaped configuration achieves optimum levels of flow uniformity and recycling rate concentration, and almost no pebbles are stagnated in the bed. Moreover, the optimum choice among the arc-shaped bed configurations is demonstrated too. Detailed information shows the quantified characteristics of bed configuration effects on flow uniformity. In addition, a good design of the pebble bed configuration is suggested

  1. 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

  2. Swiss breeder research programme

    International Nuclear Information System (INIS)

    A new initiative for a Swiss Fast Breeder Research Program has been started during 1991. This was partly the consequence of a vote in Fall 1990, when the Swiss public voted for maintaining nuclear reactors in operation, but also for a moratorium of 10 years, within which period no new reactor project should be proposed. On the other hand the Swiss government decided to keep the option 'atomic reactors' open and therefore it was essential to have programmes which guaranteed that the knowledge of reactor technology could be maintained in the industry and the relevant research organisations. There is also motivation to support a Swiss Breeder Research Program on the part of the utilities, the licensing authorities and the Paul Scherrer Institute (PSI). The utilities recognise the breeder reactor as an advanced reactor system which has to be developed further and might be a candidate, somewhere in the future, for electricity production. In so far they have great interest that a know-how base is maintained in our country, with easy access for technical questions and close attention to the development of this reactor type. The licensing authorities have a legitimate interest that an adequate knowledge of the breeder reactor type and its functions is kept at their disposal. PSI and the former EIR have had for many years a very successful basic research programme concerning breeder reactors, and were in close cooperation with EFR. The activities within this programme had to be terminated owing to limitations in personnel and financial resources. The new PSI research programme is based upon two main areas, reactor physics and reactor thermal hydraulics. In both areas relatively small but valuable basic research tasks, the results of which are of interest to the breeder community, will be carried out. The lack of support of the former Breeder Programme led to capacity problems and finally to a total termination. Therefore one of the problems which had to be solved first was

  3. Numerical simulation of nuclear pebble bed configurations

    Energy Technology Data Exchange (ETDEWEB)

    Shams, A., E-mail: shams@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Roelofs, F., E-mail: roelofs@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Komen, E.M.J., E-mail: komen@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Baglietto, E., E-mail: emiliob@MIT.EDU [Massachusetts Institute of Technology (MIT) (United States)

    2015-08-15

    Highlights: • Numerical simulations of a single face cubic centred pebble bed are performed. • Wide range of turbulence modelling techniques are used to perform these calculations. • The methods include 1-DNS, 1-LES, 3-Hybrid (RANS/LES) and 3-RANS models, respectively. • The obtained results are extensively compared to provide guidelines for such flow regimes. • These guidelines are used to perform reference LES for a limited sized random pebble bed. - Abstract: High Temperature Reactors (HTRs) are being considered all over the world. An HTR uses helium gas as a coolant, while the moderator function is taken up by graphite. The fuel is embedded in the graphite moderator. A particular inherent safety advantage of HTR designs is that the graphite can withstand very high temperatures, that the fuel inside will stay inside the graphite pebble and cannot escape to the surroundings even in the event of loss of cooling. Generally, the core can be designed using a graphite pebble bed. Some experimental and demonstration reactors have been operated using a pebble bed design. The test reactors have shown safe and efficient operation, however questions have been raised about possible occurrence of local hot spots in the pebble bed which may affect the pebble integrity. Analysis of the fuel integrity requires detailed evaluation of local heat transport phenomena in a pebble bed, and since such phenomena cannot easily be modelled experimentally, numerical simulations are a useful tool. As a part of a European project, named Thermal Hydraulics of Innovative Nuclear Systems (THINS), a benchmarking quasi-direct numerical simulation (q-DNS) of a well-defined pebble bed configuration has been performed. This q-DNS will serve as a reference database in order to evaluate the prediction capabilities of different turbulence modelling approaches. A wide range of numerical simulations based on different available turbulence modelling approaches are performed and compared with

  4. Numerical simulation of nuclear pebble bed configurations

    International Nuclear Information System (INIS)

    Highlights: • Numerical simulations of a single face cubic centred pebble bed are performed. • Wide range of turbulence modelling techniques are used to perform these calculations. • The methods include 1-DNS, 1-LES, 3-Hybrid (RANS/LES) and 3-RANS models, respectively. • The obtained results are extensively compared to provide guidelines for such flow regimes. • These guidelines are used to perform reference LES for a limited sized random pebble bed. - Abstract: High Temperature Reactors (HTRs) are being considered all over the world. An HTR uses helium gas as a coolant, while the moderator function is taken up by graphite. The fuel is embedded in the graphite moderator. A particular inherent safety advantage of HTR designs is that the graphite can withstand very high temperatures, that the fuel inside will stay inside the graphite pebble and cannot escape to the surroundings even in the event of loss of cooling. Generally, the core can be designed using a graphite pebble bed. Some experimental and demonstration reactors have been operated using a pebble bed design. The test reactors have shown safe and efficient operation, however questions have been raised about possible occurrence of local hot spots in the pebble bed which may affect the pebble integrity. Analysis of the fuel integrity requires detailed evaluation of local heat transport phenomena in a pebble bed, and since such phenomena cannot easily be modelled experimentally, numerical simulations are a useful tool. As a part of a European project, named Thermal Hydraulics of Innovative Nuclear Systems (THINS), a benchmarking quasi-direct numerical simulation (q-DNS) of a well-defined pebble bed configuration has been performed. This q-DNS will serve as a reference database in order to evaluate the prediction capabilities of different turbulence modelling approaches. A wide range of numerical simulations based on different available turbulence modelling approaches are performed and compared with

  5. Use of plutonium in pebble bed HTGRs

    International Nuclear Information System (INIS)

    This paper provides a summary of the current status of world-wide inventories of weapon-grade plutonium and plutonium from reprocessing of power reactor fuel. It addresses the use of pebble bed HTGRs for consumption of the plutonium in terms of the fuel cycle options. The requirements and neutronics aspects, and results from parameter studies conducted using pebble bed reactor types, are discussed, along with proliferation and waste disposal aspects. (author)

  6. Mechanics of the pebble bed reactor

    International Nuclear Information System (INIS)

    In a survey, the quite different type of problems which arise for the reactor designer from the mechanics of the pebble bed are demonstrated by examples. It becomes clear why the apparently simple system of a static heap of pebbles of the same diameter is such a complex problem, so that even after research and development work extending over three decades, it cannot be regarded as completely solved. (orig.)

  7. Shielding pebble transfer system for thermonuclear device

    International Nuclear Information System (INIS)

    In a system for supplying shielding pebbles to a vacuum vessel filled with the shielding pebbles in a gap of a double-walled structure, a supply port for the shielding pebbles is formed in a diverging shape, and a corny object is disposed at the center of the flow channel, or protrusions are formed in the vicinity of the supply port. Alternatively, a small object is disposed at the center of the flow channel of the supply port, and the small object is supported swingably and tiltably by elastic members. In addition, the upper plate of the vacuum vessel is slanted having the supply port of the shielding pebbles as a top, and a slanting angle relative to a horizontal axis is made greater than the resting angle of the shielding pebble accumulation layer. The shielding pebbles are jetted out from the supply port and spread to the peripheries, abut against the inner surface of the vacuum vessel, jump up and then accumulate. Accordingly, they can be accumulated dispersingly without being localized. An uniform accumulation layer is obtained to form a vacuum vessel having uniform and high shielding performance. (N.H.)

  8. Development Status of the PEBBLES Code for Pebble Mechanics: Improved Physical Models and Speed-up

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2009-12-01

    PEBBLES is a code for simulating the motion of all the pebbles in a pebble bed reactor. Since pebble bed reactors are packed randomly and not precisely placed, the location of the fuel elements in the reactor is not deterministically known. Instead, when determining operating parameters the motion of the pebbles can be simulated and stochastic locations can be found. The PEBBLES code can output information relevant for other simulations of the pebble bed reactors such as the positions of the pebbles in the reactor, packing fraction change in an earthquake, and velocity profiles created by recirculation. The goal for this level three milestone was to speedup the PEBBLES code through implementation on massively parallel computer. Work on this goal has resulted in speeding up both the single processor version and creation of a new parallel version of PEBBLES. Both the single processor version and the parallel running capability of the PEBBLES code have improved since the fiscal year start. The hybrid MPI/OpenMP PEBBLES version was created this year to run on the increasingly common cluster hardware profile that combines nodes with multiple processors that share memory and a cluster of nodes that are networked together. The OpenMP portions use the Open Multi-Processing shared memory parallel processing model to split the task across processors in a single node that shares memory. The Message Passing Interface (MPI) portion uses messages to communicate between different nodes over a network. The following are wall clock speed up for simulating an NGNP-600 sized reactor. The single processor version runs 1.5 times faster compared to the single processor version at the beginning of the fiscal year. This speedup is primarily due to the improved static friction model described in the report. When running on 64 processors, the new MPI/OpenMP hybrid version has a wall clock speed up of 22 times compared to the current single processor version. When using 88 processors, a

  9. Development Status of the PEBBLES Code for Pebble Mechanics: Improved Physical Models and Speed-up

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2009-09-01

    PEBBLES is a code for simulating the motion of all the pebbles in a pebble bed reactor. Since pebble bed reactors are packed randomly and not precisely placed, the location of the fuel elements in the reactor is not deterministically known. Instead, when determining operating parameters the motion of the pebbles can be simulated and stochastic locations can be found. The PEBBLES code can output information relevant for other simulations of the pebble bed reactors such as the positions of the pebbles in the reactor, packing fraction change in an earthquake, and velocity profiles created by recirculation. The goal for this level three milestone was to speedup the PEBBLES code through implementation on massively parallel computer. Work on this goal has resulted in speeding up both the single processor version and creation of a new parallel version of PEBBLES. Both the single processor version and the parallel running capability of the PEBBLES code have improved since the fiscal year start. The hybrid MPI/OpenMP PEBBLES version was created this year to run on the increasingly common cluster hardware profile that combines nodes with multiple processors that share memory and a cluster of nodes that are networked together. The OpenMP portions use the Open Multi-Processing shared memory parallel processing model to split the task across processors in a single node that shares memory. The Message Passing Interface (MPI) portion uses messages to communicate between different nodes over a network. The following are wall clock speed up for simulating an NGNP-600 sized reactor. The single processor version runs 1.5 times faster compared to the single processor version at the beginning of the fiscal year. This speedup is primarily due to the improved static friction model described in the report. When running on 64 processors, the new MPI/OpenMP hybrid version has a wall clock speed up of 22 times compared to the current single processor version. When using 88 processors, a

  10. Development Status of the PEBBLES Code for Pebble Mechanics: Improved Physical Models and Speed-up

    International Nuclear Information System (INIS)

    PEBBLES is a code for simulating the motion of all the pebbles in a pebble bed reactor. Since pebble bed reactors are packed randomly and not precisely placed, the location of the fuel elements in the reactor is not deterministically known. Instead, when determining operating parameters the motion of the pebbles can be simulated and stochastic locations can be found. The PEBBLES code can output information relevant for other simulations of the pebble bed reactors such as the positions of the pebbles in the reactor, packing fraction change in an earthquake, and velocity profiles created by recirculation. The goal for this level three milestone was to speedup the PEBBLES code through implementation on massively parallel computer. Work on this goal has resulted in speeding up both the single processor version and creation of a new parallel version of PEBBLES. Both the single processor version and the parallel running capability of the PEBBLES code have improved since the fiscal year start. The hybrid MPI/OpenMP PEBBLES version was created this year to run on the increasingly common cluster hardware profile that combines nodes with multiple processors that share memory and a cluster of nodes that are networked together. The OpenMP portions use the Open Multi-Processing shared memory parallel processing model to split the task across processors in a single node that shares memory. The Message Passing Interface (MPI) portion uses messages to communicate between different nodes over a network. The following are wall clock speed up for simulating an NGNP-600 sized reactor. The single processor version runs 1.5 times faster compared to the single processor version at the beginning of the fiscal year. This speedup is primarily due to the improved static friction model described in the report. When running on 64 processors, the new MPI/OpenMP hybrid version has a wall clock speed up of 22 times compared to the current single processor version. When using 88 processors, a

  11. Remote fabrication of pellet fuels for United States breeder reactors

    International Nuclear Information System (INIS)

    Goal of the program is to demonstrate the feasibility of fabricating breeder fuel in a remotely operated and maintained mode by 1985. Development for pellet fuel fabrication is in the engineering stage with much of the equipment for ceramic unit operations in final design or currently under testing. Results to date confirm that remote fabrication of pellet fuels is feasible. Several of the processes and equipment items are described in this report

  12. 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.

  13. The effects of temperatures on the pebble flow in a pebble bed high temperature reactor

    International Nuclear Information System (INIS)

    The core of a pebble bed high temperature reactor (PBHTR) moves during operation, a feature which leads to better fuel economy (online refueling with no burnable poisons) and lower fuel stress. The pebbles are loaded at the top and trickle to the bottom of the core after which the burnup of each is measured. The pebbles that are not fully burned are recirculated through the core until the target burnup is achieved. The flow pattern of the pebbles through the core is of importance for core simulations because it couples the burnup distribution to the core temperature and power profiles, especially in cores with two or more radial burnup 'zones '. The pebble velocity profile is a strong function of the core geometry and the friction between the pebbles and the surrounding structures (other pebbles or graphite reflector blocks). The friction coefficient for graphite in a helium environment is inversely related to the temperature. The Thorium High Temperature Reactor (THTR) operated in Germany between 1983 and 1989. It featured a two-zone core, an inner core (IC) and outer core (OC), with different fuel mixtures loaded in each zone. The rate at which the IC was refueled relative to the OC in THTR was designed to be 0.56. During its operation, however, this ratio was measured to be 0.76, suggesting the pebbles in the inner core traveled faster than expected. It has been postulated that the positive feedback effect between inner core temperature, burnup, and pebble flow was underestimated in THTR. Because of the power shape, the center of the core in a typical cylindrical PBHTR operates at a higher temperature than the region next to the side reflector. The friction between pebbles in the IC is lower than that in the OC, perhaps causing a higher relative flow rate and lower average burnup, which in turn yield a higher local power density. Furthermore, the pebbles in the center region have higher velocities than the pebbles next to the side reflector due to the

  14. Li4SiO4 pebbles reduction in He + 0.1% H2 purge gas and effects on tritium release properties

    International Nuclear Information System (INIS)

    Lithium orthosilicate reduction was examined by Temperature Programmed Reaction (TPR) and Temperature Programmed Desorption (TPD) methods performed in He (or Ar) + H2 purge gas flowing through pebble bed specimens. The parameters governing the kinetics and the steady-state of the reduction process to Li4SiO4-x were determined at 800 deg. C. The level x of the O-vacancy concentration at steady-state (of the order of 1.5x10-3 mole fraction) was found to be compatible with the impurities content in the specimens. Pebble pre-annealing treatments were found to affect the microstructure and the reduction mechanism. Post-irradiation tritium release by TPD tests were performed on both stoichiometric and reduced pebbles with similar results. Tritium release properties of this breeder system seem to be independent from the material reduction state (x)

  15. Multiscale Analysis of Pebble Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar; Woo Yoon; Abderrafi Ougouag

    2010-10-01

    – The PEBBED code was developed at the Idaho National Laboratory for design and analysis of pebble-bed high temperature reactors. The diffusion-depletion-pebble-mixing algorithm of the original PEBBED code was enhanced through coupling with the THERMIX-KONVEK code for thermal fluid analysis and by the COMBINE code for online cross section generation. The COMBINE code solves the B-1 or B-3 approximations to the transport equation for neutron slowing down and resonance interactions in a homogeneous medium with simple corrections for shadowing and thermal self-shielding. The number densities of materials within specified regions of the core are averaged and transferred to COMBINE from PEBBED for updating during the burnup iteration. The simple treatment of self-shielding in previous versions of COMBINE led to inaccurate results for cross sections and unsatisfactory core performance calculations. A new version of COMBINE has been developed that treats all levels of heterogeneity using the 1D transport code ANISN. In a 3-stage calculation, slowing down is performed in 167 groups for each homogeneous subregion (kernel, particle layers, graphite shell, control rod absorber annulus, etc.) Particles in a local average pebble are homogenized using ANISN then passed to the next (pebble) stage. A 1D transport solution is again performed over the pebble geometry and the homogenized pebble cross sections are passed to a 1-d radial model of a wedge of the pebble bed core. This wedge may also include homogeneous reflector regions and a control rod region composed of annuli of different absorbing regions. Radial leakage effects are therefore captured with discrete ordinates transport while axial and azimuthal effects are captured with a transverse buckling term. In this paper, results of various PBR models will be compared with comparable models from literature. Performance of the code will be assessed.

  16. Experimental and computational investigation of flow of pebbles in a pebble bed nuclear reactor

    Science.gov (United States)

    Khane, Vaibhav B.

    The Pebble Bed Reactor (PBR) is a 4th generation nuclear reactor which is conceptually similar to moving bed reactors used in the chemical and petrochemical industries. In a PBR core, nuclear fuel in the form of pebbles moves slowly under the influence of gravity. Due to the dynamic nature of the core, a thorough understanding about slow and dense granular flow of pebbles is required from both a reactor safety and performance evaluation point of view. In this dissertation, a new integrated experimental and computational study of granular flow in a PBR has been performed. Continuous pebble re-circulation experimental set-up, mimicking flow of pebbles in a PBR, is designed and developed. Experimental investigation of the flow of pebbles in a mimicked test reactor was carried out for the first time using non-invasive radioactive particle tracking (RPT) and residence time distribution (RTD) techniques to measure the pebble trajectory, velocity, overall/zonal residence times, flow patterns etc. The tracer trajectory length and overall/zonal residence time is found to increase with change in pebble's initial seeding position from the center towards the wall of the test reactor. Overall and zonal average velocities of pebbles are found to decrease from the center towards the wall. Discrete element method (DEM) based simulations of test reactor geometry were also carried out using commercial code EDEM(TM) and simulation results were validated using the obtained benchmark experimental data. In addition, EDEM(TM) based parametric sensitivity study of interaction properties was carried out which suggests that static friction characteristics play an important role from a packed/pebble beds structural characterization point of view. To make the RPT technique viable for practical applications and to enhance its accuracy, a novel and dynamic technique for RPT calibration was designed and developed. Preliminary feasibility results suggest that it can be implemented as a non

  17. Li2TiO3 pebbles reprocessing, recovery of 6Li as Li2CO3

    International Nuclear Information System (INIS)

    A process for obtaining Li2CO3 from Li2TiO3 powder by wet chemistry was developed. This is considered useful in view of the recovery of 6Li isotope from a lithium titanate breeder burned up to its end of life in a fusion reactor. The process was optimized with respect to the chemical attack of titanate and the precipitation of carbonate from aqueous solutions to get a powder, with the chemical and morphological characteristics, suitable for its re-exploitation in the fabrication of Li2TiO3 pebbles. Reprocessing was also planned to adjust the 6Li concentration to the desired value and to obtain homogeneous distribution in the powder batch. Further development concerning reprocessing of sintered Li2TiO3 pebbles is in progress exploiting the results obtained with lithium titanate powders. (orig.)

  18. Li2TiO3 pebbles reprocessing, recovery of 6Li as Li2CO3

    International Nuclear Information System (INIS)

    A process for obtaining Li2CO3 from Li2TiO3 powder by wet chemistry was developed. This is considered useful in view of the recovery of the 6Li isotope from lithium titanate breeder burned to its end of life in a fusion reactor. The process was optimized with respect to the chemical attack of titanate and the precipitation of carbonate from aqueous solutions to get a powder with chemical and morphological characteristics suitable for its reexploitation in the fabrication of Li2TiO3 pebbles. Reprocessing was also planned to adjust the 6Li concentration to the desired value and to obtain a homogeneous distribution in the powder batch. Further development concerning reprocessing of sintered Li2TiO3 pebbles is in progress exploiting the results obtained with lithium titanate powders

  19. Formation and accumulation of radiation-induced defects and radiolysis products in modified lithium orthosilicate pebbles with additions of titanium dioxide

    Science.gov (United States)

    Zarins, Arturs; Valtenbergs, Oskars; Kizane, Gunta; Supe, Arnis; Knitter, Regina; Kolb, Matthias H. H.; Leys, Oliver; Baumane, Larisa; Conka, Davis

    2016-03-01

    Lithium orthosilicate (Li4SiO4) pebbles with 2.5 wt.% excess of silicon dioxide (SiO2) are the European Union's designated reference tritium breeding ceramics for the Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM). However, the latest irradiation experiments showed that the reference Li4SiO4 pebbles may crack and form fragments under operation conditions as expected in the HCPB TBM. Therefore, it has been suggested to change the chemical composition of the reference Li4SiO4 pebbles and to add titanium dioxide (TiO2), to obtain lithium metatitanate (Li2TiO3) as a second phase. The aim of this research was to investigate the formation and accumulation of radiation-induced defects (RD) and radiolysis products (RP) in the modified Li4SiO4 pebbles with different contents of TiO2 for the first time, in order to estimate and compare radiation stability. The reference and the modified Li4SiO4 pebbles were irradiated with accelerated electrons (E = 5 MeV) up to 5000 MGy absorbed dose at 300-990 K in a dry argon atmosphere. By using electron spin resonance (ESR) spectroscopy it was determined that in the modified Li4SiO4 pebbles, several paramagnetic RD and RP are formed and accumulated, like, E' centres (SiO33-/TiO33-), HC2 centres (SiO43-/TiO3-) etc. On the basis of the obtained results, it is concluded that the modified Li4SiO4 pebbles with TiO2 additions have comparable radiation stability with the reference pebbles.

  20. Lithium titanate pebbles reprocessing by wet chemistry

    International Nuclear Information System (INIS)

    An original dissolution method for irradiated Li2TiO3 in aqueous H2O2 was developed. One could easily obtain fine Li2TiO3 powders from the solution through drying and calcination. Li2TiO3 pebbles (size ∼0.6 mm, above 90% TD) were obtained from the 'reprocessed' powders. These solutions were also suitable for the formation of a sol emulsion in 2-ethyl-hexanol-1, from which gelled microspheres of lithium titanate could be obtained. Locally prepared Li2TiO3 reprocessed and supplied pebble batches were tested for tritium release by temperature programmed desorption (TPD) methods in He + 0.1%H2 (R-gas) after their short irradiations in a thermal neutron flux. The relative TPD data were compared. A qualitative correlation was developed between peak characteristics and pebble microstructure

  1. Fuel feeding with pebbles, installation and experience

    International Nuclear Information System (INIS)

    The AVR reactor is a graphite moderated high temperature reactor cooled with helium (at 10 bar). It was the first reactor which operated with spherical fuel elements. The fuel elements have a diameter of 60 mm. They consists of graphite in which the fuel is embedded as coated particles, and weigh about 200 g. The core is cylindrical, with a diameter of 3 m and 3 m high. About 100,000 pebbles are accommodated in it. The fuelling equipment differs considerably from the fuelling and fuel removal machines of previously operated reactors with rod-shaped or block-shaped fuel elements, because of the spherical fuel elements. The tasks of the fuelling equipment are the addition, turning over and removal of fuel elements and test pebbles. The burnup also has to be measured and all pebbles have to be counted and recorded. One can talk of a completely new development, as there were no tested components for all these tasks. (orig.)

  2. Numerical simulation on friction coefficient effect of pebble flow dynamics in two-dimensional pebble-bed reactor

    International Nuclear Information System (INIS)

    In order to investigate the pebble flow dynamics in the high-temperature reactor core and based on the two-dimensional experiments of pebble flow dynamics, discrete element method (DEM) was used to simulate the pebble flow dynamics. The mean flow stream lines, standard deviation and the mean residence time of the pebble flow zone generated by markers were compared and analyzed. The results show that ball friction coefficient has little effect on the pebble flow field. With the pebble friction coefficient increasing, the horizontal diffusion of pebbles decreases and the pebble flow seems to be more uniform. The wall friction coefficient has little effect on the horizontal diffusion. While the wall friction coefficient increases, the flow tends to be more uneven. (authors)

  3. 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.''

  4. International strategies for breeder development

    International Nuclear Information System (INIS)

    This paper studies the perspectives of breeder reactors development. The near term context has led some experts to the conclusion that breeder reactor technology is too far ahead of its time. Some have compared breeders to the supersonic airplane, Concorde: good technical performance but failure in its economic dimensions. In this paper, the author points out the major shortcomings of such an assessment which may be valid in the short time. However, with a short-term market-dominated perspective that uses an 8% discount rate, one can neglect every thing that is going to happen in 50 years. 6 refs., 11 figs

  5. Exploring new coolants for nuclear breeder reactors

    International Nuclear Information System (INIS)

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F2Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F2Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F2Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void reactivity coefficients

  6. Fabrication of Li4SiO4 pebbles by wet method with modified powders synthesized via sol–gel process

    International Nuclear Information System (INIS)

    Li4SiO4 pebbles have been recognized as attractive tritium breeder materials in the fusion reactor blanket of international thermonuclear experimental reactor (ITER). In this work, we present a facile method to prepare Li4SiO4 pebbles of high density and sphericity by using a directive wet method with the Li4SiO4 powders synthesized via sol–gel process. The Li4SiO4 powders were prepared with two-step calcinating method, followed by a ball-milling process. Thermal and phase analysis, morphologies and sintering behaviors observations of the pebbles were carried out systematically. Experimental results show that the pure phase powders with white color that prepared by using two-step calcinating method is different from the powders prepared by the traditional direct calcinating method. The subsequent ball milling process proves to be effective to improve the relative density of the sintered body. When sintered at the temperature as low as 850 °C for 4 h, the favorable Li4SiO4 pebbles with uniform size (∼1 mm), good sphericity (1.02), and high density (above 90% T.D.) were fabricated by using a directive wet method. The as-fabricated pebbles hold good potential as tritium breeding materials for blankets

  7. Construction of PREMUX and preliminary experimental results, as preparation for the HCPB breeder unit mock-up testing

    International Nuclear Information System (INIS)

    Highlights: • PREMUX has been constructed as preparation for a future out-of-pile thermo-mechanical qualification of a HCPB breeder unit mock-up. • The rationale and constructive details of PREMUX are reported in this paper. • PREMUX serves as a test rig for the new heater system developed for the HCPB-BU mock-up. • PREMUX will be used as benchmark for the thermal and thermo-mechanical models developed in ANSYS for the pebble beds of the HCPB-BU. • Preliminary results show the functionality of PREMUX and the good agreement of the measured temperatures with the thermal model developed in ANSYS. - Abstract: One of the European blanket designs for ITER is the Helium Cooled Pebble Bed (HCPB) blanket. The core of the HCPB-TBM consists of so-called breeder units (BUs), which encloses beryllium as neutron multiplier and lithium orthosilicate (Li4SiO4) as tritium breeder in form of pebble beds. After the design phase of the HCPB-BU, a non-nuclear thermal and thermo-mechanical qualification program for this device is running at the Karlsruhe Institute of Technology. Before the complex full scale BU testing, a pre-test mock-up experiment (PREMUX) has been constructed, which consists of a slice of the BU containing the Li4SiO4 pebble bed. PREMUX is going to be operated under highly ITER-relevant conditions and has the following goals: (1) as a testing rig of new heater concept based on a matrix of wire heaters, (2) as benchmark for the existing finite element method (FEM) codes used for the thermo-mechanical assessment of the Li4SiO4 pebble bed, and (3) in situ measurement of thermal conductivity of the Li4SiO4 pebble bed during the tests. This paper describes the construction of PREMUX, its rationale and the experimental campaign planned with the device. Preliminary results testing the algorithm used for the temperature reconstruction of the pebble bed are reported and compared qualitatively with first analyses completed with the FEM codes

  8. Granular Dynamics in Pebble Bed Reactor Cores

    Science.gov (United States)

    Laufer, Michael Robert

    This study focused on developing a better understanding of granular dynamics in pebble bed reactor cores through experimental work and computer simulations. The work completed includes analysis of pebble motion data from three scaled experiments based on the annular core of the Pebble Bed Fluoride Salt-Cooled High- Temperature Reactor (PB-FHR). The experiments are accompanied by the development of a new discrete element simulation code, GRECO, which is designed to offer a simple user interface and simplified two-dimensional system that can be used for iterative purposes in the preliminary phases of core design. The results of this study are focused on the PB-FHR, but can easily be extended for gas-cooled reactor designs. Experimental results are presented for three Pebble Recirculation Experiments (PREX). PREX 2 and 3.0 are conventional gravity-dominated granular systems based on the annular PB-FHR core design for a 900 MWth commercial prototype plant and a 16 MWth test reactor, respectively. Detailed results are presented for the pebble velocity field, mixing at the radial zone interfaces, and pebble residence times. A new Monte Carlo algorithm was developed to study the residence time distributions of pebbles in different radial zones. These dry experiments demonstrated the basic viability of radial pebble zoning in cores with diverging geometry before pebbles reach the active core. Results are also presented from PREX 3.1, a scaled facility that uses simulant materials to evaluate the impact of coupled fluid drag forces on the granular dynamics in the PB-FHR core. PREX 3.1 was used to collect first of a kind pebble motion data in a multidimensional porous media flow field. Pebble motion data were collected for a range of axial and cross fluid flow configurations where the drag forces range from half the buoyancy force up to ten times greater than the buoyancy force. Detailed analysis is presented for the pebble velocity field, mixing behavior, and residence time

  9. Thorium utilization in a pebble bed reactor

    International Nuclear Information System (INIS)

    Thorium reserves in the earth's crust are much more than those of uranium, which today measure about 1.5 million tonnes of reasonably assured resources, plus 3 million tonnes of estimated additional resources. These large amount of thorium reserves, also available in Turkey encourages to focus on the utilization of thorium. The most remarkable applications of the use of thorium have been in high temperature reactors. The high temperature pebble bed reactor, which has been chosen as the basis for this study, is a close approximation of the thorium utilizing German reactor THTR. Pebble bed reactors have some unique features which are suitable to burn thorium. (i) The fuel is loaded in the form of coated particles, which are embedded in the graphite matrix of the fuel pebbles, allowing exceptionally high heavy metal burnups; and (ii) the continuous (on-line) fuel loading allows a high utilization factor. The criticality search of the pebble bed reactor is computed by the use of the SCALE4.4 code, CSASIX and KENOVa modules. And the in-core fuel management is computed via SCALE4.4 code, ORIGEN-S module

  10. Effect of wall structure on pebble stagnation behavior in pebble bed reactor

    International Nuclear Information System (INIS)

    Highlights: • DEM study of wall structure role in preventing near wall crystallization is carried out. • Suggestions on pebble’s kinematic parameters and wall structure design are provided. • Triangle is better than arc and sawtooth shapes for wall structure design. • Wall structure size should be close to the scale of pebble diameter. • Suitable intervals can prevent crystallization without significantly increasing the flow resistance. - Abstract: Crystallization of pebbles in pebble bed is a crucial problem in high temperature gas-cooled pebble-bed reactors. This phenomenon usually happens along the internal surface and leads to a large number of stagnated pebbles, which poses a threat to reactor safety. In real reactor engineering, wall structures have been utilized to avoid this problem. This article verifies the crystallization phenomenon through DEM (discrete element method) simulation, and explains how wall structures work in preventing crystallization. Moreover, several kinematic parameters have been adopted to evaluate wall structures with different shapes, sizes and intervals. Detailed information shows the impact of wall structure on flow field in pebble bed. Lastly, the preferred characteristics of an effective wall structure are suggested for reactor engineering

  11. 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

  12. "Smart pebble" design for environmental monitoring applications

    Science.gov (United States)

    Valyrakis, Manousos; Pavlovskis, Edgars

    2014-05-01

    Sediment transport, due to primarily the action of water, wind and ice, is one of the most significant geomorphic processes responsible for shaping Earth's surface. It involves entrainment of sediment grains in rivers and estuaries due to the violently fluctuating hydrodynamic forces near the bed. Here an instrumented particle, namely a "smart pebble", is developed to investigate the exact flow conditions under which individual grains may be entrained from the surface of a gravel bed. This could lead in developing a better understanding of the processes involved, while focusing on the response of the particle during a variety of flow entrainment events. The "smart pebble" is a particle instrumented with MEMS sensors appropriate for capturing the hydrodynamic forces a coarse particle might experience during its entrainment from the river bed. A 3-axial gyroscope and accelerometer registers data to a memory card via a microcontroller, embedded in a 3D-printed waterproof hollow spherical particle. The instrumented board is appropriately fit and centred into the shell of the pebble, so as to achieve a nearly uniform distribution of the mass which could otherwise bias its motion. The "smart pebble" is powered by an independent power to ensure autonomy and sufficiently long periods of operation appropriate for deployment in the field. Post-processing and analysis of the acquired data is currently performed offline, using scientific programming software. The performance of the instrumented particle is validated, conducting a series of calibration experiments under well-controlled laboratory conditions. "Smart pebble" allows for a wider range of environmental sensors (e.g. for environmental/pollutant monitoring) to be incorporated so as to extend the range of its application, enabling accurate environmental monitoring which is required to ensure infrastructure resilience and preservation of ecological health.

  13. Conceptual Design Studies of a Passively Safe Thorium Breeder Pebble Bed Reactor

    OpenAIRE

    Wols, F.J.

    2015-01-01

    Nuclear power plants are expected to play an important role in the worldwide electricity production in the coming decades, since they provide an economically attractive, reliable and low-carbon source of electricity with plenty of resources available for at least the coming hundreds of years. However, the design of nuclear reactors can be improved significantly in terms of safety, by designing reactors with fully passive safety systems, and sustainability, by making more efficient use of natu...

  14. Conceptual Design Studies of a Passively Safe Thorium Breeder Pebble Bed Reactor

    NARCIS (Netherlands)

    Wols, F.J.

    2015-01-01

    Nuclear power plants are expected to play an important role in the worldwide electricity production in the coming decades, since they provide an economically attractive, reliable and low-carbon source of electricity with plenty of resources available for at least the coming hundreds of years. Howeve

  15. Pebble Accretion and the Diversity of Planetary Systems

    CERN Document Server

    Chambers, J E

    2016-01-01

    I examine the standard model of planet formation, including pebble accretion, using numerical simulations. Planetary embryos large enough to become giant planets do not form beyond the ice line within a typical disk lifetime unless icy pebbles stick at higher speeds than in experiments using rocky pebbles. Systems like the Solar System (small inner planets, giant outer planets) can form if (i) icy pebbles are stickier than rocky pebbles, and (ii) the planetesimal formation efficiency increases with pebble size, which prevents the formation of massive terrestrial planets. Growth beyond the ice line is dominated by pebble accretion. Most growth occurs early, when the surface density of pebbles is high due to inward drift of pebbles from the outer disk. Growth is much slower after the outer disk is depleted. The outcome is sensitive to the disk radius and turbulence level, which control the lifetime and maximum size of pebbles. The outcome is sensitive to the size of the largest planetesimals since there is a th...

  16. Transmutation of nuclear wastes with gas-cooled pebble-bed ads

    International Nuclear Information System (INIS)

    Transmutation of nuclear wastes is being explored for its application to waste management, a fundamental issue for nuclear industry. Several concepts are under consideration, mainly fast breeder reactors and accelerator driven systems (ADS). Inside this second category, we are analysing a helium-cooled graphite moderated sub-critical assembly, which uses as fuel units a small amount of transuranics diluted, in the form of TRISO coated particles, in graphite pebbles. This configuration (PBT) allows for neutron spectra that, taking advantage of the existence of huge capture resonances in the epithermal region, increase in a substantial factor the system transmutation efficiency. Neutronic studies to determine transmutation performance and thermal behaviour are presented and discussed together with an analysis of the additional studies to address before going into detailed design activities. (author)

  17. International cooperation on breeder reactors

    International Nuclear Information System (INIS)

    In March 1977, as the result of discussions which began in the fall of 1976, the Rockefeller Foundation requested International Energy Associates Limited (IEAL) to undertake a study of the role of international cooperation in the development and application of the breeder reactor. While there had been considerable international exchange in the development of breeder technology, the existence of at least seven major national breeder development programs raised a prima facie issue of the adequacy of international cooperation. The final product of the study was to be the identification of options for international cooperation which merited further consideration and which might become the subject of subsequent, more detailed analysis. During the course of the study, modifications in U.S. breeder policy led to an expansion of the analysis to embrace the pros and cons of the major breeder-related policy issues, as well as the respective views of national governments on those issues. The resulting examination of views and patterns of international collaboration emphasizes what was implicit from the outset: Options for international cooperation cannot be fashioned independently of national objectives, policies and programs. Moreover, while similarity of views can stimulate cooperation, this cannot of itself provide compelling justification for cooperative undertakings. Such undertakings are influenced by an array of other national factors, including technological development, industrial infrastructure, economic strength, existing international ties, and historic experience

  18. 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.

  19. 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

  20. A study on evaluation of pebble flow velocity with modification of the kinematic model for pebble bed reactor

    International Nuclear Information System (INIS)

    Highlights: ► A modified kinematic method is proposed for analysis of pebble flow velocity. ► Experiments are performed to derive the coefficients and to verify the results. ► The method and result can be used for the advanced analysis of pebble bed reactor. - Abstract: A pebble bed reactor is filled by a large number of pebbles, which are randomly piled up in the core region. During the process of fuel loading and extraction, the pebbles flow downward through the core. The basic physics of the dense granular flow such as pebble flow is not fully understood; hence, the dynamic core of the pebble bed reactor has been a subject of concern among designers and regulators. The kinematic model is one of the representative models for the reconstruction of the granular flow velocity, however, it is noted that there are some limitations in the reconstruction ability. In this study, a modified kinematic model was proposed to enhance the reconstruction ability of the pebble velocity profile. Pebble flow experiments were performed to derive the coefficients needed for the modified kinematic model and to verify the reconstruction ability and the applicability of the proposed method in the annular core. The modified kinematic model can contribute to accurate velocity evaluation as well as large applicability for the specific core types such as an annular core. Also, the results can be used for reference data in the design of a pebble bed reactor

  1. Mechanical performance of irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Dalle-Donne, M.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1998-01-01

    For the Helium Cooled Pebble Bed (HCPB) Blanket, which is one of the two reference concepts studied within the European Fusion Technology Programme, the neutron multiplier consists of a mixed bed of about 2 and 0.1-0.2 mm diameter beryllium pebbles. Beryllium has no structural function in the blanket, however microstructural and mechanical properties are important, as they might influence the material behavior under neutron irradiation. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating it. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from these irradiation experiments, emphasizing the effects of irradiation of essential material properties and trying to elucidate the processes controlling the property changes. The microstructure, the porosity distribution, the impurity content, the behavior under compression loads and the compatibility of the beryllium pebbles with lithium orthosilicate (Li{sub 4}SiO{sub 4}) during the in-pile irradiation are presented and critically discussed. Qualitative information on ductility and creep obtained by hardness-type measurements are also supplied. (author)

  2. Preliminary Neutronics Analysis Of Fuel Pebble With Thorium Fuel Cycle

    International Nuclear Information System (INIS)

    A new fuel pebble was designed based on Thorium fuel cycle. 231Pa has been added into fuel pebble for obtaining the minimum reactivity swing. The results show that the new designed pebble fuel with 7.0 % 233U enrichment adding 3.2% 231Pa, the keff is to be controlled up to 65 GWd/t; the other design with 8.0 % 233U enrichment requires 3.9% 231Pa, the keff therefore is remain up to 80 GWd/t. About 95% of loaded 231Pa in fuel pebble is depleted after 120 GWd/t. The results imply that it is optimistic to design the fuel pebble with 233U, 231Pa and 232Th; but some effects such as fuel temperature effect, distribution of TRISO particle in pebble fuel, etc. are required to investigate. (author)

  3. Fusion Breeder Program interim report

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.; Lee, J.D.; Neef, W.

    1982-06-11

    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.

  4. 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

  5. PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

    International Nuclear Information System (INIS)

    This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity. The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use

  6. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores. [PEBBLE code

    Energy Technology Data Exchange (ETDEWEB)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases.

  7. PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1981-09-01

    This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity. The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.

  8. 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)

  9. Pebble Delivery for Inside-Out Planet Formation

    CERN Document Server

    Hu, Xiao; Chatterjee, Sourav

    2014-01-01

    Inside-Out Planet Formation (IOPF; Chatterjee & Tan 2014, hereafter CT14) is a scenario for sequential in situ planet formation at the pressure traps of retreating dead zone inner boundaries (DZIBs) motivated to explain the many systems with tightly packed inner planets (STIPs) discovered by Kepler. The scenario involves build-up of a pebble-dominated protoplanetary ring, supplied by radial drift of pebbles from the outer disk. It may also involve further build-up of planetary masses to gap-opening scales via continued pebble accretion. Here we study radial drift & growth of pebbles delivered to the DZIB in fiducial IOPF disk models.

  10. Tritium analyses of COBRA-1A2 beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, D.L. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    Selected tritium measurements have been completed for the COBRA-1A2 experiment C03 and D03 beryllium pebbles. The completed results, shown in Tables 1, 2, and 3, include the tritium assay results for the 1-mm and 3-mm C03 pebbles, and the 1-mm D03 pebbles, stepped anneal test results for both types of 1-mm pebbles, and the residual analyses for the stepped-anneal specimens. All results have been reported with date-of-count and are not corrected for decay. Stepped-anneal tritium release response is provided in addenda.

  11. Phenomenological method investigation of pebble flow dynamics in two-dimensional two-region pebble-bed reactor

    International Nuclear Information System (INIS)

    By means of the four basic forms of the phenomenological method, the experimental research was carried out according to the principle of similarity criterion to simulate 2D pebble flow dynamics of high-temperature gas-cooled reactor. The result indicates that the test with circulating pebble-loading mode obviously presents better the situation in the real pebble bed reactor. The pebble flow dynamics spreads from bottom to top and from middle to sides. The movement of pebbles in central region is faster than that in annulus region and has no laminar characteristics performance. The mixed zone exists between central region and annulus region, and the distinct stagnant zone also exists at pebble bed bottom corner. (authors)

  12. Design and technology development of solid breeder blanket cooled by supercritical water in Japan

    International Nuclear Information System (INIS)

    successfully fabricated. It withstood the high heat flux test at 2.7 MW m-2. Also, a correlation parameter of the Li2TiO3 pebble bed made by the sol-gel method was verified by measurement of the thermal conductivity of the breeder pebble bed, which is one of the most important design data. (author)

  13. Numerical simulation on pebble dynamics of two-dimensional two-region pebble-bed reactor using phenomenological method

    International Nuclear Information System (INIS)

    Discrete element method was used to simulate the pebble dynamics in the high-temperature gas-cooled reactor core, based on the two-dimensional pebble dynamics experiments. Phenomenological method was used to analyze the formation of the two-region distribution, the central, mixing and stagnant regions, and the velocity distribution. The simulation results show that a stable central region is formed, mixing zones between the central and annular regions and stagnant regions are observed in current simulation. The closer to the bottom of the pebble bed, the more uneven the vertical pebble velocity and the bigger the horizontal diffusion. (authors)

  14. 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)

  15. A Pebble Bed Reactor cross section methodology

    International Nuclear Information System (INIS)

    A method is presented for the evaluation of microscopic cross sections for the Pebble Bed Reactor (PBR) neutron diffusion computational models during convergence to an equilibrium (asymptotic) fuel cycle. This method considers the isotopics within a core spectral zone and the leakages from such a zone as they arise during reactor operation. The randomness of the spatial distribution of fuel grains within the fuel pebbles and that of the fuel and moderator pebbles within the core, the double heterogeneity of the fuel, and the indeterminate burnup of the spectral zones all pose a unique challenge for the computation of the local microscopic cross sections. As prior knowledge of the equilibrium composition and leakage is not available, it is necessary to repeatedly re-compute the group constants with updated zone information. A method is presented to account for local spectral zone composition and leakage effects without resorting to frequent spectrum code calls. Fine group data are pre-computed for a range of isotopic states. Microscopic cross sections and zone nuclide number densities are used to construct fine group macroscopic cross sections, which, together with fission spectra, flux modulation factors, and zone buckling, are used in the solution of the slowing down balance to generate a new or updated spectrum. The microscopic cross-sections are then re-collapsed with the new spectrum for the local spectral zone. This technique is named the Spectral History Correction (SHC) method. It is found that this method accurately recalculates local broad group microscopic cross sections. Significant improvement in the core eigenvalue, flux, and power peaking factor is observed when the local cross sections are corrected for the effects of the spectral zone composition and leakage in two-dimensional PBR test problems.

  16. Survey of dust production in pebble bed reactor cores

    International Nuclear Information System (INIS)

    Highlights: → We review potential sources of the graphite dust found in the German pebble bed reactors. → Available literature on graphite wear coefficients in pebble bed core-like conditions is reviewed. → Limited conclusions and remaining open questions are discussed. - Abstract: Graphite dust produced via mechanical wear from the pebbles in a pebble bed reactor is an area of concern for licensing. Both the German pebble bed reactors produced graphite dust that contained activated elements. These activation products constitute an additional source term of radiation and must be taken under consideration during the conduct of accident analysis of the design. This paper discusses the available literature on graphite dust production and measurements in pebble bed reactors. Limited data is available on the graphite dust produced from the AVR and THTR-300 pebble bed reactors. Experiments that have been performed on wear of graphite in pebble-bed-like conditions are reviewed. The calculation of contact forces, which are a key driving mechanism for dust in the reactor, are also included. In addition, prior graphite dust predictions are examined, and future areas of research are identified.

  17. Neutronic modeling of pebble bed reactors in APOLLO2

    International Nuclear Information System (INIS)

    In this thesis we develop a new iterative homogenization technique for pebble bed reactors, based on a 'macro-stochastic' transport approximation in the collision probability method. A model has been developed to deal with the stochastic distribution of pebbles with different burnup in the core, considering spectral differences in homogenization and depletion calculations. This is generally not done in the codes presently used for pebble bed analyses, where a pebble with average isotopic composition is considered to perform the cell calculation. Also an iterative core calculation scheme has been set up, where the low-order RZ SN full-core calculation computes the entering currents in the spectrum zones subdividing the core. These currents, together with the core keff, are then used as surface source in the fine-group heterogeneous calculation of the multi-pebble geometries. The developed method has been verified using reference Monte Carlo simulations of a simplified PBMR- 400 model. The pebbles in this model are individually positioned and have different randomly assigned burnup values. The APOLLO2 developed method matches the reference core keff within ± 100 pcm, with relative differences on the production shape factors within ± 4%, and maximum discrepancy of 3% at the hotspot. Moreover, the first criticality experiment of the HTR-10 reactor was used to perform a first validation of the developed model. The computed critical number of pebbles to be loaded in the core is very close to the experimental value of 16890, only 77 pebbles less. A method to calculate the equilibrium reactor state was also developed and applied to analyze the simplified PBMR-400 model loaded with different fuel types (UO2, Pu, Pu + MA). The potential of the APOLLO2 method to compute different fluxes for the different pebble types of a multi-pebble geometry was used to evaluate the bias committed by the average composition pebble approximation. Thanks to a 'compensation of error', this

  18. 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.

  19. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

  20. Optimization of MOX fuel cycles in pebble bed HTGR

    International Nuclear Information System (INIS)

    Compared with light water reactor (LWR), the pebble bed high temperature gas-cooled reactor (HTGR) is able to operate in a full mixed oxide (MOX) fuelled core without significant change to core structure design. Based on a reference design of 250 MW pebble bed HTGR, four MOX fuel cycles were designed and evaluated by VSOP program package, including the mixed Pu-U fuel pebbles and mixed loading of separate Pu-pebbles and U-pebbles. Some important physics features were investigated and compared for these four cycles, such as the effective multiplication factor of initial core, the pebble residence time, discharge burnup, and temperature coefficients. Preliminary results show that the overall performance of one case is superior to other equivalent MOX fuel cycles on condition that uranium fuel elements and plutonium fuel elements are separated as the different fuel pebbles and that the uranium fuel elements are irradiated longer in the core than the plutonium fuel elements, and the average discharge burnup of this case is also higher than others. (authors)

  1. Regulation of the pebble flow in a pebble-bed reactor

    International Nuclear Information System (INIS)

    The cylindrical core tank of the pebble bed reactor has a funnel-shaped outlet at its bottom end with one single discharge tube. The cylindrical part of the pebble bed has a height-to-diameter ratio of 0.5. In order to achieve an approximately constant vertical velocity of the OTTO program in the Juelich AVR and the Uentrop THTR reactors, respectively, there is a small cone mounted on supports in the outlet. Its tip is pointed upward and its largest diameter has a distance to the wall of the funnel sufficient to accommodate the diameter of the spheres. At high powers the outlet and the cone are built up of graphite blocks. They are equipped with numerous vertical channels for coolant passage. The cone may also be a rotationally symmetrical body with a rhombic longitudinal section which rests upon the outlet on radial bars. (orig./PW)

  2. Uncertainty and sensitivity analysis of filling fraction of pebble bed in pebble bed HTR

    International Nuclear Information System (INIS)

    Highlights: • An analysis approach is proposed to conduct SAU analysis of filling fraction of pebble bed core. • The contribution of uncertainty in filling fraction to key parameters of pebble bed core is quantified. • The primary drivers of the uncertainty in the key parameters are identified by sensitivity analysis. • Mechanism of effect of uncertainty in the filling fraction to key parameters is studied in depth. - Abstract: The filling fraction of pebble bed in each small region has some uncertainty, which will contribute to the total uncertainty in the key parameters of pebble bed core, such as power peak, axial offset, keff and so on. In fact, the heavy metal content and graphite content of the corresponding region will change synchronously due to the perturbation of filling fraction but the ratio of atomic number density of moderator to the fuel (carbon–uranium) is still constant. To investigate these effects, the Chinese demonstration plant HTR-PM was selected as the research object and the VSOP code and CUSA package were used to conduct detailed analysis of the influence of the uncertainty in the filling fraction on the HTR-PM output variables of interest, based on the propagation of input uncertainties by using statistical sampling method to calculate uncertainty and sensitivity information from the simulation results. At the same time, uncertainty and sensitivity analysis of uranium loading had also been conducted for comparative analysis to study the mechanism of effect of uncertainty in the filling fraction to key parameters, and therefore only the heavy metal content of the corresponding region changes in the presence of perturbation of uranium loading. Finally, the propagated uncertainty in the power peak, axial offset and keff of HTR-PM core was obtained and the primary drivers of the uncertainty in the key parameters were identified by sensitivity analysis

  3. 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

  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. 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)

  6. Calibration of a pebble bed configuration for direct numerical simulation

    International Nuclear Information System (INIS)

    The appearance of hot spots in the pebble bed cores of High Temperature Reactors (HTR) may affect the integrity of the pebbles. A good prediction of the flow and heat transport in such a pebble bed core is a challenge for available turbulence models. Such models need to be validated in order to gain trust in the simulation of these types of flow configurations. Direct Numerical Simulation (DNS) can serve as a reference for validation, however, it poses restrictions in terms of flow parameters and numerical tools corresponding to the available computational resources. In the present study, a wide range of numerical simulations has been performed in order to calibrate a pebble bed configuration for DNS which may serve as reference for validation. (author)

  7. Building massive compact planetesimal disks from the accretion of pebbles

    CERN Document Server

    Moriarty, John

    2015-01-01

    We present a model in which planetesimal disks are built from the combination of planetesimal formation and accretion of radially drifting pebbles onto existing planetesimals. In this model, the rate of accretion of pebbles onto planetesimals quickly outpaces the rate of direct planetesimal formation in the inner disk. This allows for the formation of a high mass inner disk without the need for enhanced planetesimal formation or a massive protoplanetary disk. Our proposed mechanism for planetesimal disk growth does not require any special conditions to operate. Consequently, we expect that high mass planetesimal disks form naturally in nearly all systems. The extent of this growth is controlled by the total mass in pebbles that drifts through the inner disk. Anything that reduces the rate or duration of pebble delivery will correspondingly reduce the final mass of the planetesimal disk. Therefore, we expect that low mass stars (with less massive protoplanetary disks), low metallicity stars and stars with gian...

  8. Pebble bed reactor with a feeding device for absorber materials

    International Nuclear Information System (INIS)

    Description of a second shutdown device for pebble bed reactors with small absorber pebbles, which, if required, can be let off a storage tank and regularly trickle through a dispersion cone into the pebble bed. In the normal state the storage tank is in a low position with its outlet being obstructed by absorber pebbles filling a cylinder in which slides a piston which is firmly connected with the storage tank. By giving pressure over a line a piston in a pneumatic cylinder can be moved which lifts the storage tank. The cylinder is emptied by lifting the piston and the outlet is released. The level of the storage tank is measured by means of a probe. The whole device is installed in the prestressed concrets ceiling of the reactor. The device is proposed to be set into motion for a short moment from time to time in order to prove its operatability. (orig.)

  9. On the growth of pebble-accreting planetesimals

    CERN Document Server

    Visser, Rico G

    2015-01-01

    Pebble accretion is a new mechanism to quickly grow the cores of planets. In pebble accretion, gravity and gas drag conspire to yield large collisional cross sections for small particles in protoplanetary disks. However, before pebble accretion commences, aerodynamical deflection may act to prevent planetesimals from becoming large, because particles tend to follow gas streamlines. We derive the planetesimal radius where pebble accretion is initiated and determine the growth timescales of planetesimals by sweepup of small particles. We obtain the collision efficiency factor as the ratio of the numerically-obtained collisional cross section to the planetesimal surface area, from which we obtain the growth timescales. Integrations are conducted in the potential flow limit (steady, inviscid) and in the Stokes flow regime (steady, viscid). Only particles of stopping time $t_s \\ll t_X$ where $t_X\\approx10^3$ s experience aerodynamic deflection. Even in that case, the planetesimal's gravity always ensures positive ...

  10. US/FRG joint report on the pebble bed high temperature reactor resource conservation potential and associated fuel cycle costs

    International Nuclear Information System (INIS)

    Independent analyses at ORNL and KFA have led to the general conclusion that the flexibility in design and operation of a high-temperature gas-cooled pebble-bed reactor (PBR) can result in favorable ore utilization and fuel costs in comparison with other reactor types, in particular, with light-water reactors (LWRs). Fuel reprocessign and recycle show considerable promise for reducing ore consumption, and even the PBR throwaway cycle is competitive with fuel recycle in an LWR. The best performance results from the use of highly enriched fuel. Proliferation-resistant measures can be taken using medium-enriched fuel at a modest ore penalty, while use of low-enriched fuel would incur further ore penalty. Breeding is possible but net generation of fuel at a significant rate would be expensive, becoming more feasible as ore costs increase substantially. The 233U inventory for a breeder could be produced by prebreeders using 235U fuel

  11. Control rod for a pebble bed nuclear reactor

    International Nuclear Information System (INIS)

    In order to leave the density of the pebble bed unchanged when the control rod is driven in and out, the tip of the control rod is provided with moving parts in the form of conical hemispheres or spoons. These parts move close to the control rod when it is driven in and spread out due to the effect of gravity when it is driven out. This loosens the heap of pebble shaped operating elements again. (DG)

  12. Reconstructing the transport history of pebbles on Mars

    OpenAIRE

    Szabó, Tímea; Domokos, Gábor; Grotzinger, John P.; Jerolmack, Douglas J.

    2015-01-01

    The discovery of remarkably rounded pebbles by the rover Curiosity, within an exhumed alluvial fan complex in Gale Crater, presents some of the most compelling evidence yet for sustained fluvial activity on Mars. While rounding is known to result from abrasion by inter-particle collisions, geologic interpretations of sediment shape have been qualitative. Here we show how quantitative information on the transport distance of river pebbles can be extracted from their shape alone, using a combin...

  13. Steam chemical reactivity of Be pebbles and Be powder

    International Nuclear Information System (INIS)

    This paper reports the results of chemical reactivity experiments for Be pebbles (2 and 0.2 mm diameter) and Be powder (14-31 μm diameter) exposed to steam at elevated temperatures, 350-900 deg. C for pebbles and 400-500 deg. C for powders. We measured BET specific surface areas of 0.12 m2/g for 2 mm pebbles, 0.24 m2/g for 0.2 mm pebbles and 0.66-1.21 m2/g for Be powder samples. These experiments showed a complex reactivity behavior for the material, dependent primarily on the test temperature. Average H2 generation rates for powder samples, based on measured BET surface areas, were in good agreement with previous measurements for fully dense consolidated powder metallurgy (CPM)-Be. Rates for the Be pebbles, based on measured BET surface areas, were systematically lower than the CPM-Be rates, possibly because of different surface and bulk features for the pebbles, especially surface layer impurities, that contribute to the measured BET surface area and influence the oxidation process at the material surface

  14. Recovery and recycling of lithium value from spent lithium titanate (Li2TiO3) pebbles

    Science.gov (United States)

    Mandal, D.

    2013-09-01

    In the first generation fusion reactors the fusion of deuterium (D) and tritium (T) is considered to produce energy to meet the future energy demand. Deuterium is available in nature whereas, tritium is not. Lithium-6 (Li6) isotope has the ability to produce tritium in the n, α nuclear reaction with neutrons. Thus lithium-based ceramics enriched by Li6 isotope are considered for the tritium generation for its use in future fusion reactors. Lithium titanate is one such Li-based ceramic material being considered for its some attractive properties viz., high thermal and chemical stability, high thermal conductivity, and low tritium solubility. It is reported in the literature, that the burn up of these pebbles in the fusion reactor will be limited to only 15-17 atomic percentage. At the end of life, the pebbles will contain more than 45% unused Li6 isotope. Due to the high cost of enriched Li6 and the waste disposal considerations, it is necessary to recover the unused Li from the spent lithium titanate pebbles. Till date, only the feasibilities of different processes are reported, but no process details are available. Experiments were carried out for the recovery of Li from simulated Li2TiO3 pebbles and to reuse of lithium in lithium titanate pebble fabrication. The details of the experiments and results are discussed in this paper. Simulated lithium titanate (Li2TiO3) pebbles. The objective of the study is to develop a process which can be used to recover lithium value form the spent Li2TiO3 pebbles from future fusion reactor. The Li2TiO3 pebbles used in the study were synthesized and fabricated by the solid state reaction process developed by Mandal et al. described in details somewhere else [1,2]. Spherical Li2TiO3 pebbles of size 1.0 mm were used and the properties of the Li2TiO3 pebbles used in the study are shown in Table 1. Hydrochloric acid (HCl), of 99.8% purity, purchased from Merck and Loba Chemicals, Mumbai, India. To leach lithium from Li2TiO3

  15. Pebbles and Branching Programs for Tree Evaluation

    CERN Document Server

    Cook, Stephen; Wehr, Dustin; Braverman, Mark; Santhanam, Rahul

    2010-01-01

    We introduce the Tree Evaluation Problem, show that it is in logDCFL (and hence in P), and study its branching program complexity in the hope of eventually proving a superlogarithmic space lower bound. The input to the problem is a rooted, balanced d-ary tree of height h, whose internal nodes are labeled with d-ary functions on [k] = {1,...,k}, and whose leaves are labeled with elements of [k]. Each node obtains a value in [k] equal to its d-ary function applied to the values of its d children. The output is the value of the root. We show that the standard black pebbling algorithm applied to the binary tree of height h yields a deterministic k-way branching program with Theta(k^h) states solving this problem, and we prove that this upper bound is tight for h=2 and h=3. We introduce a simple semantic restriction called "thrifty" on k-way branching programs solving tree evaluation problems and show that the same state bound of Theta(k^h) is tight (up to a constant factor) for all h >= 2 for deterministic thrift...

  16. Status of the pebble bed modular reactor

    International Nuclear Information System (INIS)

    Eskom is the South African state electricity utility, with an installed capacity of 38397 MW at the end of 1996 (some 98% of all national generating assets). It is largely coal-based with a small proportion (5%) nuclear. As part of Eskom's long-term planning process, investigations have been made into new power generation options. On reconsidering the nuclear option, Eskom identified two key issues: cost and public acceptance. It was considered that both of these were driven by the safety issues related to potential accidents and the only way to obtain competitive costs with nuclear power was to remove the potential (however remote) for accidents with significant off-site consequences. The only reactor type that was seen to meet this safety standard was the pebble bed modular reactor (PBMR). This paper discusses the PBMR project history, plant performance and design, its benefits, safety features, and current status. It concludes that the PBMR will provide South Africa with a competitive option for coastal generation and, internationally, it will be highly competitive with virtually all other generation options. (author)

  17. Pebble red modular reactor - South Africa

    International Nuclear Information System (INIS)

    In 1995 the South African Electricity Utility, ESKOM, was convinced of the economical advantages of high temperature gas-cooled reactors as viable supply side option. Subsequently planning of a techno/economic study for the year 1996 was initiated. Continuation to the construction phase of a prototype plant will depend entirely on the outcome of this study. A reactor plant of pebble bed design coupled with a direct helium cycle is perceived. The electrical output is limited to about 100 MW for reasons of safety, economics and flexibility. Design of the reactor will be based on internationally proven, available technology. An extended research and development program is not anticipated. New licensing rules and regulations will be required. Safety classification of components will be based on the merit of HTGR technology rather than attempting to adhere to traditional LWR rules. A medium term time schedule for the design and construction of a prototype plant, commissioning and performance testing is proposed during the years 2002 and 2003. Pending the performance outcome of this plant and the current power demand, series production of 100 MWe units is foreseen. (author)

  18. An economic analysis of fusion breeders

    International Nuclear Information System (INIS)

    This paper presents a study of the economic performance of Fission/Fusion Hybrid devices. This work takes fusion breeder cost estimates and applies methodology and cost factors used in the fission reactor programs to compare fusion breeders with Liquid Metal Fast Breeder Reactors (LMFBR). The results of the analysis indicate that the Hybrid will be in the same competitive range as proposed LMFBRs and have the potential to provide economically competitive power in a future of rising uranium prices. The sensitivity of the results to variations in key parameters is included

  19. A method for estimating maximum static rainfall retention in pebble mulches used for soil moisture conservation

    Science.gov (United States)

    Peng, Hongtao; Lei, Tingwu; Jiang, Zhiyun; Horton, Robert

    2016-06-01

    Mulching of agricultural fields and gardens with pebbles has long been practiced to conserve soil moisture in some semi-arid regions with low precipitation. Rainfall interception by the pebble mulch itself is an important part of the computation of the water balance for the pebble mulched fields and gardens. The mean equivalent diameter (MED) was used to characterize the pebble size. The maximum static rainfall retention in pebble mulch is based on the water penetrating into the pores of pebbles, the water adhering to the outside surfaces of pebbles and the water held between pebbles of the mulch. Equations describing the water penetrating into the pores of pebbles and the water adhering to the outside surface of pebbles are constructed based on the physical properties of water and the pebble characteristics. The model for the water between pebbles of the mulch is based on the basic equation to calculate the water bridge volume and the basic coordination number model. A method to calculate the maximum static rainfall retention in the pebble mulch is presented. Laboratory rain simulation experiments were performed to test the model with measured data. Paired sample t-tests showed no significant differences between the values calculated with the method and the measured data. The model is ready for testing on field mulches.

  20. Stepped-anneal helium release in 1-mm beryllium pebbles from COBRA-1A2

    Energy Technology Data Exchange (ETDEWEB)

    Oliver, B.M. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    Stepped-anneal helium release measurements on two sets of fifteen beryllium pebbles irradiated in the Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-w), are reported. The purpose of the measurements was to determine the helium release characteristics of the beryllium using larger sample sizes and longer anneal times relative to earlier measurements. Sequential helium analyses were conducted over a narrower temperature range from approximately 800 C to 1100 C in 100 C increments, but with longer anneal time periods. To allow for overnight and unattended operation, a temperature controller and associated circuitry were added to the experimental setup. Observed helium release was nonlinear with time at each temperature interval, with each step being generally characterized by an initial release rate followed by a slowing of the rate over time. Sample Be-C03 showed a leveling off in the helium release after approximately 3 hours at a temperature of 890 C. Sample Be-D03, on the other hand, showed a leveling off only after {approximately}12 to 24 hours at a temperature of 1100 C. This trend is consistent with that observed in earlier measurements on single microspheres from the same two beryllium lots. None of the lower temperature steps showed any leveling off of the helium release. Relative to the total helium concentrations measured earlier, the total helium releases observed here represent approximately 80% and 92% of the estimated total helium in the C03 and D03 samples, respectively.

  1. Neutronic features of pebble-bed reactors for transmutation applications

    International Nuclear Information System (INIS)

    Pebble-bed reactors offer very appealing characteristics for radioactivity confinement and for withstanding thermal transients. Besides that, pebble-bed reactors have a peculiar degree of freedom in the radius of the active core of the pebble (where the fuel is located) as compared to the outer radius of the pebble, which has a coating of pure graphite. By varying the aforementioned radius, very different types of neutron spectra can be formed, which in turn gives very different values of the average cross sections that govern the isotopic composition evolution, and particularly the elimination of the most relevant transuranics. Preliminary conclusions of this work show that there is a very broad design window for exploiting the transmutation capabilities of pebble-bed reactors in a scenario of inherent safety features. A 99,9% elimination of Pu-239 associated to a 99% elimination of Pu-240 and Pu-241 can be reached, with some increment of the Pu-242 contents (which is extremely long-lived, less radio-toxic and decays into the natural nuclide U-238). Am and Cm are also transmuted to a significant level, although some residual higher A actinides will remain. (authors)

  2. 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

  3. Design and analysis of breeding blanket with helium cooled solid breeder for ITER-TBM

    International Nuclear Information System (INIS)

    Test blanket module (TBM) is one of important components in ITER. Some of related blanket technologies of future fusion, such as tritium self-sufficiency, the exaction of high-grade heat, design criteria and safety requirements and environmental impacts, will be demonstrated in ITER-TBM. In ITER device, the three equatorial ports have allocated for TBM testing. China had proposed to develop independently the ITER-TBM with helium cooled solid breeder in 12th meeting of test blanket workgroup (TBWG-12). In this work, the preliminary design and analysis for Chinese HCSB TBM will be carried out. The TBM must be contains the function of the first wall, breeding blanket, shield and structure. Finally, in the period of preliminary investigation, HCSB TBM design adopt modularization concept which is helium as coolant and tritium purge gas, ferritic/martensitic steel as structural material, Lithium orthosilicate (Li4SiO4) as tritium breeder, beryllium pebble as neutron multiplier. TBM is allocated in standard vertical frame port. HCSB TBM consist of first wall, backplate, breeding sub-modules, caps, grid and support plate, and breeding sub-modules is arranged by layout of 2 x 6 in blanket box. In this paper, main components of HCSB TBM will be described in detail, also performance analysis of main components have been completed. (authors)

  4. A European proposal for an ITER water-cooled solid breeder blanket

    International Nuclear Information System (INIS)

    The water-cooled solid breeder blanket concept proposed here aims to replace the shielding blanket for the enhanced performance phase of the international thermonuclear experimental reactor (ITER). The nominal performances are as follows: an average neutron wall load of 1 MW m-2 which corresponds to a fusion power of about 1.5 GW, and an average neutron fluence of 1 MWy m-2. The proposed blanket concept has been designed to accept a power increase of about 30% and power transients up to 3-5 GW for a short time. This blanket concept is based on a breeder inside tube (BIT)-type blanket with poloidal breeding elements made of 316 L-type stainless steel and filled with lithium metazirconate and beryllium pebbles. Inlet and outlet water temperatures of 160 and 200 C have been considered with a medium-pressure cooling system during plasma burn. The diameters of the breeding elements are compatible with the space available in test fission reactor core channels, making in-pile testing, required for blanket development and qualification, easier. A conservative approach using qualified materials, a blanket concept easily testable in fission reactors and on-going mock-up testing, which can be qualified using the blanket test module during the basic performance phase of ITER, will allow the blanket reliability required for the enhanced performance phase to be achieved. (orig.)

  5. Conceptual study of ferromagnetic pebbles for heat exhaust in fusion reactors with short power decay length

    OpenAIRE

    Gierse, N.; Coenen, J W; C. Thomser; Panin, A.; Ch. Linsmeier; Unterberg, B.; Philipps, V

    2015-01-01

    Ferromagnetic pebbles are investigated as high heat flux (q∥) plasma facing components in fusion devices with short power decay length (λq) on a conceptual level. The ability of a pebble concept to cope with high heat fluxes is retained and extended by the acceleration of ferromagnetic pebbles in magnetic fields. An alloying concept suited for fusion application is outlined and the compatibility of ferromagnetic pebbles with plasma operation is discussed. Steel grade 1.4510 is chosen as a ...

  6. Conceptual study of ferromagnetic pebbles for heat exhaust in fusion reactors with short power decay length

    Directory of Open Access Journals (Sweden)

    N. Gierse

    2015-03-01

    The key results of this study are that very high heat fluxes are accessible in the operation space of ferromagnetic pebbles, that ferromagnetic pebbles are compatible with tokamak operation and current divertor designs, that the heat removal capability of ferromagnetic pebbles increases as λq decreases and, finally, that for fusion relevant values of q∥ pebble diameters below 100 μm are required.

  7. Computational prediction of dust production in pebble bed reactors

    International Nuclear Information System (INIS)

    Highlights: ► Finite element analysis of frictional contact. ► Plasticity taken into account for nuclear graphite at room temperature. ► Prediction of order of magnitude for dust loading in PBRs. ► Archard wear model for wear mass calculations. - Abstract: This paper describes the computational modeling and simulation of graphite pebbles in frictional contacts as anticipated in a pebble bed reactor. For the high temperature gas-cooled reactor, the potential dust generation from frictional contact at the surface of pebbles and the subsequent lift-off and transport of dust and absorbed fission products are of safety concern at elevated temperatures under an air ingress accident. The aim of this work is to perform a computational study to estimate the quantity of the nuclear grade graphite dust produces from a typical anticipated configuration.

  8. In-core fuel management for pebble-bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Milian Perez, Daniel; Rodriguez Garcia, Lorena; Garcia Hernandez, Carlos; Milian Lorenzo, Daniel, E-mail: dperez@instec.cu, E-mail: cgh@instec.cu, E-mail: dmilian@instec.cu [Higher Institute of Technologies and Applied Sciences, Havana (Cuba); Velasco, Abanades, E-mail: abanades@etsii.upm.es [Department of Simulation of Thermo Energy Systems, Polytechnic University of Madrid (Spain)

    2013-07-01

    In this paper a calculation procedure to reduce the power peak in the core of a Very High Temperature pebble bed Reactor is presented. This procedure combines the fuel depletion and the neutronic behavior of the fuel in the reactor core, modeling once-through-then-out cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times, obtaining the asymptotic fuel-loading pattern. The procedure consists in several coupled computational codes, which are used iteratively until convergence is reached. The utilization of the MCNPX 2.6e, as one of these computational codes, is validated through the calculation of benchmarks announced by IAEA (IAEA-TECDOC-1249, 2001). To complete the verification of the calculation procedure a base case described in (Annals of Nuclear Energy 29 (2002) 1345-1364), was performed. The procedure has been applied to a model of Pebble Bed Modular Reactor (200 MW) design. (author)

  9. Fabrication of beryllide pebble as advanced neutron multiplier

    International Nuclear Information System (INIS)

    Highlights: • A new beryllide granulation process that combined process with a plasma sintering method for electrode fabrication and a rotating electrode method (REM) for granulation was suggested. • The beryllide electrode fabrication process was investigated for mass production. • As optimized beryllide electrode indicated higher ductility and was sintered at a lower temperature for a shorter time. • It appears to be more able to not only withstand the thermal shock from arc-discharge during granulation but also produce beryllide pebbles on a large scale. • These optimization results can reduce the time for electrode fabrication by 40%, they suggest the possibility of great reductions in time and cost for mass production of beryllide pebbles. - Abstract: Fusion reactors require advanced neutron multipliers with great stability at high temperatures. Beryllium intermetallic compounds, called beryllides such as Be12Ti, are the most promising materials for use as advanced neutron multipliers. However, few studies have been conducted on the development of mass production methods for beryllide pebbles. A granulation process for beryllide needs to have both low cost and high efficiency. To fabricate beryllide pebbles, a new granulation process is established in this research by combining a plasma sintering method for beryllide synthesis and a rotating electrode method using a plasma-sintered electrode for granulation. The fabrication process of the beryllide electrode is investigated and optimized for mass production. The optimized beryllide electrode exhibits higher ductility and can be sintered at a lower temperature for a shorter time, indicating that it is more suitable not only for withstanding the thermal shock from arc-discharge during granulation but also for producing the beryllide pebbles on a large scale. Accordingly, because these optimization results can reduce the time required for electrode fabrication by 40%, they suggest the possibility of

  10. BEATRIX-II: A multi-national solid breeder experiment

    International Nuclear Information System (INIS)

    BEATRIX-II is an IEA program focused on tritium recovery experiments on lithium ceramic materials in a fast neutron reactor which partially simulates the environment of a fusion blanket. In addition to data on the performance of Li2O and Li2ZrO3, the BEATRIX-II program offers information on innovative technologies associated with tritium recovery. The successful execution of the BEATRIX-II program also offers a precedent for the structure, schedule and interfaces that other international programs should consider. Japan, Canada, and USA re participants in the BEATRIX-II program with primary responsibilities being assigned to Japan Atomic Research Institute, Atomic Energy of Canada Ltd., Battelle Pacific Northwest Laboratory, and Westinghouse Hanford Company. The purpose of the BEATRIX-II experiment is to conduct in situ tritium recovery experiments on ceramic solid breeder materials under irradiation conditions which expanded the burnup, irradiation damage, tritium production, and temperature regimes previously investigated. A liquid metal, fast neutron reactor was selected because spatial variations in tritium and heat production are minimized and temporal variations in the lithium burnup rate (burnout) are also minimal. The Fast Flux Test Facility was selected because it possessed a high neutron flux, excellent control and monitoring capabilities and ready access for a tritium recovery experiment

  11. PEBBED ANALYSIS OF HOT SPOTS IN PEBBLE-BED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Abderrafi M. Ougouag; Hans D. Gougar; William K. Terry; Frederik Reitsma; Wessel Joubert

    2005-09-01

    The Idaho National Laboratory’s PEBBED code and simple probability considerations are used to estimate the likelihood and consequences of the accumulation of highly reactive pebbles in the region of peak power in a pebble-bed reactor. The PEBBED code is briefly described, and the logic of the probability calculations is presented in detail. The results of the calculations appear to show that hot-spot formation produces only moderate increases in peak accident temperatures, and no increases at all in normal operating temperatures.

  12. TEM study of impurity segregations in beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Klimenkov, M., E-mail: michael.klimenkov@kit.edu [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Chakin, V.; Moeslang, A. [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R. [Institute for Applied Materials – Materials and Biomechanics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2014-12-15

    Beryllium is planned to be used as a neutron multiplier in the Helium-cooled Pebble Bed European concept of a breeding blanket of demonstration power reactor DEMO. In order to evaluate the irradiation performance, individual pebbles and constrained pebble beds were neutron-irradiated at temperatures typical of fusion blankets. Beryllium pebbles 1 mm in diameter produced by the rotating electrode method were subjected to a TEM study before and after irradiation at High Flux Reactor, Petten, Netherlands at 861 K. The grain size varied in a wide range from sub-micron size up to several tens of micrometers, which indicated formation bimodal grain size distribution. Based on the application of combined electron energy loss spectroscopy and energy dispersive X-ray spectroscopy methods, we suggest that impurity precipitates play an important role in controlling the mechanical properties of beryllium. The impurity elements were present in beryllium at a sub-percent concentration form beryllide particles of a complex (Fe/Al/Mn/Cr)B composition. These particles are often ordered along dislocations lines, forming several micron-long chains. It can be suggested that fracture surfaces often extended along these chains in irradiated material.

  13. Transmutation of plutonium in pebble bed type high temperature reactors

    International Nuclear Information System (INIS)

    The pebble bed type High Temperature Reactor (HTR) has been studied as a uranium-free burner of reactor grade plutonium. In a parametric study, the plutonium loading per pebble as well as the type and size of the coated particles (CPs) have been varied to determine the plutonium consumption, the final plutonium burnup, the k∞ and the temperature coefficients as a function of burnup. The plutonium loading per pebble is bounded between 1 and 3 gr Pu per pebble. The upper limit is imposed by the maximal allowable fast fluence for the CPs. A higher plutonium loading requires a longer irradiation time to reach a desired burnup, so that the CPs are exposed to a higher fast fluence. The lower limit is determined by the temperature coefficients, which become less negative with increasing moderator-actinide ratio. A burnup of about 600 MWd/kgHM can be reached. With the HTR's high efficiency of 40%, a plutonium supply of 1520 kg/GWea is achieved. The discharges of plutonium and minor actinides are then 450 and 110 kg/GWea, respectively. (author)

  14. TEM study of impurity segregations in beryllium pebbles

    Science.gov (United States)

    Klimenkov, M.; Chakin, V.; Moeslang, A.; Rolli, R.

    2014-12-01

    Beryllium is planned to be used as a neutron multiplier in the Helium-cooled Pebble Bed European concept of a breeding blanket of demonstration power reactor DEMO. In order to evaluate the irradiation performance, individual pebbles and constrained pebble beds were neutron-irradiated at temperatures typical of fusion blankets. Beryllium pebbles 1 mm in diameter produced by the rotating electrode method were subjected to a TEM study before and after irradiation at High Flux Reactor, Petten, Netherlands at 861 K. The grain size varied in a wide range from sub-micron size up to several tens of micrometers, which indicated formation bimodal grain size distribution. Based on the application of combined electron energy loss spectroscopy and energy dispersive X-ray spectroscopy methods, we suggest that impurity precipitates play an important role in controlling the mechanical properties of beryllium. The impurity elements were present in beryllium at a sub-percent concentration form beryllide particles of a complex (Fe/Al/Mn/Cr)B composition. These particles are often ordered along dislocations lines, forming several micron-long chains. It can be suggested that fracture surfaces often extended along these chains in irradiated material.

  15. Pebble Bed Reactor review update. Fiscal year 1979 annual report

    International Nuclear Information System (INIS)

    Updated information is presented on the Pebble Bed Reactor (PBR) concept being developed in the Federal Republic of Germany for electricity generation and process heat applications. Information is presented concerning nuclear analysis and core performance, fuel cycle evaluation, reactor internals, and safety and availability

  16. Pebble Bed Reactor review update. Fiscal year 1979 annual report

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    Updated information is presented on the Pebble Bed Reactor (PBR) concept being developed in the Federal Republic of Germany for electricity generation and process heat applications. Information is presented concerning nuclear analysis and core performance, fuel cycle evaluation, reactor internals, and safety and availability.

  17. Fabrication of Li2TiO3 pebbles by a freeze drying process

    International Nuclear Information System (INIS)

    Li2TiO3 pebbles were successfully fabricated by using a freeze drying process. The Li2TiO3 slurry was prepared using a commercial powder of particle size 0.5–1.5 μm and the pebble pre-form was prepared by dropping the slurry into liquid nitrogen through a syringe needle. The droplets were rapidly frozen, changing their morphology to spherical pebbles. The frozen pebbles were dried at −10 °C in vacuum. To make crack-free pebbles, some glycerin was employed in the slurry, and long drying time and a low vacuum condition were applied in the freeze drying process. In the process, the solid content in the slurry influenced the spheroidicity of the pebble green body. The dried pebbles were sintered at 1200 °C in an air atmosphere. The sintered pebbles showed almost 40% shrinkage. The sintered pebbles revealed a porous microstructure with a uniform pore distribution and the sintered pebbles were crushed under an average load of 50 N in a compressive strength test. In the present study, a freeze drying process for fabrication of spherical Li2TiO3 pebbles is introduced. The processing parameters, such as solid content in the slurry and the conditions of freeze drying and sintering, are also examined

  18. Tritium trapping in silicon carbide in contact with solid breeder under high flux isotope reactor irradiation

    International Nuclear Information System (INIS)

    The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C β-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ∼3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation

  19. Canadian fusion breeder blanket program: Irradiation facilities at chalk river*1

    Science.gov (United States)

    Hastings, I. J.; Burton, D. G.; Celli, A.; Delaney, R. D.; Fehrenbach, P. J.; Howe, L. M.; Larson, L. L.; MacEwen, S. R.; Miller, J. M.; Naeem, T. A.; Sawicki, J. A.; Swanson, M. L.; Verrall, R. A.; Zee, R. H.

    1986-11-01

    The major irradiation facility at Chalk River Nuclear Laboratories (CRNL) is the NRU research reactor. Both unvented and vented capsule experiments on candidate blanket ceramics can be performed. In the unvented tests, tritium release data (HT-to-HTO ratio, tritium retention) are obtained by post-irradiation heating of the breeder ceramic in the presence of a sweep gas. Four tests have been completed on Li 2O and LiAlO 2. Effects of sweep gas composition, extraction vessel material and ceramic properties have been determined. Two unvented irradiations under the BEATRIX international breeder exchange program have been completed; analysis is underway. The vented tests involve long-term irradiation of candidate blanket materials. CRITIC-I, scheduled for mid-1986 under BEATRIX, will examine ANL-fabricated Li 2O in a six-month irradiation at 700-1200 K, varying sweep gas composition, with on-line HT/HTO measurement. Additionally, accelerator simulation techniques are available, using 70 kV and 2.0 MV mass separators, a 2.5 MV Van de Graaff accelerator and a tandem accelerator super-conducting cyclotron, the latter allowing irradiation with protons, deuterons or helium at 18-20 MeV.

  20. 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.

  1. 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.

  2. Tritium transport in lithium ceramics porous media

    International Nuclear Information System (INIS)

    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

  3. 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)

  4. Pore Scale Thermal Hydraulics Investigations of Molten Salt Cooled Pebble Bed High Temperature Reactor with BCC and FCC Configurations

    Directory of Open Access Journals (Sweden)

    Shixiong Song

    2014-01-01

    CFD results and empirical correlations’ predictions of pressure drop and local Nusselt numbers. Local pebble surface temperature distributions in several default conditions are investigated. Thermal removal capacities of molten salt are confirmed in the case of nominal condition; the pebble surface temperature under the condition of local power distortion shows the tolerance of pebble in extreme neutron dose exposure. The numerical experiments of local pebble insufficient cooling indicate that in the molten salt cooled pebble bed reactor, the pebble surface temperature is not very sensitive to loss of partial coolant. The methods and results of this paper would be useful for optimum designs and safety analysis of molten salt cooled pebble bed reactors.

  5. Modeling of realistic pebble bed reactor geometries using the Serpent Monte Carlo code

    International Nuclear Information System (INIS)

    Highlights: • The explicit stochastic geometry model in Serpent is documented. • A pebble bed criticality benchmark was calculated demonstrating the geometry model. • Stochastic pebble configurations were obtained from discrete element simulations. • Results deviate from experiments but are in line with example calculations. - Abstract: This paper documents the models available in Serpent for high temperature reactor (HTR) calculations. It is supplemented by a calculation example of ASTRA critical pebble bed experiments. In the pebble bed reactor modeling, different methods have been used to model the double heterogeneity problem occurring in pebble bed reactor calculations. A solution was sought to avoid unphysical simplifications in the pebble bed modeling and the stochastic geometry modeling features available in the Monte Carlo code Serpent were applied for exact placement of pebbles and fuel particles. Randomly packed pebble beds were produced in discrete element method (DEM) simulations and fuel particles were positioned randomly inside the pebbles. Pebbles and particles are located using a Cartesian search mesh, which provides necessary computational efficiency. Serpent uses Woodcock delta-tracking which provides efficient neutron tracking in the complicated geometries. This detailed pebble bed modeling approach was tested by calculating the ASTRA criticality benchmark experiment done at the Kurchatov Institute in 2004. The calculation results are in line with the sample calculations provided with the benchmark documentation. The material library selected for the calculations has a major effect on the results. The difference in graphite absorption cross section is considered the cause of this result. The model added in Serpent is very efficient with a calculation time slightly higher than with a regular lattice approximation. It is demonstrated that Serpent can be used for pebble bed reactor calculations with minimal geometric approximations as it

  6. 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

  7. 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

  8. The role and problems of the breeder

    International Nuclear Information System (INIS)

    World uranium resources are discussed and it is concluded that the period of availability of uranium for use in the present type of nuclear power stations is not much greater than that of oil. The neutron economies of fast and thermal reactors are compared, and the advantages of the breeder for the world uranium economy are demonstrated. The main impediments to the use of the fast breeder are considerations of safety, public acceptance and economics. Fast reactor safety is discussed and the health hazards and possible mis-use of plutonium for terrorism and weapons proliferation are considered. It is widely accepted that the U.K. cannot economically justify the development of the breeder alone and is likely to choose to co-operate with Western Europe. A public enquiry in the U.K. seems certain and would be welcomed by the nuclear industry. (author)

  9. Ceramic materials for fission and fusion nuclear reactors

    International Nuclear Information System (INIS)

    A general survey on the ceramics for nuclear applications is presented. For the fission nuclear reactor, the ceramics materials are almost totally used as fuel e.g. (U,Pu)O2; other types of ceramics, e.g. Uranium-Plutonium carbide and nitride, have been investigated as potential nuclear fuels. The (U,Pu)N compound is to be the fuel for the space nuclear power reactor in the U.S.A. For the fusion nuclear reactor, the ceramics should be the fundamental materials for many components: first wall, breeder, RF heating systems, insulant and shielding parts, etc. In recent years many countries are involved on the research and development of ceramic compounds with the principal purpose of being used in the fusion powerplant (year 2010-2020 ?). An effort has been even made to verify if it is possible to use more ceramic components in the fission nuclear plant (probably differntly disigned) to improve the safety level

  10. 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,

  11. 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

  12. Comparison analysis of fusion breeder blanket concepts

    International Nuclear Information System (INIS)

    Based on the wide survey, the development status and key issues of fusion breeder blanket concepts are summarized. Two types of blanket concepts, i.e. solid and liquid breeder blanket, were compared and assessed in terms of engineering feasibility, tritium recovery and control, economic and safety aspects, etc. The advantages and disadvantages of the two types of blanket concepts were clarified from the viewpoint of technology realization and development potential. This study may act as a valuable reference for fusion blanket concept selection and design. (authors)

  13. Structural Ceramics

    Science.gov (United States)

    1986-01-01

    This publication is a compilation of abstracts and slides of papers presented at the NASA Lewis Structural Ceramics Workshop. Collectively, these papers depict the scope of NASA Lewis' structural ceramics program. The technical areas include monolithic SiC and Si3N4 development, ceramic matrix composites, tribology, design methodology, nondestructive evaluation (NDE), fracture mechanics, and corrosion.

  14. 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.)

  15. 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

  16. Tritium release from neutron irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik

    1998-01-01

    One of the most important open issues related to beryllium for fusion applications refers to the kinetics of the tritium release as a function of neutron fluence and temperature. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating the beryllium response under neutron irradiation. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from the above mentioned irradiation experiments, trying to elucidate the tritium release controlling processes. In agreement with previous studies it has been found that release starts at about 500-550degC and achieves a maximum at about 700-750degC. The observed release at about 500-550degC is probably due to tritium escaping from chemical traps, while the maximum release at about 700-750degC is due to tritium escaping from physical traps. The consequences of a direct contact between beryllium and ceramics during irradiation, causing tritium implanting in a surface layer of beryllium up to a depth of about 40 mm and leading to an additional inventory which is usually several times larger than the neutron-produced one, are also presented and the effects on the tritium release are discussed. (author)

  17. 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

  18. Set-up of a pre-test mock-up experiment in preparation for the HCPB Breeder Unit mock-up experimental campaign

    International Nuclear Information System (INIS)

    Highlights: ► As preparation for the HCPB-TBM Breeder Unit out-of-pile testing campaign, a pre-test experiment (PREMUX) has been prepared and described. ► A new heater system based on a wire heater matrix has been developed for imitating the neutronic volumetric heating and it is compared with the conventional plate heaters. ► The test section is described and preliminary thermal results with the available models are presented and are to be benchmarked with PREMUX. ► The PREMUX integration in the air cooling loop L-STAR/LL in the Karlsruhe Institute for Technology is shown and future steps are discussed. -- Abstract: The complexity of the experimental set-up for testing a full-scaled Breeder Unit (BU) mock-up for the European Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) has motivated to build a pre-test mock-up experiment (PREMUX) consisting of a slice of the BU in the Li4SiO4 region. This pre-test aims at verifying the feasibility of the methods to be used for the subsequent testing of the full-scaled BU mock-up. Key parameters needed for the modeling of the breeder material is also to be determined by the Hot Wire Method (HWM). The modeling tools for the thermo-mechanics of the pebble beds and for the mock-up structure are to be calibrated and validated as well. This paper presents the setting-up of PREMUX in the L-STAR/LL facility at the Karlsruhe Institute of Technology. A key requirement of the experiments is to mimic the neutronic volumetric heating. A new heater concept is discussed and compared to several conventional heater configurations with respect to the estimated temperature distribution in the pebble beds. The design and integration of the thermocouple system in the heater matrix and pebble beds is also described, as well as other key aspects of the mock-up (dimensions, layout, cooling system, purge gas line, boundary conditions and integration in the test facility). The adequacy of these methods for the full-scaled BU mock-up is

  19. Status of the EXOTIC-8 programme and first in-pile results for Li{sub 2}TiO{sub 3} pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Van der Laan, J.G.; Stijkel, M.P. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Conrad, R.

    1998-03-01

    After renewal of the Tritium Measuring Station the HFR is again fully operational for in-pile breeder irradiations. The EXOTIC-8 series has started with first three experiments on June 12, 1997. First in-pile results have been obtained for Li{sub 2}TiO{sub 3}-pebbles supplied by CEA: preliminary analyses indicate satisfactory in-pile behaviour with fast recovery from transient conditions. Five further experiments have been defined which implies that in the present planning EXOTIC-8 is filled completely up to Fall`98 and 2 of 4 positions are occupied up to Spring`99. P.I.E. results will be obtained from Spring`98 onwards. (J.P.N.)

  20. 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)

  1. 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)

  2. Thorium and plutonium utilisation in pebble-bed modular reactor

    International Nuclear Information System (INIS)

    Thorium and plutonium utilisation in a high temperature gas-cooled pebble-bed reactor is investigated with the aim to predict the economic value of vast thorium reserves in Turkey. A pebble-bed reactor of the type designed by PBMR Pty. of South Africa is taken as the investigated system. The equilibrium core of a PBMR is considered and neutronics analyses of such a core are performed through the use of the SCALE-4.4 computer code system KENOV.a module. Various cross-section libraries are used to calculate the criticality of the core. Burn-up calculations of the core are performed by coupling the KENOV.a module with the ORIGEN-S module. Calculations are carried out for various U-Th, U-Pu-Th and U-Pu combinations. The results are preliminary in nature and the work is currently proceeding as planned. (author)

  3. Uranium assessment for the Precambrian pebble conglomerates in southeastern Wyoming

    International Nuclear Information System (INIS)

    This volume is a geostatistical resource estimate of uranium and thorium in quartz-pebble conglomerates, and is a companion to Volume 1: The Geology and Uranium Potential to Precambrian Conglomerates in the Medicine Bow Mountains and Sierra Madre of Southeastern Wyoming; and to Volume 2: Drill-Hole Data, Drill-Site Geology, and Geochemical Data from the Study of Precambrian Uraniferous Conglomerates of the Medicine Bow Mountains and the Sierra Madre of Southeastern Wyoming

  4. Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

    International Nuclear Information System (INIS)

    Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the 235U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of keff with MCNP5 and ENDF/B-VII.0 neutron nuclear data are greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of keff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

  5. Modeling of laminar forced convection in spherical- pebble packed beds

    International Nuclear Information System (INIS)

    There are many parameters that have significant effects on forced convection heat transfer in packed beds, including Reynolds and Prandtl numbers of flow, porosity, pebble geometry, local flow conditions, wall and end effects. In addition, there have been many experimental investigations on forced convection heat transfer in packed beds and each have studied the effect of some of these parameters. Yet, there is not a reliable correlation that includes the effect of main parameters: at the same time, the prediction of precise correct limits for very low and high Reynolds numbers is off hand. In this article a general well-known model of convection heat transfer from isothermal bodies, next to some previous reliable experimental data has been used as a basis for a more comprehensive and accurate correlation to calculate the laminar constant temperature pebble-fluid forced convection heat transfer in a homogeneous saturated bed with spherical pebbles. Finally, for corroboration, the present results are compared with previous works and show a very good agreement for laminar flows at any Prandtl number and all porosities

  6. Pebbling and Branching Programs Solving the Tree Evaluation Problem

    CERN Document Server

    Wehr, Dustin

    2010-01-01

    We study restricted computation models related to the Tree Evaluation Problem}. The TEP was introduced in earlier work as a simple candidate for the (*very*) long term goal of separating L and LogDCFL. The input to the problem is a rooted, balanced binary tree of height h, whose internal nodes are labeled with binary functions on [k] = {1,...,k} (each given simply as a list of k^2 elements of [k]), and whose leaves are labeled with elements of [k]. Each node obtains a value in [k] equal to its binary function applied to the values of its children, and the output is the value of the root. The first restricted computation model, called Fractional Pebbling, is a generalization of the black/white pebbling game on graphs, and arises in a natural way from the search for good upper bounds on the size of nondeterministic branching programs (BPs) solving the TEP - for any fixed h, if the binary tree of height h has fractional pebbling cost at most p, then there are nondeterministic BPs of size O(k^p) solving the heigh...

  7. Studies on air ingress for pebble bed reactors

    International Nuclear Information System (INIS)

    A loss-of-coolant accident (LOCA) has been considered a critical event for helium-cooled pebbled bed reactors. Following helium depressurization, it is anticipated that unless countermeasures are taken air will enter the core through the break and then by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure and graphite pebbles. Thus, without any mitigating features a LOCA will lead to an air ingress event. The INEEL is studying such an event with two well-respected light water reactor transient response codes: RELAP5/ATHENA and MELCOR. To study the degree of graphite oxidation occurring due to an air ingress event, a MELCOR model of a reference pebble bed design was constructed. A modified version of MELCOR developed at INEEL, which includes graphite oxidation capabilities, and molecular diffusion of air into helium was used for these calculations. Results show that the lower reflector graphite consumes all of the oxygen before reaching the core. The results also show a long time delay between the time that the depressurization phase of the accident is over and the time that natural circulation air through the core occurs. (author)

  8. Tritium release from highly neutron irradiated constrained and unconstrained beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V., E-mail: vladimir.chakin@kit.edu; Rolli, R.; Vladimirov, P.; Moeslang, A.

    2015-06-15

    Highlights: • For the irradiated constrained beryllium pebbles, the tritium release occurs easier than for the unconstrained ones. • Tritium retention in the irradiated constrained and unconstrained beryllium pebbles decreases with increasing irradiation temperature. • Formation of sub-grains in the constrained beryllium pebbles facilitate the open porosity network formation. - Abstract: Beryllium is the reference neutron multiplier material in the Helium Cooled Pebble Bed (HCPB) breeding blanket of fusion power plants. Significant tritium inventory accumulated in beryllium as a result of neutron-induced transmutations could become a safety issue for the operation of such blankets as well as for the nuclear waste utilization. To provide a related materials database, a neutron irradiation campaign of beryllium pebbles with diameters of 0.5 and 1 mm at 686–1006 K, the HIDOBE-01 experiment, has been performed in the HFR in Petten, the Netherlands, producing up to 3020 appm helium and 298 appm tritium. Thermal desorption tests of irradiated unconstrained and constrained beryllium pebbles were performed in a purge gas flow using a quadrupole mass-spectrometer (QMS) and an ionization chamber. Compared to unconstrained pebbles, constrained beryllium pebbles have an enhanced tritium release at all temperatures investigated. Small elongated sub-grains formed under irradiation in the constrained pebbles promote formation of numerous channels for facilitated tritium release.

  9. Tritium release from highly neutron irradiated constrained and unconstrained beryllium pebbles

    International Nuclear Information System (INIS)

    Highlights: • For the irradiated constrained beryllium pebbles, the tritium release occurs easier than for the unconstrained ones. • Tritium retention in the irradiated constrained and unconstrained beryllium pebbles decreases with increasing irradiation temperature. • Formation of sub-grains in the constrained beryllium pebbles facilitate the open porosity network formation. - Abstract: Beryllium is the reference neutron multiplier material in the Helium Cooled Pebble Bed (HCPB) breeding blanket of fusion power plants. Significant tritium inventory accumulated in beryllium as a result of neutron-induced transmutations could become a safety issue for the operation of such blankets as well as for the nuclear waste utilization. To provide a related materials database, a neutron irradiation campaign of beryllium pebbles with diameters of 0.5 and 1 mm at 686–1006 K, the HIDOBE-01 experiment, has been performed in the HFR in Petten, the Netherlands, producing up to 3020 appm helium and 298 appm tritium. Thermal desorption tests of irradiated unconstrained and constrained beryllium pebbles were performed in a purge gas flow using a quadrupole mass-spectrometer (QMS) and an ionization chamber. Compared to unconstrained pebbles, constrained beryllium pebbles have an enhanced tritium release at all temperatures investigated. Small elongated sub-grains formed under irradiation in the constrained pebbles promote formation of numerous channels for facilitated tritium release

  10. Evaluation of the Accumulation of Trace Metals (as, U, CR, CU, PB, Zn) on Iron-Manganese Coatings on in Situ Stream Pebbles and Emplaced Substrates

    Science.gov (United States)

    Turpin, M. M.; Blake, J.; Crossey, L. J.; Ali, A.; Hansson, L.

    2015-12-01

    Exposure to trace metals (As, U, Cr, Cu, Pb, Zn) has potential negative health effects on human populations and wildlife. Geothermal waters often have elevated concentrations of trace elements and understanding the geochemical cycling of these elements can be challenging. Previous studies have utilized in situ stream pebbles and glass or ceramic substrates with iron-manganese oxide coatings to understand contamination and or chemical cycling. This project's main focus is to develop an ideal tracing method using adsorption onto substrate surfaces and to define key parameters that are necessary for the phenomenon of adsorption between trace metals and these surface coatings to occur. Sampling locations include the Jemez River and Rio San Antonio in the Jemez mountains, northern New Mexico. Both streams have significant geothermal inputs. Pebbles and cobbles were gathered from the active stream channel and 6mm glass beads and 2 X1 in. ceramic plates were placed in streams for three weeks to allow for coating accumulation. Factors such as leachate type, water pH, substrate type, coating accumulation period and leach time were all considered in this experiment. It was found that of the three leachates (aqua regia, 10% aqua regia and hydroxylamine), hydroxylamine was the most effective at leaching coatings without dissolving substrates. Samples leached with aqua regia and 10% aqua regia were found to lose weight and mass over the following 5, 7, and 10 day measurements. Glass beads were determined to be more effective than in stream pebbles as an accumulation substrate: coatings were more easily controlled and monitored. Samples leached with hydroxylamine for 5 hours and 72 hours showed little difference in their leachate concentrations, suggesting that leach time has little impact on the concentration of leachate samples. This research aims to find the best method for trace metal accumulation in streams to aid in understanding geochemical cycling.

  11. A COMPARISON OF PEBBLE MIXING AND DEPLETION ALGORITHMS USED IN PEBBLE-BED REACTOR EQUILIBRIUM CYCLE SIMULATION

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; Frederik Reitsma; Wessel Joubert

    2009-05-01

    Recirculating pebble-bed reactors are distinguished from all other reactor types by the downward movement through and reinsertion of fuel into the core during operation. Core simulators must account for this movement and mixing in order to capture the physics of the equilibrium cycle core. VSOP and PEBBED are two codes used to perform such simulations, but they do so using different methods. In this study, a simplified pebble-bed core with a specified flux profile and cross sections is used as the model for conducting analyses of two types of burnup schemes. The differences between the codes are described and related to the differences observed in the nuclide densities in pebbles discharged from the core. Differences in the methods for computing fission product buildup and average number densities lead to significant differences in the computed core power and eigenvalue. These test models provide a key component of an overall equilibrium cycle benchmark involving neutron transport, cross section generation, and fuel circulation.

  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. Alternate fuel cycles for fast breeder reactors

    International Nuclear Information System (INIS)

    In this contribution to the syllabus for Subgroup 5D, a full range of alternate breeder fuel cycle options is developed and explored as to energy supply capability, resource utilizations, performance characteristics and technical features that pertain to proliferation resistance. Breeding performance information is presented for designs based on Pu/U, Pu/Th, 233 U/U, etc. with oxide, carbide or metal fuel; with lesser emphasis, heterogeneous and homogeneous concepts are presented. A potential proliferation resistance advantage of a symbiotic system of a Pu/U core, Th blanket breeder producing 233 U for utilization in dispersed LWR's is identified. LWR support ratios for various reactor and fuel types and the increase in uranium consumption with higher support ratios are identified

  14. What can fast breeders do for Ontario

    International Nuclear Information System (INIS)

    Fast reactors have the potential of significantly reducing Ontario's demand for natural resources while meeting virtually any requirements for nuclear power this province may have. The breeding efficiency of the fast reactors does not affect the overall uranium consumption of the system to any significant extent. It is, however, an important economic factor in a breeder/burner system. To minimize the resource consumption, the fast reactors should be introduced in Ontario at the onset of the next century. The 'breeder-burner' mix of reactors can effectively reduce the fissile inventory of the whole power system (including the inventory in irradiated fuel storage bays). For the nuclear capacity growth scenarios thought to be applicable in Ontario, the fast reactor systems have about the same or lower requirements for natural uranium as the best (self-sustaining thorium) CANDU cycles. Compared to all other advanced CANDU cycles, the fast reactors yield a substantial resource saving. (auth)

  15. BEATRIX: The international breeder materials exchange

    International Nuclear Information System (INIS)

    The BEATRIX experiment is an IEA-sponsored effort that involves the exchange of solid breeder materials and shared irradiation testing among research groups in several countries. The materials will be tested in both closed capsules (to evaluate material lifetime) and opened capsules (to evaluate purge-flow tritium recovery). Pre- and post-irradiation measurement of thermophysical and mechanical properties will also be carried out

  16. Technological questions of the breeder fuel cycle

    International Nuclear Information System (INIS)

    Since the contributions by the Karlsruhe Nuclear Research Center to the construction of SNR 300 have been completed to a large extent and irradiated KNK II fuel subassemblies have now become available, the possibility and necessity arise of concentrating efforts on the breeder fuel cycle. This work was started in 1980. The 17 papers presented at this seminar will provide a survey of intermediate results obtained until today. (orig./HP)

  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. The fast breeder reactor Rapsodie (1962)

    International Nuclear Information System (INIS)

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors)

  19. Fast Breeder Development: EDF's point of view

    International Nuclear Information System (INIS)

    This paper presents EDF's views and contributions to fast breeder development and to the French SFR trilateral program. Utility requirements are first outlined, based on the approach followed for the EPR reactor. R and D contributions are presented in the areas of core physics, safety, technology innovations, materials, deployment and fuel cycle scenarios. The paper also deals with some of the issues of the 2020 French prototype as seen by EDF.

  20. 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.

  1. The breeder's broke - who will save it

    International Nuclear Information System (INIS)

    In February, Bonn must decide on whether to tear down the most expensive building of the country: The Fast Breeder at Kalkar, which once seemed to lead to a glorious future. DM 3.500 million have been spent on it already and DM 1.000 million more will be needed. But the state has no money. The report is given by Wolfgang Hoffmann and Horst Bieber. (orig.)

  2. Future designs of breeder reactors (Europe, USA)

    International Nuclear Information System (INIS)

    Sodium-cooled reactors with a fast neutron core today are the only fission reactors that offer the reactor physics required for the breeding process and the complete conversion of U-238 or Th-232 into fissile fuel. There are currently five prototype breeder reactors in operation in England, France, and the USSR. The trends observable in development work aim at reducing capital cost, enhancing and improving passive shutdown performance, and simplifying the fuel cycle. (orig.)

  3. 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.

  4. 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

  5. Analysis of the HCPB breeder blanket bock-up experiment for ITER using SUSD3D code

    International Nuclear Information System (INIS)

    In order to validate new nuclear cross-section evaluations, method development and design of the helium-cooled pebble bed (HCPB) test blanket module of ITER a benchmark experiment was performed this year at the Frascati Neutron Generator (FNG) in the scope of the EFF (European Fusion File) project in Europe. The objective of this experiment is to study the tritium breeding ratio and other nuclear quantities in a breeder blanket in order to establish and improve the quality of related JEFF nuclear data. The experiment consists of a metallic beryllium set-up with two double layers of breeder material (Li2CO3 powder). The reaction rate measurements include the Li2CO3 pellets (tritium breeding ratio), activation foils, and neutron and gamma spectrometers inserted at several axial and lateral locations in the block. Our task is to perform the deterministic transport, and cross section sensitivity and uncertainty analysis. The role of the cross-section sensitivity and uncertainty analysis is to optimise the design of the benchmark, and to assist in the interpretation of the measurement results. The paper presents the pre- and post- analysis of the benchmark experiment. (author)

  6. History and evolution of the breeder reactor

    International Nuclear Information System (INIS)

    The concept of the breeder reactor is almost as old as the idea of the nuclear reactor itself. From the very first years following the discovery of nuclear fission, scientists and technicians tried to turn mankind's eternal dream into reality; that is, enjoy an abundant source of energy without using up our raw material reserves. Nuclear energy offered several solutions to realize this dream. One of them, fusion, seemed out of our grasp in the near future. But fission of 235U was possible, and the Manhattan Project soon furnished ample proof of this theory. However, everyone working in this field was conscious of the fact that thermal neutron reactors make very inefficient use of the energy potential contained in natural uranium. The solution was to use in a core sufficiently rich in fissile matter, the excess neutrons to convert the 238U, so poorly used by other types of reactors, into fissile 239Pu. Regeneration, or 'breeding' of fuel, can multiply the energy drawn from a ton of uranium by a factor of 50 to 100. This would enable us to ward off the specter of an energy shortage and the rapid depletion of uranium mines. As early as 1945 in Los Alamos, Enrico Fermi stated: 'The country which first develops a breeder reactor will have a great competitive edge in atomic energy.' The development of the breeder reactor in the USA and around the world is discussed

  7. Neutronic simulation of a pebble bed reactor considering its double heterogeneous nature

    International Nuclear Information System (INIS)

    Highlights: ► A new model is successfully developed for a pebble bed reactor simulation. ► In the model, the double heterogeneous nature is considered using MCNP5 code. ► The initial and full core criticality, control rod worth, etc. are calculated to validate it. ► Results confirm the capability of Monte Carlo codes in modeling complex geometries. - Abstract: In pebble bed reactors, the core is filled with thousands of graphite and fuel pebbles. Fuel pebbles in these reactors consist of TRISO particles, which are embedded in a graphite matrix stochastically. The reactor core is also stochastically filled with pebbles. These two stochastic geometries comprise the so-called double heterogeneous nature of this type of reactor. In this paper, a pebble bed reactor, the HTR-10, is used to demonstrate a treatment of this double heterogeneity using the MCNP5 Monte Carlo code and MATLAB programming. In this technique, TRISO particles are modeled in a pebble using the expanded FILL and LATTICE features of MCNP5. MATLAB is used to generate random numbers which represent the location of pebbles in the core. Centers of pebbles are generated stochastically and uniformly and then transferred into the MCNP5 input file as the centers of spherical surfaces. In this model, there is no approximation to the actual geometry. In other words, the double heterogeneous nature is preserved while truncating neither the pebbles in the core nor the particles in the pebble matrix. Finally, to validate the model, benchmark problems of IAEA are used. Very good agreement with experimental results is observed.

  8. Ceramic Methyltrioxorhenium

    CERN Document Server

    Herrmann, R; Eickerling, G; Helbig, C; Hauf, C; Miller, R; Mayr, F; Krug von Nidda, H A; Scheidt, E W; Scherer, W; Herrmann, Rudolf; Troester, Klaus; Eickerling, Georg; Helbig, Christian; Hauf, Christoph; Miller, Robert; Mayr, Franz; Nidda, Hans-Albrecht Krug von; Scheidt, Ernst-Wilhelm; Scherer, Wolfgang

    2006-01-01

    The metal oxide polymeric methyltrioxorhenium [(CH3)xReO3] is an unique epresentative of a layered inherent conducting organometallic polymer which adopts the structural motifs of classical perovskites in two dimensions (2D) in form of methyl-deficient, corner-sharing ReO5(CH3) octahedra. In order to improve the characteristics of polymeric methyltrioxorhenium with respect to its physical properties and potential usage as an inherentconducting polymer we tried to optimise the synthetic routes of polymeric modifications of 1 to obtain a sintered ceramic material, denoted ceramic MTO. Ceramic MTO formed in a solvent-free synthesis via auto-polymerisation and subsequent sintering processing displays clearly different mechanical and physical properties from polymeric MTO synthesised in aqueous solution. Ceramic MTO is shown to display activated Re-C and Re=O bonds relative to MTO. These electronic and structural characteristics of ceramic MTO are also reflected by a different chemical reactivity compared with its...

  9. 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

  10. Study of Li 2TiO 3 + 5 mol% TiO 2 lithium ceramics after long-term neutron irradiation

    Science.gov (United States)

    Chikhray, Y.; Shestakov, V.; Maksimkin, O.; Turubarova, L.; Osipov, I.; Kulsartov, T.; Kuykabayeba, A.; Tazhibayeva, I.; Kawamura, H.; Tsuchiya, K.

    2009-04-01

    Given work presents the results of complex material-science studies of 1 mm diameter ceramic pebbles manufactured of Li 2TiO 3 + 5 mol% TiO 2 ceramics before and after long-time neutron irradiation. Ceramic samples were placed in specially ampoules (six items) made of stainless steel Cr18Ni10Ti which were vacuumized and filled with helium. Irradiation of ampoules was carried out in the loop channel of WWRK reactor (Almaty, Kazakhstan) during 223 days at 6 MW power. After irradiation light-colored pebbles became grey-colored due to structure changes which generation of grey-colored inclusions (lithium oxide) with low density and microhardness. There is a radiation softening of lithium ceramic and that effect is higher for lower irradiation temperature 760 K than for 920 K. The value of maximum permissible load (pebble crash limit) at that is low and comprises ˜37.9 N. The content of residual tritium is higher for ceramic irradiated at 760 K (6.6 ± 0.6 × 10 11 Bq/kg) than for ceramic irradiated at 920 K (17 ± 3 × 10 10 Bq/kg). The size change indicates that pebble increase more after irradiation at 760 K than at 920 K where the bigger portion of tritium leaves the pebble. X-ray analysis shows radiation modification of Li 2TiO 3 + 5 mol% TiO 2 phase composition and generation of new phases: LiTi 2O 4, LiTiO 2 and Li 4Ti 5O 12.

  11. Study of Li{sub 2}TiO{sub 3} + 5 mol% TiO{sub 2} lithium ceramics after long-term neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Chikhray, Y. [Kazakh National University, Almaty (Kazakhstan)], E-mail: john@physics.kz; Shestakov, V. [Kazakh National University, Almaty (Kazakhstan); Maksimkin, O.; Turubarova, L.; Osipov, I. [Institute of Nuclear Physics, Almaty (Kazakhstan); Kulsartov, T.; Kuykabayeba, A.; Tazhibayeva, I. [National Nuclear Center, Kurchatov (Kazakhstan); Kawamura, H.; Tsuchiya, K. [JAEA, Oarai (Japan)

    2009-04-30

    Given work presents the results of complex material-science studies of 1 mm diameter ceramic pebbles manufactured of Li{sub 2}TiO{sub 3} + 5 mol% TiO{sub 2} ceramics before and after long-time neutron irradiation. Ceramic samples were placed in specially ampoules (six items) made of stainless steel Cr18Ni10Ti which were vacuumized and filled with helium. Irradiation of ampoules was carried out in the loop channel of WWRK reactor (Almaty, Kazakhstan) during 223 days at 6 MW power. After irradiation light-colored pebbles became grey-colored due to structure changes which generation of grey-colored inclusions (lithium oxide) with low density and microhardness. There is a radiation softening of lithium ceramic and that effect is higher for lower irradiation temperature 760 K than for 920 K. The value of maximum permissible load (pebble crash limit) at that is low and comprises {approx}37.9 N. The content of residual tritium is higher for ceramic irradiated at 760 K (6.6 {+-} 0.6 x 10{sup 11} Bq/kg) than for ceramic irradiated at 920 K (17 {+-} 3 x 10{sup 10} Bq/kg). The size change indicates that pebble increase more after irradiation at 760 K than at 920 K where the bigger portion of tritium leaves the pebble. X-ray analysis shows radiation modification of Li{sub 2}TiO{sub 3} + 5 mol% TiO{sub 2} phase composition and generation of new phases: LiTi{sub 2}O{sub 4}, LiTiO{sub 2} and Li{sub 4}Ti{sub 5}O{sub 12}.

  12. 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)

  13. Cold trapping of traces of tritiated water from the helium loops of a fusion breeder blanket

    International Nuclear Information System (INIS)

    The ITER Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) will comprise three helium loops designed for: tritium extraction from the breeder zone, heat removal, and purification of the coolant. The process step envisaged for tritium extraction as well as for coolant purification includes a cryogenic cold trap as main component for the removal of tritiated water vapour (mainly HTO, H2O). The concentrations of water in the gas streams are expected to be extremely small, i.e. of the order of 10 ppm by volume. In this paper, we describe first runs with a cold trap using helium as the carrier gas at flow rates of 0.1 and 1.0 m3/h. The range of water vapour concentration in the helium carrier gas was 0.5 to >200 ppmv. The experiments have demonstrated the ability of the cold trap to remove water vapour efficiently from the He stream down to concentrations of less than 0.02 ppmv when the inlet water concentration is in the range of 300-650 ppmv or higher

  14. Gas Reactor International Cooperative Program. Interim report. Safety and licensing evaluaion of German Pebble Bed Reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    The Pebble Bed Gas Cooled Reactor, as developed in the Federal Republic of Germany, was reviewed from a United States Safety and Licensing perspective. The primary concepts considered were the steam cycle electric generating pebble bed (HTR-K) and the process heat pebble bed (PNP), although generic consideration of the direct cycle gas turbine pebble bed (HHT) was included. The study examines potential U.S. licensing issues and offers some suggestions as to required development areas.

  15. 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...

  16. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Barbara H. Dolphin; James W. Sterbentz; Luka Snoj; Igor Lengar; Oliver Köberl

    2013-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.

  17. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Barbara H. Dolphin; James W. Sterbentz; Luka Snoj; Igor Lengar; Oliver Köberl

    2012-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.

  18. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    International Nuclear Information System (INIS)

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.

  19. Renewable lower reflector structure for a high temperature pebble bed reactor

    International Nuclear Information System (INIS)

    The description is given of a renewable lower reflector structure for a high temperature pebble bed reactor of the type comprising a cylindrical or prismatic graphite vessel wrapped in concrete and terminating at its lower end with a conical or pyramidal bottom fitted with a central aperture allowing the pebbles to be discharged by gravity. This structure includes a bed of several layers of protective graphite pebbles on the bottom and, fitted vertically so as to be removable along the centre line of the central aperture through the reflector and the concrete, a graphite block drilled in its centre to allow the discharge of the fuel pebbles and the protective pebbles. The graphite block rises above the level of the central aperture by an extent corresponding to the thickness of the bed when the reactor is working

  20. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  1. Challenges in Forming the Solar System's Giant Planet Cores via Pebble Accretion

    CERN Document Server

    Kretke, K A

    2014-01-01

    Though ~10 Earth mass rocky/icy cores are commonly held as a prerequisite for the formation of gas giants, theoretical models still struggle to explain how these embryos can form within the lifetimes of gaseous circumstellar disks. In recent years, aerodynamic-aided accretion of "pebbles," objects ranging from centimeters to meters in size, has been suggested as a potential solution to this long-standing problem. While pebble accretion has been demonstrated to be extremely effective in local simulations that look at the detailed behavior of these pebbles in the vicinity of a single planetary embryo, to date there have been no global simulations demonstrating the effectiveness of pebble accretion in a more complicated, multi-planet environment. Therefore, we have incorporated the aerodynamic-aided accretion physics into LIPAD, a Lagrangian code which can follow the collisional / accretional / dynamical evolution of a protoplanetary system, to investigate the how pebble accretion manifests itself in the larger ...

  2. Production of various sizes and some properties of beryllium pebbles by the rotating electrode method

    Energy Technology Data Exchange (ETDEWEB)

    Iwadachi, T.; Sakamoto, N.; Nishida, K. [NGK Insulators Ltd., Nagoya (Japan); Kawamura, H.

    1998-01-01

    The particle size distribution of beryllium pebbles produced by the rotating electrode method was investigated. Particle size depends on some physical properties and process parameters, which can practicaly be controlled by varying electrode angular velocities. The average particle sizes produced were expressed by the hyperbolic function with electrode angular velocity. Particles within the range of 0.3 and 2.0 mm in diameter are readily produced by the rotating electrode method while those of 0.2 mm in diameter are also fabricable. Sphericity and surface roughness were good in each size of pebble. Grain sizes of the pebbles are 17 {mu} m in 0.25 mm diameter pebbles and 260 {mu} m in 1.8 mm diameter pebbles. (author)

  3. DEM simulation of pebble flow in HTR-10 core by phenomenological method

    International Nuclear Information System (INIS)

    The 10 MW High Temperature Gas-cooled Reactor (HTR-10), developed at Tsinghua University, is an important test advanced reactor in the world. The pebble flow is of fundamental significance for the HTR-10. The discrete element method validated by experiments was used to study pebble flow in the HTR-10 core by the phenomenological method. A 1 : 1 scale computational model to the HTR-10 was utilized to simulate the motion of 27000 spheres, including the flows with different frictional coefficients and base angles. It is found that the pebble flow inside the HTR-10 core is uniform. The stagnant region does not exist. The larger the frictional coefficient or the base angle is, the more uniform the pebble flow is. When the frictional coefficient is 0.8, the HTR-10 maintains a normal discharge operation without stagnant pebbles. This work is important to further optimization of HTR design and development. (authors)

  4. 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.)

  5. Integrated design approach of the pebble bed modular using models

    International Nuclear Information System (INIS)

    The Pebble Bed Modular Reactor (PBMR) is the first pebble bed reactor that will be utilised in a high temperature direct Brayton cycle configuration. This implies that there are a number of unique features in the PBMR that extend from the German experience base. One of the challenges in the design of the PBMR is managing the integrated design process between the designers, the physicists and the analysts. This integrated design process is managed through model-based development work. Three-dimensional CAD models are constructed of the components and parts in the reactor. From the CAD models, CFD models, neutronic models, shielding models, FEM models and other thermodynamic models are derived. These models range from very simple models to extremely detailed and complex models. The models are used in legacy software as well as commercial off-the-shelf software. The different models are also used in code-to-code comparisons to verify the results. This paper will briefly discuss the different models and the interaction between the models, showing the iterative design process that is used in the development of the reactor at PBMR. (author)

  6. Dynamics of a small direct cycle pebble bed HTR

    International Nuclear Information System (INIS)

    The Dutch market for combined generation of heat and power identifies a unit size of 40 MW thermal for the conceptual design of a nuclear cogeneration plant. The ACACIA system provides 14 MW(e) electricity combined with 17 t/h of high temperature steam (220 deg. C, 10 bar) with a pebble bed high temperature reactor directly coupled with a helium compressor and a helium turbine. To come to quantitative statements about the ACACIA transient behaviour, a calculational coupling between the high temperature reactor core analysis code package Panthermix (Panther-Thermix/Direkt) and the thermal hydraulic code RELAP5 for the energy conversion system has been made. This paper will present the analysis of safety related transients. The usual incident scenarios Loss of Coolant Incident (LOCI) and Loss of Flow Incident (LOFI) have been analysed. Besides, also a search for the real maximum fuel temperature (inside a fuel pebble anywhere in the core) has been made. It appears that the maximum fuel temperatures are not reached during a LOFI or LOCI with a halted mass flow rate, but for situations with a small mass flow rate, 1-0.5%. As such, a LOFI or LOCI does not represent the worst-case scenario in terms of maximal fuel temperature. (author)

  7. Design windows for accelerator driven pebble-bed transmutators

    International Nuclear Information System (INIS)

    Nuclear waste transmutation can be achieved by different strategies. In this paper, the studies are focused in the 'Once Through' scenario, consisting in the nuclear waste transmutation until a maximum burnup (BU) is achieved. After transmutation, the fuel elements can be disposed in a Deep Storage Facility (DSF.) The main advantage of this strategy is that only one reprocess step is necessary. The drawback of this strategy consists mainly in the need of a fuel element design capable of withstanding very high burn-ups. It has been demonstrated that pebbles fuel elements in a pebble bed reactor design can withstand 700 MWd/Kg BU. This reactor presents the possibility of attainment different neutron spectrum with different fuel element designs, presents good safety characteristics, and the possibility of replacing the fuel elements easily inside the reactor (necessary for recycling strategies.) The transmutation process can be achieved in two steps. The first one, as a critical reactor, and the second one, as a subcritical assembly driven by an accelerator. In this paper, the optimum spectrum for the 'Once Through' strategy is presented, and some safety characteristics of the subcritical assembly are introduced. (authors)

  8. 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)

  9. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  10. 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

  11. Neutronics design for a fusion breeder

    International Nuclear Information System (INIS)

    As a fusion breeder, one of the most important figure is support ratio which reflects the economic and fuel production performance of the system to a great extent. In this paper, the support ratio is calculated by using one dimension transport program ANISN and optimized by adjusting 6Li enrichment and blanket arrangement. The radial distribution of producted U-233 is also taken into account. Measures are taken for better blanket design, and satisfactory results are obtained. Tritium breeding ratio T reaches 1.11 and the support ratio is enhanced from 11 to 14. The engineering, safety and environment performance are improved

  12. Coated fuel particle temperature analysis of the pebble-bed modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    In the 200 MWe pebble-bed modular high temperature gas-cooled reactor (HTR-PM), each sphere fuel element contains approximately 12,000 coated fuel particles scattered in the inner graphite matrix with a diameter of 50 mm to form the fuel zone, while the outer shell with a thickness of 5 mm is a fuel-free zone made up of the same graphite material. The coated fuel particle, with a diameter of less than 1 mm, consists of a UO2 kernel in 0.5 mm diameter and multiple coated ceramic layers. The HTR-PM has good inherent safety properties, one of which is exhibited like that, under some transient or accidental situations leading to an unexpected power increase, the reactor can shut down itself automatically or be brought down to a very low power level only by the negative temperature coefficient of reactivity due to the fuel temperature rise. During the calculation of the fuel element temperature with the pebble bed reactor analysis software THERMIX, which was originally developed by the German KFA-Juelich, a uniform power density in the fuel zone is assumed, without considering the temperature difference between the coated fuel particles and the surrounding graphite matrix. In this paper, the reactor temperature feedback characteristics and the nuclear power during a rapid reactivity introduction accident are analyzed in detail for two cases, i.e. taking into account the coated fuel particle temperature or not. The calculation results show that, the coated fuel particle temperature rises more quickly than the graphite matrix, and then the reactor power descends after a limited increase due to the higher negative temperature coefficient of reactivity of the fuel particle compared with that of the graphite moderator. Besides, the calculation conservation of the THERMIX code is revealed, and the safety properties of the HTR-PM are illustrated as well. (authors)

  13. 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.

  14. Ceramic glossary

    International Nuclear Information System (INIS)

    This book is a 2nd edition that contains new terms reflecting advances in high technology applications of ceramic materials. Definitions for terms which materials scientists, engineers, and technicians need to know are included

  15. The Shippingport Pressurized Water Reactor and Light Water Breeder Reactor

    International Nuclear Information System (INIS)

    This report discusses the Shippingport Atomic Power Station, located in Shippingport, Pennsylvania, which was the first large-scale nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. A program was started in 1953 at the Bettis Laboratory to confirm the practical application of nuclear power for large-scale electric power generation. It led to the development of zirconium alloy (Zircaloy) clad fuel element containing bulk actinide oxide ceramics (UO2, ThO2, ThO2 -- UO2, ZrO2 -- UO2) as nuclear reactor fuels. The program provided much of the technology being used for design and operation of the commercial, central-station nuclear power plants now in use. The Shippingport Pressurized Water Reactor (PWR) began initial power operation on December 18, 1957, and was a reliable electric power producer until February 1974. In 1965, subsequent to the successful operation of the Shippingport PWR (UO2, ZrO2 -- UO2 fuels), the Bettis Laboratory undertook a research and development program to design and build a Light Water Breeder Reactor (LWBR) core for operation in the Shippingport Station. Thorium was the fertile fuel in the LWBR core and was the base oxide for ThO2 and ThO2 -- UO2 fuel pellets. The LWBR core was installed in the pressure vessel of the original Shippingport PWR as its last core before decommissioning. The LWBR core started operation in the Shippingport Station in the autumn of 1977 and finished routine power operation on October 1, 1982. Successful LWBR power operation to over 160% of design lifetime demonstrated the performance capability of the core for both base-load and swing-load operation. Postirradiation examinations confirmed breeding and successful performance of the fuel system

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

    International Nuclear Information System (INIS)

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

  17. 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

  18. 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

  19. Fast breeder reactor fuel reprocessing in France

    International Nuclear Information System (INIS)

    Simultaneous with the effort on fast breeder reactors launched several years ago in France, equivalent investigations have been conducted on the fuel cycle, and in particular on reprocessing, which is an indispensable operation for this reactor. The Rapsodie experimental reactor was associated with the La Hague reprocessing plant AT1 (1 kg/day), which has reprocessed about one ton of fuel. The fuel from the Phenix demonstration reactor is reprocessed partly at the La Hague UP2 plant and partly at the Marcoule pilot facility, undergoing transformation to reprocess all the fuel (TOR project, 5 t/y). The fuel from the Creys Malville prototype power plant will be reprocessed in a specific plant, which is in the design stage. The preliminary project, named MAR 600 (50 t/y), will mobilize a growing share of the CEA's R and D resources, as the engineering needs of the UP3 ''light water'' plant begins to decline. Nearly 20 tonnes of heavy metals irradiated in fast breeder reactors have been processed in France, 17 of which came from Phenix. The plutonium recovered during this reprocessing allowed the power plant cycle to be closed. This power plant now contains approximately 140 fuel asemblies made up with recycled plutonium, that is, more than 75% of the fuel assemblies in the Phenix core

  20. 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

  1. Industrial solar breeder project using concentrator photovoltaics

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, R; Wohlgemuth, J; Burkholder, J; Levine, A; Storti, G; Wrigley, C; McKegg, A

    1979-08-01

    The purpose of this program is to demonstrate the use of a concentrating photovoltaic system to provide the energy for operating a silicon solar cell production facility, i.e., to demonstrate a solar breeder. Solarex has proposed to conduct the first real test of the solar breeder concept by building and operating a 200 kW(e) (peak) concentrating photovoltaic system based on the prototype and system design developed during Phase I. This system will provide all of the electrical and thermal energy required to operate a solar cell production line. This demonstration would be conducted at the Solarex Rockville facility, with the photovoltaic array located over the company parking lot and on an otherwise unusable flood plain. Phase I of this program included a comprehensive analysis of the application, prototype fabrication and evaluation, system design and specification, and a detailed plan for Phases II and III. A number of prototype tracking concentrator solar collectors were constructed and operated. Extensive system analysis was performed to design the Phase II system as a stand-alone power supply for a solar cell production line. Finally, a detailed system fabrication proposal for Phase II and an operation and evaluation plan for Phase III were completed. These proposals included technical, management, and cost plans for the fabrication and exercise of the proposed system.

  2. 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

  3. Improved structural materials for fast breeder reactors

    International Nuclear Information System (INIS)

    Electricity plays a crucial role in the economic development of our country. Coal is the primary fuel for generation of electricity in India as in many other countries. In India, generation of power by nuclear reactors is very important because of (i) availability of large thorium resource, (ii) constraints on setting up of fossil fuel based power plants and (iii) the negligibly small green house gas emissions by nuclear energy. The nuclear programme of the country is being implemented in three stages: (i) pressurized heavy water reactors of the CANDU type, (ii) sodium-cooled fast reactors and (iii) thorium-based reactors. Sodium-cooled fast reactor (SFR) technology is envisioned to make use of the large thorium reserves available. India has undertaken and made rapid strides in developing SFR technology and building of fast reactors for energy generation. A Fast Breeder Test Reactor (FBTR) of 40 MWt is operating successfully for over 25 years at Indira Gandhi Centre for Atomic Research. Based on the design, construction and operational experience, a 500 MWe Prototype Fast Breeder Reactor (PFBR) has been designed indigenously and is in an advanced stage of construction. Its design is being further optimised for enhanced economy with respect to cost of electricity production, for use in commercial reactors. Currently, several R and D programmes are under implementation for the development of new materials required for improved economy of commercial fast reactors

  4. The fast breeder reactor. v. 1

    International Nuclear Information System (INIS)

    The Energy Committee's report was prepared after hearing evidence (the minutes of which are published in Volume II) from the Central Electricity Generating Board, the United Kingdom Atomic Energy Authority and the Department of Energy. Memoranda received from other interested bodies or individuals were also considered and members of the Committee visited fast breeder projects in France, West Germany and Japan. As well as the development of the fast reactors, the economics and timescale were reviewed. The particular case of the fast breeder reactor and proposed fuel reprocessing plant at Dounreay was considered. The main conclusion is that major expenditure on fast reactor programmes can only be justified if there is a potential economic case, i.e. if the fuel cycle costs are lower than for PWRs. This would only be the case if uranium costs increased greatly. It is not considered worthwhile to participate in the European Fast Reactor although this should be reviewed in 1993 and 1997. The Committee agree with the Government's decision to cease funding the PFR in 1994 and endorses the need to regenerate the local economy which will be affected by this decision. (UK)

  5. Solid breeder blanket design and tritium breeding

    International Nuclear Information System (INIS)

    Thermonuclear D-T power plants will have to be tritium self-sufficient. In addition to recovering the energy carried by the fusion neutrons (about 80% of the fusion energy), the blanket of the reactor will thus have to breed tritium to replace that burnt in the fusion process. This paper is an attempt to cover in a concise way the questions of tritium breeding, and the influence of this issue on the design of, and the material selection for, power reactor blanket relying on the use of solid breeder materials. Tritium breeding requirements - to breed one tritium per fusion neutron - are shown to be quite demanding. To meet them, the blanket must incorporate, in addition to a tritium breeding lithium compound, a neutron multiplier so as to compensate for neutron losses. Presently prefered lithium compounds are Li2O, LiAlO2, Li2ZrO3, Li4SiO4. The neutron multiplier considered in most design concepts is beryllium. Furthermore, the blanket must be designed with a view to minimizing these neutron losses (search for compactness and high coverage ratio of the plasma while minimizing the amount of structures and coolant). The design guidelines are justified and the technological problems which limit their implementation are discussed and illustrated with typical designs of solid breeder blanket. (orig.)

  6. Fabrication of Li{sub 2}TiO{sub 3} pebbles by a freeze drying process

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang-Jin, E-mail: lee@mokpo.ac.kr [Department of Advanced Materials Science and Engineering, Mokpo National University, Muan 534-729 (Korea, Republic of); Park, Yi-Hyun [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Yu, Min-Woo [Department of Advanced Materials Science and Engineering, Mokpo National University, Muan 534-729 (Korea, Republic of)

    2013-11-15

    Li{sub 2}TiO{sub 3} pebbles were successfully fabricated by using a freeze drying process. The Li{sub 2}TiO{sub 3} slurry was prepared using a commercial powder of particle size 0.5–1.5 μm and the pebble pre-form was prepared by dropping the slurry into liquid nitrogen through a syringe needle. The droplets were rapidly frozen, changing their morphology to spherical pebbles. The frozen pebbles were dried at −10 °C in vacuum. To make crack-free pebbles, some glycerin was employed in the slurry, and long drying time and a low vacuum condition were applied in the freeze drying process. In the process, the solid content in the slurry influenced the spheroidicity of the pebble green body. The dried pebbles were sintered at 1200 °C in an air atmosphere. The sintered pebbles showed almost 40% shrinkage. The sintered pebbles revealed a porous microstructure with a uniform pore distribution and the sintered pebbles were crushed under an average load of 50 N in a compressive strength test. In the present study, a freeze drying process for fabrication of spherical Li{sub 2}TiO{sub 3} pebbles is introduced. The processing parameters, such as solid content in the slurry and the conditions of freeze drying and sintering, are also examined.

  7. Optimization of OTTO Fuel Management in Pebble-Bed Reactors Using Particle Swarm Algorithm

    International Nuclear Information System (INIS)

    Pebble-Bed nuclear reactors feature highly flexible in-core fuel management capabilities due to on-line fueling and thermo-mechanical robust fuel design. Fuel pebbles with various fissile and fertile materials can be loaded into the reactor core at different rates. The fuel pebbles may be recirculated in the core several times until reaching their target burnup, or reach their target burnup in single pass through the core (OTTO- Once-Through-Then-Out fueling Scheme). Pebble-bed reactors have relatively efficient neutron economy since they operate with low excess reactivity and hence minimize the use of neutron poisons and control rods. Moreover, the fuel pebble robust design permits high burnup levels (up to 140000 MWD/THM). The flexibility of the fuel management operations allows enhancing fuel utilization. Traditionally fuel cycle design decisions were made using expert opinions and parametric studies. In this work, we have used the Particle Swarm Optimization (PSO) algorithm to optimize fuel utilization of pebble-bed reactors running OTTO fuel management. Optimization was carried out also for cores with Th232 as fertile material. Preliminary calculations were performed for a large core with 2 radial fuel loading zones. Results of the optimal fuel utilization performed for cores with UO2 fuel and cores with (Th- U)O2. Future work will include optimization of cores fuelled with separate seed (U) and blanket (Th) fuel pebbles and with advanced modular core configuration, like the PBMR400

  8. Mechanical compression tests of beryllium pebbles after neutron irradiation up to 3000 appm helium production

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V., E-mail: vladimir.chakin@kit.edu [Karlsruhe Institute of Technology, Institite for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R.; Moeslang, A. [Karlsruhe Institute of Technology, Institite for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2015-04-15

    Highlights: • Compression tests of highly neutron irradiated beryllium pebbles have been performed. • Irradiation hardening of beryllium pebbles decreases the steady-state strain-rates. • The steady-state strain-rates of irradiated beryllium pebbles exceed their swelling rates. - Abstract: Results: of mechanical compression tests of irradiated and non-irradiated beryllium pebbles with diameters of 1 and 2 mm are presented. The neutron irradiation was performed in the HFR in Petten, The Netherlands at 686–968 K up to 1890–2950 appm helium production. The irradiation at 686 and 753 K cause irradiation hardening due to the gas bubble formation in beryllium. The irradiation-induced hardening leads to decrease of steady-state strain-rates of irradiated beryllium pebbles compared to non-irradiated ones. In contrary, after irradiation at higher temperatures of 861 and 968 K, the steady-state strain-rates of the pebbles increase because annealing of irradiation defects and softening of the material take place. It was shown that the steady-state strain-rates of irradiated beryllium pebbles always exceed their swelling rates.

  9. METHODS FOR MODELING THE PACKING OF FUEL ELEMENTS IN PEBBLE BED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Abderrafi M. Ougouag; Joshua J. Cogliati; Jan-Leen Kloosterman

    2005-09-01

    Two methods for the modeling of the packing of pebbles in the pebble bed reactors are presented and compared. The first method is based on random generation of potential centers for the pebbles, followed by rejection of points that are not compatible with the geometric constraint of no (or limited) pebbles overlap. The second method models the actual physical packing process, accounting for the dynamic of pebbles as they are dropped onto the pebble bed and as they settle therein. A simplification in the latter model is the assumption of a starting point with very dilute packing followed by settling. The results from the two models are compared and the properties of the second model and the dependence of its results on many of the modeling parameters are presented. The first model (with no overlap allowed) has been implemented into a code to compute Dancoff factors. The second model will soon be implemented into that same code and will also be used to model flow of pebbles in a reactor and core densification in the simulation of earthquakes. Both methods reproduce experimental values well, with the latter displaying a high level of fidelity.

  10. 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.

  11. Mass loss of Li{sub 2}TiO{sub 3} pebbles and Li{sub 4}SiO{sub 4} pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Kashimura, Hideaki, E-mail: kadzu@nucl.kyushu-u.ac.jp [Interdisciplinary Graduate School of Engineering Science, Kyushu University, 6-10-1, Hakozaki, Higashi-ku, Fukuoka 812-8581 (Japan); Nishikawa, Masabumi; Katayama, Kazunari; Matsuda, Shohei; Shimozori, Motoki; Fukada, Satoshi [Interdisciplinary Graduate School of Engineering Science, Kyushu University, 6-10-1, Hakozaki, Higashi-ku, Fukuoka 812-8581 (Japan); Hoshino, Tsuyoshi [Blanket Irradiation and Analysis Group, Fusion Research and Development Directorate, Japan Atomic Energy Agency, 2-166, Oaza-Obuchi-Aza-Omotedate, Rokkasho-mura, Kamikita-gun, Aomori 039-3212 (Japan)

    2013-10-15

    It has been known that water vapor is released from ceramic breeder materials into the purge gas due to desorption of adsorbed water under dry atmosphere and due to the water formation reaction under hydrogen atmosphere. However, an effect of water vapor in the purge gas to Li mass loss has not been understood. In this study, mass loss of Li{sub 2}TiO{sub 3} (NFI) and Li{sub 4}SiO{sub 4} (FzK) under hydrogen atmosphere (1000 Pa H{sub 2}/Ar), and mass loss of Li{sub 2}TiO{sub 3} (NFI) and Li{sub 2}TiO{sub 3} with additional Li which is in a developmental stage (JAEA) under water vapor atmosphere (50 Pa H{sub 2}O/Ar) were compared, respectively. It was found that under hydrogen atmosphere Li mass loss of Li{sub 4}SiO{sub 4} and Li{sub 2}TiO{sub 3} is same degree although the amount of water vapor released from Li{sub 4}SiO{sub 4} is larger than that from Li{sub 2}TiO{sub 3}. It was clarified with regard to Li{sub 2}TiO{sub 3} that Li mass loss in water vapor atmosphere is larger than that in hydrogen atmosphere. Mass loss of Li{sub 2}TiO{sub 3} with additional Li (JAEA) was larger than that of Li{sub 2}TiO{sub 3} (NFI). It was observed by X-ray analysis that Li deposits formed on the inner wall of the quartz tube contain Li{sub 2}SiO{sub 3}.

  12. Monte Carlo criticality calculation for Pebble-type HTR-PROTEUS core

    International Nuclear Information System (INIS)

    These days, pebble-bed and other High-Temperature Gas-cooled Reactor (HTGR) designs are once again in vogue in connection with hydrogen production. In this study, as a part of establishing Monte Carlo computation system for HTGR core analysis, some criticality calculations for pebble-type HTGR were carried out using MCNP code. Firstly, the pebble-bed cores of HTR-PROTEUS critical facility in Swiss were selected for the benchmark model, and, after the detailed MCNP modeling of the whole facility, criticality calculations were performed. It was also investigated the homogenization effect of TRISO fuel on criticality

  13. Effect of non-uniform porosity distribution on thermalhydraulics in a pebble bed reactor

    International Nuclear Information System (INIS)

    In pebble bed reactors, the porosity profile shows strong fluctuations near the wall. These changes in fuel density affect local power density, coolant velocity, and temperature distribution. This paper describes the pebFoam code, capable of calculating pebble bed thermohydraulics including non-uniform porosity distributions for arbitrary geometries, and investigates the changes in velocity, pressure drop, and helium and pebble temperatures when using a nonuniform porosity distribution instead of a uniform distribution. Results show only minor changes in temperature profiles and pressure drop for full power steady state calculations, though the velocity profile shows a clear increase in velocity near the wall. (author)

  14. Uranium deposits in Proterozoic quartz-pebble conglomerates

    International Nuclear Information System (INIS)

    This report is the result of an effort to gather together the most important information on uranium deposits in Proterozoic quartz-pebble conglomerates in the United States of America, Canada, Finland, Ghana, South Africa and Australia. The paper discusses the uranium potential (and in some cases also the gold potential in South Africa, Western Australia and Ghana) in terms of ores, sedimentation, mineralization, metamorphism, placers, geologic formations, stratigraphy, petrology, exploration, tectonics and distribution. Geologic history and application of geologic models are also discussed. Glacial outwash and water influx is also mentioned. The uranium deposits in a number of States in the USA are covered. The Witwatersrand placers are discussed in several papers. Refs, figs, tabs

  15. Pebble bed reactor with one-zone core

    International Nuclear Information System (INIS)

    The claim deals with measures to differentiate the flow rate and to remove spherical fuel elements in the core of a pebble bed reactor. Hence the vertical rate of the fuel elements in the border region is for example twice as much as in the centre. A central funnel-shaped outlet on the floor of the core container over which a conical body is placed with its peak pointing upwards, or also the forming of several outlets can be used to adjust to a certain exit rate for the fuel elements. The main target of the invention is a radially extensively constant coolant outlet temperature at the outlet of the core which determines the effectiveness of the connected heat exchanger and thus contributes to economy. (orig./PW)

  16. Neutron wave experiment in a graphite pebble-bed system

    International Nuclear Information System (INIS)

    The propagation of neutron waves through a Type-AVR graphite pebble-bed is studied. Use of a sinusoidally modulated source of neutrons is equivalent to 'poisoning' a moderator with a 1/v poison. The inverse relaxation length of the neutron wave amplitude and the variation of the phase angle as function of position are dependent upon the frequency of modulation and the neutron diffusion and thermalization parameters of the media in which the waves are being propagated. The diffusion coefficient D0 of a system of graphite spheres is determined to a high accuracy. In the termal energy range a streaming correction of 14,8% is necessary if for calculation the graphite of the spheres is homogenized. (orig.)

  17. 'Once through' cycles in the pebble bed HTR

    International Nuclear Information System (INIS)

    In the pebble bed HTR the 'Once Through' cycles achieve a favorable conservation of uranium resources due to their high burnup and due to the relatively low fissile inventory. A detailed study is given for cycles with highly enriched uranium and thorium, 20% enriched uranium and thorium, and for the low (approximately 8%) enriched cycle. The recommended cycle is based on the known THTR fuel element in the Th/U (93%) cycle. The variant with separate Seed elements and Breed elements presents the best pioneer in view of later recycling and thermal breeding. The minimum proliferation risk is achieved in the Th/U (20%) cycle basing on the fuel element type of the AVR, due to the low amount and high denaturization of the disloaded plutonium. (orig.)

  18. Modeling of tritium transport in a pin-type solid breeder blanket

    International Nuclear Information System (INIS)

    This section of the pin-type solid breeder blanket study is the first detailed attempt at modeling tritium inventory and release within a fusion reactor blanket based on actual tritium generation and thermal hydraulic profiles, rather, than a simple average unit cell extrapolation or some assumed exponential profiles. The DIFFUSE 83 code was found to give inventory results consistent with previous modeling efforts and with the general spherical grain model. The inventories for this blanket design calculated using DIFFUSE were found to be very satisfactory, less than 14 g at steady-state for a 129,000 kg LiAlO2 blanket. This result is reasonable compared with a BCSS LiALO2 blanket inventory calculated by GA Technologies. DIFFUSE was found to be very useful in approximating tritium inventories during transient startup/shutdown modes. The evaluations of transient inventories in this study appear to be the most detailed to date. The results suggest the need for controlling coolant flow during start-up to maintain high breeder temperatures and low tritium inventory, and the use of pre-heated coolant to bake out tritium inventory after shut-down. DIFFUSE modeling of breeder pins of 100% theoretical density indicates very limited tritium release from the LiAlO2 ceramic, suggesting batch processing of the pins for tritium extraction at the end of blanket lifetime. Preliminary analysis of other surface and radiative trapping effects shows DIFFUSE to be potentially a very useful tool in approximating and evaluating experimental results. Additional DIFFUSE analysis of these effects given the available experimental data is warranted

  19. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  20. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    International Nuclear Information System (INIS)

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel digital

  1. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Leland M. Montierth

    2013-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

  2. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Leland M. Montierth

    2014-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

  3. 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

  4. 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

  5. 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)

  6. 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

  7. Special topics reports for the reference tandem mirror fusion breeder: liquid metal MHD pressure drop effects in the packed bed blanket. Vol. 1

    International Nuclear Information System (INIS)

    Magnetohydrodynamic (MHD) effects which result from the use of liquid metal coolants in magnetic fusion reactors include the modification of flow profiles (including the suppression of turbulence) and increases in the primary loop pressure drop and the hydrostatic pressure at the first wall of the blanket. In the reference fission-suppressed tandem mirror fusion breeder design concept, flow profile modification is a relatively minor concern, but the MHD pressure drop in flowing the liquid lithium coolant through an annular packed bed of beryllium/thorium pebbles is directly related to the required first wall structure thickness. As such, it is a major concern which directly impacts fissile breeding efficiency. Consequently, an improved model for the packed bed pressure drop has been developed. By considering spacial averages of electric fields, currents, and fluid flow velocities the general equations have been reduced to simple expressions for the pressure drop. The averaging approach results in expressions for the pressure drop involving a constant which reflects details of the flow around the pebbles. Such details are difficult to assess analytically, and the constant may eventually have to be evaluated by experiment. However, an energy approach has been used in this study to bound the possible values of the constant, and thus the pressure drop. In anticipation that an experimental facility might be established to evaluate the undetermined constant as well as to address other uncertainties, a survey of existing facilities is presented

  8. Flow distribution of pebble bed high temperature gas cooled reactors using Large Eddy Simulation

    International Nuclear Information System (INIS)

    The simulation of complex three-dimensional gas flow through the gaps of the spherical fuel elements (fuel pebbles) of Pebble Bed Modulator Reactor is performed. This will help in understanding the highly three-dimensional, complex flow phenomena in pebble bed caused by flow curvature. The flow of this type has distinctive features, which strongly affect the boundary layer behavior. The transition from a laminar to turbulent flow around this curved flow occurs at different Reynolds (Re) numbers. Noncircular curved flows as in the pebble-bed situation need to be investigated. In this study, Large Eddy Simulation (LES) is used in modeling the turbulence to overcome the shortcoming of the Reynolds Average Navier-Stokes approach. (author)

  9. Computational and experimental prediction of dust production in pebble bed reactors, Part II

    International Nuclear Information System (INIS)

    Highlights: • Custom-built high temperature, high pressure tribometer is designed. • Two different wear phenomena at high temperatures are observed. • Experimental wear results for graphite are presented. • The graphite wear dust production in a typical Pebble Bed Reactor is predicted. -- Abstract: This paper is the continuation of Part I, which describes the high temperature and high pressure helium environment wear tests of graphite–graphite in frictional contact. In the present work, it has been attempted to simulate a Pebble Bed Reactor core environment as compared to Part I. The experimental apparatus, which is a custom-designed tribometer, is capable of performing wear tests at PBR relevant higher temperatures and pressures under a helium environment. This environment facilitates prediction of wear mass loss of graphite as dust particulates from the pebble bed. The experimental results of high temperature helium environment are used to anticipate the amount of wear mass produced in a pebble bed nuclear reactor

  10. Who cracked the pebbles in the gravel pit - lithostatic pressure or a bunch of faults?

    Science.gov (United States)

    Tuitz, Christoph; Exner, Ulrike; Grasemann, Bernhard; Preh, Alexander

    2010-05-01

    The occurrence of radially, brittle fractured pebbles from unconsolidated sediments were investigated in a gravel pit south of St. Margarethen (Burgenland, Austria). The outcrop is located in the Neogene Eisenstadt-Sopron Basin, which is a sub-basin on the SE border of the Vienna Basin. The sediments, which were deposited during the Sarmatian and Pannonian (12.7-7.2 Ma), represent a succession of deltaic gravels with intercalations of shallow-marine calcareous sands. Extensional tectonics in these sediments resulted in the generation of conjugate sets of predominately WNW- and subordinate ESE-dipping normal faults (shear deformation bands). These faults were primarily localized in meter-thick gravel layers and, with increasing displacement, eventually cross-cut other lithologies. The gravel layers contain a significant number of cracked pebbles. Detailed structural mapping of the distribution of cracked pebbles revealed their preferential occurrence in the vicinity of the normal faults and, in these, within zones of roughly uniform-sized pebbles. The findings indicated a strong relation to the mechanics of faulting within the sediment. To find the controlling factors for the localization of pebble fracturing, the grain-size distribution and shape and the number of point contacts of the pebbles were statistically measured. Furthermore, on the basis of point load tests, a breakage criterion was statistically defined which characterizes the breakage behaviour of the pebbles. The results were then used as input parameters for numerical modelling. The Discrete Element Method was applied to simulate the effect of overburden on a certain volume of particles (i.e. the pebbles). In numerical uniaxial compression simulations, the magnitude and the distribution of contact forces between the particles were monitored during compressive loading and repetitively compared with the breakage criterion. If a particle in the simulation reached the criterion, it was automatically

  11. Modularity of the MIT Pebble Bed Reactor for use by the commercial power industry

    OpenAIRE

    Hanlon-Hyssong, Jaime E.

    2008-01-01

    CIVINS The Modular Pebble Bed Reactor is a small high temperature helium cooled reactor that is being considered for both electric power and hydrogen production. Pebble bed reactors are being developed in South Africa, China and the US. To make smaller 120 Mwe reactors economically competitive with larger 1500 Mwe traditional light water reactors changes in the way these plants are built are needed. Economies of production need to be sufficiently large to compete with economies of sca...

  12. STUDI PEMODELAN DAN PERHITUNGAN TRANSPORT MONTE CARLO DALAM TERAS HTR PEBBLE BED

    Directory of Open Access Journals (Sweden)

    Zuhair .

    2013-01-01

    Full Text Available Konsep sistem energi VHTR baik yang berbahan bakar pebble (VHTR pebble bed maupun blok prismatik (VHTR prismatik menarik perhatian fisikawan reaktor nuklir. Salah satu kelebihan teknologi bahan bakar bola adalah menawarkan terobosan teknologi pengisian bahan bakar tanpa harus menghentikan produksi listrik. Selain itu, partikel bahan bakar pebble dengan kernel uranium oksida (UO2 atau uranium oksikarbida (UCO yang dibalut TRISO dan pelapisan silikon karbida (SiC dianggap sebagai opsi utama dengan pertimbangan performa tinggi pada burn-up bahan bakar dan temperatur tinggi. Makalah ini mendiskusikan pemodelan dan perhitungan transport Monte Carlo dalam teras HTR pebble bed. HTR pebble bed adalah reaktor berpendingin gas temperatur tinggi dan bermoderator grafit dengan kemampuan kogenerasi. Perhitungan dikerjakan dengan program MCNP5 pada temperatur 1200 K. Pustaka data nuklir energi kontinu ENDF/B-V dan ENDF/B-VI dimanfaatkan untuk melengkapi analisis. Hasil perhitungan secara keseluruhan menunjukkan konsistensi dengan nilai keff yang hampir sama untuk pustaka data nuklir yang digunakan. Pustaka ENDF/B-VI (66c selalu memproduksi keff lebih besar dibandingkan ENDF/B-V (50c maupun ENDF/B-VI (60c dengan bias kurang dari 0,25%. Kisi BCC memprediksi keff hampir selalu lebih kecil daripada kisi lainnya, khususnya FCC. Nilai keff kisi BCC lebih dekat dengan kisi FCC dengan bias kurang dari 0,19% sedangkan dengan kisi SH bias perhitungannya kurang dari 0,22%. Fraksi packing yang sedikit berbeda (BCC= 61%, SH= 60,459% tidak membuat bias perhitungan menjadi berbeda jauh. Estimasi keff ketiga model kisi menyimpulkan bahwa model BCC lebih bisa diadopsi dalam perhitungan HTR pebble bed dibandingkan model FCC dan SH. Verifikasi hasil estimasi ini perlu dilakukan dengan simulasi Monte Carlo atau bahkan program deterministik lainnya guna optimisasi perhitungan teras reaktor temperatur tinggi.   Kata-kunci: kernel, TRISO, bahan bakar pebble, HTR pebble bed

  13. Prototype studies on the nondestructive online burnup determination for the modular pebble bed reactors

    International Nuclear Information System (INIS)

    Highlights: • Prototype study of online burnup measurement for HTR proves its feasibility. • Calibration and its correction of burnup assay device is discussed and verified. • Analysis of simulated gamma spectra shows good performance of spectra-unfolding method. - Abstract: The online fuel pebble burnup determination in future modular pebble bed reactor is implemented by measuring nondestructively the activity of the monitoring nuclide Cs-137 with HPGe detector on a pebble-by-pebble basis. Based on a full size prototype the feasibility is investigated. The prototype was first tested by using double sources to show that a precision of 2.8% (1σ) can be achieved in the determination of the Cs-137 net counting rate. Then, the relationship between the Cs-137 activity and the net counting rate recorded in the HPGe detector is calibrated with a standard Cs-137 source contained in the center of a graphite sphere with the same dimension as a real fuel pebble. Because the self attenuation of the calibration source differs with a fuel pebble, a correction factor of 1.07 ± 0.02 (p = 0.95) to the calibration is derived by using the efficiency transfer method. Last, by analyzing the spectra generated with KORIGEN software followed by Monte Carlo simulation, it is predicted that the relative standard deviation of the Cs-137 net counting rate can be still controlled below 3.5% despite of the presence of all the interfering peaks. The results demonstrate the feasibility of utilizing HPGe gamma spectrometry in the online determination of the pebble burnup in future modular pebble bed reactors

  14. Experimental studies on heat transfer and pressure drop in pebble bed test facility

    International Nuclear Information System (INIS)

    Indian program for development of high temperature reactor and its utilization to supply process heat aimed to develop alternate fuel carrier to substitute petroleum based transport fuel, which has very small reserves in India and results in large import bills. Hydrogen is an attractive energy carrier for transport applications. It can be produced by splitting water which requires either electricity or process heat at high temperatures or both depending upon the process selected. BARC is carrying out design of a 600 MWth reactor capable of supplying process heat at around 1000 °C as required for hydrogen production. For this reactor various design options with respect to fuel configurations, such as prismatic bed and pebble bed were considered for thermal hydraulics analysis. Coolant options such as molten lead and molten salt were analyzed. Studies carried out indicate selection of pebble bed reactor core with molten salt as coolant. Thermal-hydraulic studies are required for pebble bed reactor. With this in view, a pebble bed test facility has been setup to study the heat transfer and pressure drop in pebble bed. Water is used as a working medium for the facility. The paper deals with the description of the pebble bed test facility and the experimental results of heat transfer and pressure drop. It also deals with the assessment of correlations for heat transfer and pressure drop in pebble bed geometry. Pressure drop experiments in the pebble bed test facility have been performed for Raynolds number ranges from 3000-12000. Various pressure drop correlations have been compared with the experimental data. It has been found that that the correlation given by Leva et. al. matches well with the experimental data. Various heat transfer correlations have also been compared. Heat transfer experiments are nearing completion

  15. Steady-state thermal-hydraulic of pebble bed blanket on hybrid reactor

    International Nuclear Information System (INIS)

    This paper gives thermal-hydraulic studies of pebble bed blanket on Hybrid Reactor. The concept of whole pebble bed blanket and the cooling methods are presented. The thermal-hydraulic characteristics of pebble bed are summarized. The theoretical model and code for solving heat transfer and flowing are presented. By using this code the calculation and analysis of thermal hydraulic of pebble bed Blanket of Hybrid Reactor are also given. In order to improve the flexibility, safety and economy, the authors select pebble beds not only to breed Tritium, but also to breed fission material and to multiply neutron. 5 MPa Helium is used as coolant and 0.05 MPa-0.1 MPa Helium is used as Purge gas. The heat transfer mechanisms of pebble bed are very complicated which include conduction, convection and radiation. In order to study the thermal-hydraulic of the bed, the authors just simply consider it as homogeneous and continuous binary phase medium as that used in the porous medium at the condition that the size of the bed is much greater than that of the balls. The coolant or the purge gas flowing through the bed is just considered existing a cooling source in the bed. It also significantly influences the effective conductivity's of the bed. Porous fraction, the main factor of the bed depends on the geometry position and parameters. From this model, one can obtain the thermal-hydraulic governing equations of the bed

  16. TEM study of beryllium pebbles after neutron irradiation up to 3000 appm helium production

    Energy Technology Data Exchange (ETDEWEB)

    Klimenkov, M., E-mail: michael.klimenkov@kit.edu [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Chakin, V.; Moeslang, A. [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R. [Institute for Applied Materials – Materials and Biomechanics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2013-11-15

    Beryllium is planned to be used as a neutron multiplier in the Helium Cooled Pebble Bed (HCPB) European concept of a breeding blanket of DEMO. In order to evaluate the irradiation performance, individual pebbles and constrained pebble beds were neutron irradiated at temperatures typical for fusion blanket. Beryllium pebbles with a diameter of 1 mm produced by the Rotating Electrode Method were subjected to a TEM study after irradiation at the HFR, Petten, at temperatures of 686, 753, 861, and 968 K. The helium production in the pebbles was calculated in the range from 2090 to 3090 appm. Gas bubbles as disks of hexagonal shape were observed for all four irradiation temperatures. The disks were oriented in the (0 0 0 1) basal plane with a height directed along the [0 0 0 1] “c” axis. The average diameters of the bubbles increase from 7.5 to 80 nm with increasing irradiation temperature, the bulk densities accordingly decrease from 4.4 × 10{sup 22} to 3.8 × 10{sup 20} m{sup −3}. With increasing irradiation temperature, the swelling of the pebbles increases from 0.6% at 686 K up to 6.5% at 968 K.

  17. Thermal-hydraulics numerical analyses of Pebble Bed Advanced High Temperature Reactor hot channel

    International Nuclear Information System (INIS)

    Background: The thermal hydraulics behavior of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) hot channel was studied. Purpose: We aim to analyze the thermal-hydraulics behavior of the PB-AHTR, such as pressure drop, temperature distribution of coolant and pebble bed as well as thermal removal capacity in the condition of loss of partial coolant. Methods: We used a modified FLUENT code which was coupled with a local non-equilibrium porous media model by introducing a User Defined Scalar (UDS) in the calculation domain of the reactor core and subjoining different resistance terms (Ergun and KTA) to calculate the temperature of coolant, solid phase of pebble bed and pebble center in the core. Results: Computational results showed that the resistance factor has great influence on pressure drop and velocity distribution, but less impact on the temperature of coolant, solid phase of pebble bed and pebble center. We also confirmed the heat removal capacity of the PB-AHTR in the condition of nominal and loss of partial coolant conditions. Conclusion: The numerical analyses results can provide a useful proposal to optimize the design of PB-AHTR. (authors)

  18. A comparison of fusion breeder/fission client and fission breeder/fission client systems for electrical energy production

    International Nuclear Information System (INIS)

    A parametric study that evaluated the economic performance of breeder/client systems is described. The linkage of the breeders to the clients was modelled using the stockpile approach to determine the system doubling time. Since the actual capital costs of the breeders are uncertain, a precise prediction of the cost of a breeder was not attempted. Instead, the breakeven capital cost of a breeder relative to the capital cost of a client reactor was established by equating the cost of electricity from the breeder/client system to the cost of a system consisting of clients alone. Specific results are presented for two breeder/client systems. The first consisted of an LMFBR with LWR clients. The second consisted of a DT fusion reactor (with a 238U fission suppressed blanket) with LWR clients. The economics of each system was studied as a function of the cost of fissile fuel from a conventional source. Generally, the LMFBR/LWR system achieved relatively small breakeven capital cost ratios; the maximum ratio computed was 2.2 (achieved at approximately triple current conventional fissile material cost). The DTFR/LWR system attained a maximum breakeven capital cost ratio of 4.5 (achieved at the highest plasma quality (ignited device) and triple conventional fissile cost)

  19. Light-water breeder reactors: preliminary safety and environmental information document. Volume III

    International Nuclear Information System (INIS)

    Information is presented concerning prebreeder and breeder reactors based on light-water-breeder (LWBR) Type 1 modules; light-water backfit prebreeder supplying advanced breeder; light-water backfit prebreeder/seed-blanket breeder system; and light-water backfit low-gain converter using medium-enrichment uranium, supplying a light-water backfit high-gain converter

  20. 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.)

  1. Chemical form of tritium released from solid breeder materials and the influences of it on a bred tritium recovery systems

    International Nuclear Information System (INIS)

    The ratio of HTO in total tritium was measured at release of the bred tritium to the purge gas with hydrogen using the thermal release after irradiation method, where neutron irradiation was performed at JRR-3 reactor in JAERI or KUR reactor in Kyoto University. It is experimentally confirmed in this study that not a small portion of bred tritium is released to the purge gas in the form of HTO form ceramic breeder materials even when hydrogen is added to the purge gas. The chemical composition is to be decided by the competitive reaction at the grain surface of a ceramic breeder material where desorption reaction, isotope exchange reaction 1, isotope exchange reaction 2 and water formation reaction are considered to take part. Observation in this study implies that it is necessary to have a bred tritium recovery system applicable for both HT and HTO form to recover whole bred tritium. The chemical composition also decides the amount of tritium transferable to the cooling water of the electricity generation system through the structural material in the blanket system. Permeation behavior of tritium through some structural materials at various conditions are also discussed. (author)

  2. 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

  3. Simulation of physical parameter for in-pipe tritium breeder

    International Nuclear Information System (INIS)

    It is necessary to build in-pipe tritium breeder in our country in order to assess breeder material of tritium breeder module (TBM) and to find the release law of tritium. The irradiation vessel is one of the key components of TBM. The physical parameters about in-pipe tritium breeder were simulated with MCNP code. The values of the self-shielding factor, equivalent cross-section, daily production of tritium and total heating power are separately 0.435, 1.09 x 10-22 cm2, 2.8 x 1010 Bq and 8.2 kW. And they would provide necessary data for designing the irradiation vessel. (authors)

  4. Fast-breeder-power reactor records in the INIS database

    International Nuclear Information System (INIS)

    This report presents a statistical analysis of more than 19,700 records of publications concerned with research and technology in the field of fast breeder power fission reactors which are included in the INIS Bibliographic Database for the period from 1970. to 1999. The main objectives of this bibliometric study were: to make an inventory of the fast breeder power reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of fast breeder power reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in fast breeder reactors research and technology. The quantitative data in this report are obtained for various properties of relevant INIS records such as year of publication, secondary subject categories, countries of publication, language, publication types, literary types, etc. (author)

  5. 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.

  6. A comprehensive model for the prediction of tritium behavior in solid breeder materials during steady-state and transient conditions

    International Nuclear Information System (INIS)

    In recent years, the area of tritium transport and release from Li-base ceramics in fusion blankets has become increasingly important particularly in conjunction with the growing amount of data available from in-pile tritium recovery experiments. Key variables that can strongly affect the tritium inventory and the kinetics of release, such as purge gas composition, temperature, solid breeder microstructure and activation energies for bulk diffusion and for desorption have been identified. Therefore, in the current phase of research and development, there is a strong incentive to develop comprehensive predictive capabilities in order to understand the new experimental data, to extrapolate these data to different ranges of conditions of interest, and to provide a necessary tool for fusion blanket design analysis. The objectives of this research are: (1) to develop new models for tritium transport in solid breeders to better describe the complex multistep phenomena that characterize tritium release, (2) to develop a computer code to predict tritium behavior, as a function of different variables and for a wide range of operating conditions, (3) to calibrate such models with existing experimental results. A comprehensive model is proposed. The sequence of transport processes leading to tritium release includes diffusion through the grain and along the grain boundaries, adsorption and desorption at the breeder surface and diffusion through the pore. A computer code called MISTRAL has been developed based on this model. The results obtained are in reasonable agreement with the experimental results, for the available set of property data, and indicate a fairly good predictive capability of the model for the analysis of several transients of interest for solid breeder fusion blankets

  7. Characterisation of thermal radiation in the near-wall region of a packed pebble bed / Maritza de Beer

    OpenAIRE

    De Beer, Maritza

    2014-01-01

    The heat transfer phenomena in the near-wall region of a randomly packed pebble bed are important in the design of a Pebble Bed Reactor (PBR), especially when considering the safety case during accident conditions. At higher temperatures the contribution of the radiation heat transfer component to the overall heat transfer in a PBR increases significantly. The wall effect present in the near-wall region of a packed pebble bed affects the heat transfer in this region. Various correlations e...

  8. Reprocessing of fast breeder reactor fuels in France

    International Nuclear Information System (INIS)

    The reprocessing of breeder reactor fuels is a direct technical descendant of the reprocessing of thermal reactor fuels which was developped first. The process used is in both cases the PUREX process, which consists in dissolution by nitric acid followed by selective extraction using TBP. In France, the application of this technique to breeder reactor fuels greatly benefited from the scientific and industrial experience initially acquired with metallic fuels of the MAGNOX type and then with oxide fuels of the LWR type

  9. Development of Liquid Type Breeder Technology for ITER-TBM

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Ki Sok; Hong, Bong Geun; Lee, Dong Won

    2008-07-15

    In relation to liquid type TBM technology development, various works are performed. We established a test loop concept to test the MHD effects and materials compatibility for the Pb-17Li breeder material. For the loop construction, electromagnetic pump and storage tank for the Pb-17Li loop was manufactured and some technical requirements are summarised. As a reference, technical literatures relevant to the liquid type TBM materials and the tritium extraction from breeder materials are also surveyed.

  10. 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.

  11. 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...

  12. The future of the Fast Breeder

    International Nuclear Information System (INIS)

    Fast Breeder Reactors (FBRs) can produce more fissile nuclei than they consume whilst, at the same time, generating energy using fast neutrons. By conversion of uranium isotope 238 into a fissionable fuel, FBRs provide over 60 times more energy than can be extracted from the uranium reserves by thermal reactors. Their development is therefore an essential objective in the next century, particularly for those industrialised countries that have little or no energy resources of their own. The European countries which have been engaged in the development of FBRs for more than 25 years have decided to collaborate in an advanced design, the European Fast Reactor (EFR) which uses the best of previous national projects and draws on extensive operating experience from FBR plants in Europe. The naturally safe characteristics and technological features of sodium-cooled Fast Reactors will be fully utilised in an EFR design which meets the same safety level as the Light Water Reactors (LWRs). Owing to technical progress and series construction effect, the EFR is expected to achieve competitiveness with contemporary LWRs with the higher capital cost of the Fast Reactor offset by its markedly lower fuel cycle cost. (author)

  13. 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.)

  14. Characteristics of microstructure and tritium release properties of different kinds of beryllium pebbles for application in tritium breeding modules

    Energy Technology Data Exchange (ETDEWEB)

    Kurinskiy, P.; Vladimirov, P.; Moeslang, A. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany). Inst. for Applied Materials - Applied Materials Physics (IAM-AWP); Rolli, R. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany). Inst. for Applied Materials - Materials Biomechanics (IAM-WBM); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, Barcelona (Spain)

    2013-07-01

    Beryllium pebbles with diameters of 1 mm are considered to be perspective material for the use as neutron multiplier in tritium breeding modules of fusion reactors. Up to now, the main concept of helium-cooled breeding blanket in ITER project foresees the use of 1 mm beryllium pebbles fabricated by company NGK, Japan. It is notable that beryllium pebbles of other types are commercially available at the market. Presented work is dedicated to a study of characteristics of microstructure, packaging density and parameters of tritium release of beryllium pebbles produced by Bochvar Institute, Russian Federation, and Company Materion, USA. (orig.).

  15. Rapid growth of gas-giant cores by pebble accretion

    CERN Document Server

    Lambrechts, Michiel

    2012-01-01

    The observed lifetimes of gaseous protoplanetary discs place strong constraints on gas and ice giant formation in the core accretion scenario. The approximately 10-Earth-mass solid core responsible for the attraction of the gaseous envelope has to form before gas dissipation in the protoplanetary disc is completed within 1-10 million years. Building up the core by collisions between km-sized planetesimals fails to meet this time-scale constraint, especially at wide stellar separations. Nonetheless, gas-giant planets are detected by direct imaging at wide orbital distances. In this paper, we numerically study the growth of cores by the accretion of cm-sized pebbles loosely coupled to the gas. We measure the accretion rate onto seed masses ranging from a large planetesimal to a fully grown 10-Earth-mass core and test different particle sizes. The numerical results are in good agreement with our analytic expressions, indicating the existence of two accretion regimes, one set by the azimuthal and radial particle ...

  16. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    Science.gov (United States)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  17. Industrial ceramics

    International Nuclear Information System (INIS)

    After having given the definition of the term 'ceramics', the author describes the different manufacturing processes of these compounds. These materials are particularly used in the fields of 1)petroleum industry (in primary and secondary reforming units, in carbon black reactors and ethylene furnaces). 2)nuclear industry (for instance UO2 and PuO2 as fuels; SiC for encapsulation; boron carbides for control systems..)

  18. Preliminary neutronic design of high burnup OTTO cycle pebble bed reactor

    International Nuclear Information System (INIS)

    The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM) loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble. (author)

  19. Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    T. Setiadipura

    2015-04-01

    Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

  20. 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)

  1. 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

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

    International Nuclear Information System (INIS)

    This document was prepared by the Office of the Program Director for Nuclear Energy, U.S. Department of Energy (USDOE). It sets forth the status and current activities for the development of fast breeder technology in the United States. In April 1977 the United States announced a change in its nuclear energy policy. Concern about the potential for the proliferation of nuclear weapons capability emerged as a major issue in considering whether to proceed with the development, demonstration and eventual deployment of breeder reactor energy systems. Plutonium recycle and the commercialization of the fast breeder were deferred indefinitely. This led to a reorientation of the nuclear fuel cycle program which was previously directed toward the commercialization of fuel reprocessing and plutonium recycle to the investigation of a full range of alternative fuel cycle technologies. Two major system evaluation programs, the Nonproliferation Alternative Systems Assessment Program (NASAP), which is domestic, and the International Nuclear Fuel Cycle Evaluation (INFCE), which is international, are assessing the nonproliferation advantages and other characteristics of advanced reactor concepts and fuel cycles. These evaluations will allow a decision in 1981 on the future direction of the breeder program. In the interim, the technologies of two fast breeder reactor concepts are being developed: the Liquid Metal Fast Breeder Reactor (LMFBR) and the Gas Cooled Fast Reactor (CFR). The principal goals of the fast breeder program are: LMFBR - through a strong R and D program, consistent with US nonproliferation objectives and anticipated national electric energy requirements, maintain the capability to commit to a breeder option; investigate alternative fuels and fuel cycles that might offer nonproliferation advantages; GCFR - provide a viable alternative to the LMFBR that will be consistent with the developing U.S. nonproliferation policy; provide GCFR technology and other needed

  3. Penn State geoPebble system: Design,Implementation, and Initial Results

    Science.gov (United States)

    Urbina, J. V.; Anandakrishnan, S.; Bilen, S. G.; Fleishman, A.; Burkett, P.

    2014-12-01

    The Penn State geoPebble system is a new network of wirelessly interconnected seismic and GPS sensor nodes with flexible architecture. This network will be used for studies of ice sheets in Antarctica and Greenland, as well as to investigate mountain glaciers. The network will consist of ˜150 geoPebbles that can be deployed in a user-defined spatial geometry. We present our design methodology, which has enabled us to develop these state-of- the art sensors using commercial-off-the-shelf hardware combined with custom-designed hardware and software. Each geoPebble is a self- contained, wirelessly connected sensor for collecting seismic measurements and position information. Key elements of each node encompasses a three-component seismic recorder, which includes an amplifier, filter, and 24- bit analog-to-digital converter that can sample up to 10 kHz. Each unit also includes a microphone channel to record the ground-coupled airwave. The timing for each node is available from GPS measurements and a local precision oscillator that is conditioned by the GPS timing pulses. In addition, we record the carrier-phase measurement of the L1 GPS signal in order to determine location at sub-decimeter accuracy (relative to other geoPebbles within a few kilometers radius). Each geoPebble includes 16 GB of solid-state storage, wireless communications capability to a central supervisory unit, and auxiliary measurements capability (including tilt from accelerometers, absolute orientation from magnetometers and temperature). A novel aspect of the geoPebble is a wireless charging system for the internal battery (using inductive coupling techniques). The geoPebbles include all the sensors (geophones, GPS, microphone), communications (WiFi), and power (battery and charging) internally, so the geoPebble system can operate without any cabling connections (though we do provide an external connector so that different geophones can be used). We report initial field-deployment results and

  4. Single-phase convection heat transfer characteristics of pebble-bed channels with internal heat generation

    International Nuclear Information System (INIS)

    Graphical abstract: The core of the water-cooled pebble bed reactor is the porous channels which stacked with spherical fuel elements. The gaps between the adjacent fuel elements are complex because they are stochastic and often shift. We adopt electromagnetic induction heating method to overall heat the pebble bed. By comparing and analyzing the experimental data, we get the rule of power distribution and the rule of heat transfer coefficient with particle diameter, heat flux density, inlet temperature and working fluid's Re number. Highlights: ► We adopt electromagnetic induction heating method to overall heat the pebble bed to be the internal heat source. ► The ball diameter is smaller, the effect of the heat transfer is better. ► With Re number increasing, heat transfer coefficient is also increasing and eventually tends to stabilize. ► The changing of heat power makes little effect on the heat transfer coefficient of pebble bed channels. - Abstract: The reactor core of a water-cooled pebble bed reactor includes porous channels that are formed by spherical fuel elements. This structure has notably improved heat transfer. Due to the variability and randomness of the interstices in pebble bed channels, heat transfer is complex, and there are few studies regarding this topic. To study the heat transfer characters of pebble bed channels with internal heat sources, oxidized stainless steel spheres with diameters of 3 and 8 mm and carbon steel spheres with 8 mm diameters are used in a stacked pebble bed. Distilled water is used as a refrigerant for the experiments, and the electromagnetic induction heating method is used to heat the pebble bed. By comparing and analyzing the experimental results, we obtain the governing rules for the power distribution and the heat transfer coefficient with respect to particle diameter, heat flux density, inlet temperature and working fluid Re number. From fitting of the experimental data, we obtain the dimensionless average

  5. 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

  6. Fast breeder physics and nuclear core design

    International Nuclear Information System (INIS)

    This report gathers the papers that have been presented on January 18/19, 1983 at a seminar ''Fast breeder physics and nuclear core design'' held at KfK. These papers cover the results obtained within about the last five years in the r+d program and give some indication, what still has to be done. To begin with, the ''tools'' of the core designer, i.e. nuclear data and neutronics codes are covered in a comprehensive way, the seminar emphasized the applications, however. First of all the accuracies obtained for the most important parameters are presented for the design of homogeneous and heterogeneous cores of about 1000 MWe, they are based on the results of critical experiments. This is followed by a survey on activities related to the KNK II reactor, i.e. calculations concerning a modification of the core as well as critical experiments done with respect to re-loads. Finally, work concerning reactivity worths of accident configurations is presented: the generation of reactivity worths for the input of safety-related calculations of a SNR 2 design, and critical experiments to investigate the requirements for the codes to be used for these calculations. These papers are accompanied by two contributions from the industrial partners. The first one deals with the requirements to nuclear design methods as seen by the reactor designer and then shows what has been achieved. The latter one presents state, trends, and methods of the SNR 2 design. The concluding remarks compare the state of the art reached within DeBeNe with international achievements. (orig.)

  7. 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

  8. 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)

  9. The importance of the AVR pebble-bed reactor for the future of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Pohl, P. [Arbeitsgemeinschaft Versuchsreaktor AVR GmbH, Postfach 1160, 52412 Juelich (Germany)

    2006-07-01

    The AVR pebble-bed high temperature gas-cooled reactor (HTGR) at Juelich (Germany)) operated from 1967 to 1988 and was certainly the most important HTGR project of the past. The reactor was the mass test bed for all development steps of HTGR pebble fuel. Some early fuel charges failed under high temperature conditions and contaminated the reactor. An accurate pebble measurement (Cs 137) allowed to clean the core from unwanted pebbles after 1981. The coolant activity went down and remained very low for the remaining reactor operation. A melt-wire experiment in 1986 revealed max. coolant temperatures of >1280 deg. C and fuel temperatures of >1350 deg. C, explained by under-estimated bypasses. The fuel still in the core achieved high burn-ups and showed under the extreme temperature conditions excellent fission product retention. Thus, the AVR operation qualified the HTGR fuel, and an average discharge burn-up of 112% fifa revealed an excellent fuel economy of the pebble-bed reactor. Furthermore, the AVR operation offers many meaningful data for code-to-experiment comparisons. (authors)

  10. Temperature transients of a fusion-fission ITER pebble bed reactor in loss of coolant accident

    International Nuclear Information System (INIS)

    In this preliminary scoping study, post-accident temperature transients of several fusion-fission designs utilizing ITER-FEAT-like parameters and fission pebble bed fuel technology are examined using a 1-D cylindrical MATLAB heat transfer code along with conventional fission decay heat approximations. Scenarios studied include systems with no additional passive safety features to systems with melting reflectors designed to increase emissivity after reaching a specified temperature. Results show that for a total fission power of ∼1400-2800 MW, two of the realistic variants investigated are passively safe. The crucial time, defined as the time when either any structural part of the fusion-fission tokamak reaches melting point, or when the pebble fuel reaches 1873 K, ranges from 5.7 to 76 h for the unsafe configurations. Additionally, it is illustrated that, fundamentally, the LOCA characteristics of pure fission pebble beds and fusion-fission pebble beds are different. Namely, the former depends on the pebble fuel's large thermal capacity, along with external radiation and natural convective cooling, while the latter depends significantly more on the tokamak's sizeable total internal heat capacity. This difference originates from the fusion-fission reactor's conflicting goal of having to minimize heat transfer to the magnets during normal operation. These results are discussed in the context of overall fusion-fission reactor design and safety

  11. Stability and convergence analysis of the quasi-dynamics method for the initial pebble packing

    International Nuclear Information System (INIS)

    The simulation for the pebble flow recirculation within Pebble Bed Reactors (PBRs) requires an efficient algorithm to generate an initial overlap-free pebble configuration within the reactor core. In the previous work, a dynamics-based approach, the Quasi-Dynamics Method (QDM), has been proposed to generate densely distributed pebbles in PBRs with cylindrical and annular core geometries. However, the stability and the efficiency of the QDM were not fully addressed. In this work, the algorithm is reformulated with two control parameters and the impact of these parameters on the algorithm performance is investigated. Firstly, the theoretical analysis for a 1-D packing system is conducted and the range of the parameter in which the algorithm is convergent is estimated. Then, this estimation is verified numerically for a 3-D packing system. Finally, the algorithm is applied to modeling the PBR fuel loading configuration and the convergence performance at different packing fractions is presented. Results show that the QDM is efficient in packing pebbles within the realistic range of the packing fraction in PBRs, and it is capable in handling cylindrical geometry with packing fractions up to 63.5%. (authors)

  12. On the water delivery to terrestrial embryos by ice pebble accretion

    CERN Document Server

    Sato, Takao; Ida, Shigeru

    2015-01-01

    Standard accretion disk models suggest that the snow line in the solar nebula migrated interior to the Earth's orbit in a late stage of nebula evolution. In this late stage, a significant amount of ice could have been delivered to 1 AU from outer regions in the form of mm to dm-sized "pebbles." This raises the question why the present Earth is so depleted of water (with the ocean mass being as small as 0.023% of the Earth mass). Here we quantify the amount of icy pebbles accreted by terrestrial embryos after the migration of the snow line assuming that no mechanism halts the pebble flow in outer disk regions. We use a simplified version of the coagulation equation to calculate the formation and radial inward drift of icy pebbles in a protoplanetary disk. The pebble accretion cross section of an embryo is calculated using analytic expressions presented by recent studies. We find that the final mass and water content of terrestrial embryos strongly depends on the radial extent of the gas disk, the strength of d...

  13. On the numerical assessment of the thermo-mechanical performances of the DEMO Helium-Cooled Pebble Bed breeding blanket module

    International Nuclear Information System (INIS)

    Highlights: • HCPB blanket module thermo-mechanical behavior has been investigated under normal operation and over-pressurization steady state scenarios. • A theoretical–computational approach based on the finite element method (FEM) has been followed, adopting a qualified commercial FEM code. • Under normal operation scenario, SDC-IC safety rule relevant to the loss of ductility is not fulfilled in the FW and in the hot spots of SPv. • Under over-pressurization scenario, SDC-IC safety rule relevant to the loss of ductility is not met in the hot spots of lower and upper SPv. - Abstract: Within the framework of the European DEMO Breeder Blanket Programme, a research campaign has been launched by University of Palermo, ENEA-Brasimone and Karlsruhe Institute of Technology to theoretically investigate the thermo-mechanical behavior of the Helium-Cooled Pebble Bed (HCPB) breeding blanket module of the DEMO1 blanket vertical segment, under normal operation and over-pressurization loading scenarios. The research campaign has been carried out following a theoretical–computational approach based on the finite element method (FEM) and adopting a qualified commercial FEM code. A realistic 3D FEM model of the HCPB blanket module central poloidal–radial region has been developed, including one breeder cell in the toroidal direction and all the five cells in the poloidal one. No Breeder Units have been modeled, their presence being simulated by effective thermo-mechanical loads. Two sets of uncoupled steady state thermo-mechanical analyses have been carried out with reference to the investigated loading scenarios. In particular, under normal operation scenario (level A) the module has been supposed to undergo both 8 MPa coolant pressure on its cooling channel walls and thermal deformations due to the flat-top plasma operational state thermal field, while under over-pressurization scenario (level D) it has been assumed to experience 8 MPa coolant pressure on its

  14. A panoptic model for planetesimal formation and pebble delivery

    Science.gov (United States)

    Krijt, S.; Ormel, C. W.; Dominik, C.; Tielens, A. G. G. M.

    2016-02-01

    Context. The journey from dust particle to planetesimal involves physical processes acting on scales ranging from micrometers (the sticking and restructuring of aggregates) to hundreds of astronomical units (the size of the turbulent protoplanetary nebula). Considering these processes simultaneously is essential when studying planetesimal formation. Aims: The goal of this work is to quantify where and when planetesimal formation can occur as the result of porous coagulation of icy grains and to understand how the process is influenced by the properties of the protoplanetary disk. Methods: We develop a novel, global, semi-analytical model for the evolution of the mass-dominating dust particles in a turbulent protoplanetary disk that takes into account the evolution of the dust surface density while preserving the essential characteristics of the porous coagulation process. This panoptic model is used to study the growth from sub-micron to planetesimal sizes in disks around Sun-like stars. Results: For highly porous ices, unaffected by collisional fragmentation and erosion, rapid growth to planetesimal sizes is possible in a zone stretching out to ~10 AU for massive disks. When porous coagulation is limited by erosive collisions, the formation of planetesimals through direct coagulation is not possible, but the creation of a large population of aggregates with Stokes numbers close to unity might trigger the streaming instability (SI). However, we find that reaching conditions necessary for SI is difficult and limited to dust-rich disks, (very) cold disks, or disks with weak turbulence. Conclusions: Behind the snow-line, porosity-driven aggregation of icy grains results in rapid (~104 yr) formation of planetesimals. If erosive collisions prevent this, SI might be triggered for specific disk conditions. The numerical approach introduced in this work is ideally suited for studying planetesimal formation and pebble delivery simultaneously and will help build a coherent

  15. Monolithic ceramics

    Science.gov (United States)

    Herbell, Thomas P.; Sanders, William A.

    1992-01-01

    A development history and current development status evaluation are presented for SiC and Si3N4 monolithic ceramics. In the absence of widely sought improvements in these materials' toughness, and associated reliability in structural applications, uses will remain restricted to components in noncritical, nonman-rated aerospace applications such as cruise missile and drone gas turbine engine components. In such high temperature engine-section components, projected costs lie below those associated with superalloy-based short-life/expendable engines. Advancements are required in processing technology for the sake of fewer and smaller microstructural flaws.

  16. Ceramic composition

    International Nuclear Information System (INIS)

    Improved ceramic compositions useful for cutting tools and the like are described. They are composed of an essentially homogeneous admixture of sintered powders of an aluminum oxide base material with other refractories including zirconium oxide, titanium oxide, hafnium oxide, titanium nitride, zirconium nitride, and tungsten or molybdenum carbide. In addition to their common and improved properties of hardness and strength, many of these compositions may be made by simple cold-pressing and sintering procedures. This avoids the known drawbacks of conventional hot press production

  17. Pebble Accretion Rates for Planetesimals: Hydrodynamics Calculations with Direct Particle Integration

    Science.gov (United States)

    Hughes, Anna; Boley, Aaron

    2015-12-01

    The formation and growth of planetesimals are fundamental to planet building. However, in our understanding of planet formation, there are a number of processes that limit the formation of planetesimals such as particle bouncing, fragmentation, and inward radial drift due to gas drag. Such processes seemingly make growth beyond mm to cm sizes difficult. In this case, the protoplanetary disk may become rich in pebble-sized solids as opposed to km-sized planetesimals. If a small number of large planetesimals do manage to form, then gas-drag effects can allow those seeds to efficiently accrete the abundant pebbles from the nebula and grow to planet sizes. We present self-consistent hydrodynamic simulations with direct particle integration and gas-drag coupling to evaluate the rate of planetesimal growth due to pebble accretion. We explore a range of particle sizes and nebular conditions using wind tunnel numerical experiments.

  18. The growth of planets by pebble accretion in evolving protoplanetary discs

    CERN Document Server

    Bitsch, Bertram; Johansen, Anders

    2015-01-01

    The formation of planets depends on the underlying protoplanetary disc structure, which influences both the accretion and migration rates of embedded planets. The disc itself evolves on time-scales of several Myr during which both temperature and density profiles change as matter accretes onto the central star. Here we use a detailed model of an evolving disc to determine the growth of planets by pebble accretion and their migration through the disc. Cores that reach their pebble isolation mass accrete gas to finally form giant planets with extensive gas envelopes, while planets that do not reach pebble isolation mass are stranded as ice giants and ice planets containing only minor amounts of gas in their envelopes. Unlike earlier population synthesis models, our model works without any artificial reductions in migration speed and for protoplanetary discs with gas and dust column densities similar to those inferred from observations. We find that in our nominal disc model the emergence of planetary embryos pr...

  19. Study of the fracture behavior of mortar and concretes with crushed rock or pebble aggregates

    Directory of Open Access Journals (Sweden)

    Sebastião Ribeiro

    2011-03-01

    Full Text Available The objective of this work was to compare the fracture energy of mortar and concretes produced with crushed rock and pebble aggregates using zero, 10, 20, 30 and 40% of aggregates mixed with standard mortar and applying the wedge splitting method to achieve stable crack propagation. The samples were cast in a special mold and cured for 28 days, after which they were subjected to crack propagation tests by the wedge splitting method to determine the fracture energies of the mortar and concrete. The concretes showed higher fracture energy than the mortar, and the concretes containing crushed rock showed higher resistance to crack propagation than all the compositions containing pebbles. The fracture energy varied from 38 to 55 J.m-2. A comparison of the number of aggregates that separated from the two concrete matrices with the highest fracture energies indicated that the concrete containing pebbles crumbled more easily and was therefore less resistant to crack propagation.

  20. Characteristics of microstructure and tritium release properties of different kinds of beryllium pebbles for application in tritium breeding modules

    Energy Technology Data Exchange (ETDEWEB)

    Kurinskiy, P., E-mail: petr.kurinskiy@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Vladimirov, P.; Moeslang, A. [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Rolli, R. [Karlsruhe Institute of Technology, Institute for Applied Materials – Materials and Biomechanics (IAM-WBM), P.O. Box 3640, Karlsruhe 76021 (Germany); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, Barcelona 08019 (Spain)

    2014-10-15

    Highlights: • Tritium release properties and characteristics of microstructure of beryllium pebbles having different sizes of grains were studied. • Fine-grained beryllium pebbles showed the best ability to release tritium compared to pebbles from another charges. • Be pebbles with the grain sizes exceeding 100 μm contain a great number of small pores and inclusions presumably referring to the history of material fabrication. • The sizes of grains are one of a key characteristic of microstructure which influences the parameters of tritium release. - Abstract: Beryllium pebbles with diameters of 1 mm are considered to be perspective material for the use as neutron multiplier in tritium breeding modules of fusion reactors. Up to now, the design of helium-cooled breeding blanket in ITER project foresees the use of 1 mm beryllium pebbles fabricated by NGK Insulators Ltd., Japan. It is notable that beryllium pebbles from Russian Federation and USA are also available and the possibility of their large-scale fabrication is under study. Presented work is dedicated to a study of characteristics of microstructure and parameters of tritium release of beryllium pebbles produced by Bochvar Institute, Russian Federation, and Materion Corporation, USA.

  1. Characteristics of microstructure and tritium release properties of different kinds of beryllium pebbles for application in tritium breeding modules

    International Nuclear Information System (INIS)

    Highlights: • Tritium release properties and characteristics of microstructure of beryllium pebbles having different sizes of grains were studied. • Fine-grained beryllium pebbles showed the best ability to release tritium compared to pebbles from another charges. • Be pebbles with the grain sizes exceeding 100 μm contain a great number of small pores and inclusions presumably referring to the history of material fabrication. • The sizes of grains are one of a key characteristic of microstructure which influences the parameters of tritium release. - Abstract: Beryllium pebbles with diameters of 1 mm are considered to be perspective material for the use as neutron multiplier in tritium breeding modules of fusion reactors. Up to now, the design of helium-cooled breeding blanket in ITER project foresees the use of 1 mm beryllium pebbles fabricated by NGK Insulators Ltd., Japan. It is notable that beryllium pebbles from Russian Federation and USA are also available and the possibility of their large-scale fabrication is under study. Presented work is dedicated to a study of characteristics of microstructure and parameters of tritium release of beryllium pebbles produced by Bochvar Institute, Russian Federation, and Materion Corporation, USA

  2. Challenges in forming the solar system's giant planet cores via pebble accretion

    Energy Technology Data Exchange (ETDEWEB)

    Kretke, K. A.; Levison, H. F., E-mail: kretke@boulder.swri.edu [Southwest Research Institute, 1050 Walnut Street, Suite 300, Boulder, CO 80302 (United States)

    2014-12-01

    Though ∼10 M {sub ⊕} mass rocky/icy cores are commonly held as a prerequisite for the formation of gas giants, theoretical models still struggle to explain how these embryos can form within the lifetimes of gaseous circumstellar disks. In recent years, aerodynamic-aided accretion of 'pebbles', objects ranging from centimeters to meters in size, has been suggested as a potential solution to this long-standing problem. While pebble accretion has been demonstrated to be extremely effective in local simulations that look at the detailed behavior of these pebbles in the vicinity of a single planetary embryo, to date there have been no global simulations demonstrating the effectiveness of pebble accretion in a more complicated, multi-planet environment. Therefore, we have incorporated the aerodynamic-aided accretion physics into LIPAD, a Lagrangian code that can follow the collisional/accretional/dynamical evolution of a protoplanetary system, to investigate how pebble accretion manifests itself in the larger planet formation picture. We find that under generic circumstances, pebble accretion naturally leads to an 'oligarchic' type of growth in which a large number of planetesimals grow to similar-sized planets. In particular, our simulations tend to form hundreds of Mars- and Earth-mass objects between 4 and 10 AU. While merging of some oligarchs may grow massive enough to form giant planet cores, leftover oligarchs lead to planetary systems that cannot be consistent with our own solar system. We investigate various ideas presented in the literature (including evaporation fronts and planet traps) and find that none easily overcome this tendency toward oligarchic growth.

  3. Challenges in forming the solar system's giant planet cores via pebble accretion

    International Nuclear Information System (INIS)

    Though ∼10 M ⊕ mass rocky/icy cores are commonly held as a prerequisite for the formation of gas giants, theoretical models still struggle to explain how these embryos can form within the lifetimes of gaseous circumstellar disks. In recent years, aerodynamic-aided accretion of 'pebbles', objects ranging from centimeters to meters in size, has been suggested as a potential solution to this long-standing problem. While pebble accretion has been demonstrated to be extremely effective in local simulations that look at the detailed behavior of these pebbles in the vicinity of a single planetary embryo, to date there have been no global simulations demonstrating the effectiveness of pebble accretion in a more complicated, multi-planet environment. Therefore, we have incorporated the aerodynamic-aided accretion physics into LIPAD, a Lagrangian code that can follow the collisional/accretional/dynamical evolution of a protoplanetary system, to investigate how pebble accretion manifests itself in the larger planet formation picture. We find that under generic circumstances, pebble accretion naturally leads to an 'oligarchic' type of growth in which a large number of planetesimals grow to similar-sized planets. In particular, our simulations tend to form hundreds of Mars- and Earth-mass objects between 4 and 10 AU. While merging of some oligarchs may grow massive enough to form giant planet cores, leftover oligarchs lead to planetary systems that cannot be consistent with our own solar system. We investigate various ideas presented in the literature (including evaporation fronts and planet traps) and find that none easily overcome this tendency toward oligarchic growth.

  4. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

  5. An analysis of the thermal behaviour of pebble-bed nuclear reactors in the case of emergencies

    International Nuclear Information System (INIS)

    In this paper, the performance of pebble-bed nuclear reactors under very severe emergencies will be analysed. Calculated hypotheses take into consideration total failure of decay heat removal systems and any other active equipment, including electric power supply. It has been shown that pebble temperatures will remain well below safety limits if the reactor design embodies a core catcher with a passive cooling reservoir and a pebble draining system which would be naturally activated by a lack of a power supply. Although these features apply to any pebble-bed reactor, particular attention is paid to accelerator-driven sub-critical assemblies, where reactivity noise produced pebble quivering has a practical negligible effect. (authors)

  6. Apply Burnable Poison For Fuel Pebble Of PBMR-400 With OTTO Refueling

    International Nuclear Information System (INIS)

    A new fuel pebble was designed by adding spherical Gd2O3 particles for obtaining the minimum reactivity swing. Optimization is done in a lattice model to determine the combination of radius and number of burnable poison (BP) particles per pebble to obtain the minimum reactivity swing. The numerical calculation so that with 740 μm and 13 particles of Gd2O3. The reactivity swing is reduced from 38% to 2.0%, whereas the k∞ is 1.06 - 1.08 for a fuel lattice with the target burnup of 55 GWd/t. (author)

  7. Absorber rod for nuclear reactors in a pebble bed of spherical operating elements

    International Nuclear Information System (INIS)

    The claim refers to the constructional configuration of an absorber rod, whose and penetrating into the pebble bed has an opening to reduce the fracture rate, so that the operating elements can escape into a channel within the absorber rod. To suit this to the direction of movement of the elements a part of the end of the rod is flexibly connected to the hollow absorber rod via a joint. In this way the mechanical load of the element particles is reduced and simultaneously one achieves that much lower force is required to insert the absorber rod into the pebble bed. (UA)

  8. Hydrologic and Aquatic Species Implications of the Proposed Pebble Mine, Bristol Bay, Alaska

    OpenAIRE

    Cundy, Fiona

    2012-01-01

    Bristol Bay, Alaska is one of the last ecosystems left on earth that haqs gone unaltered by human impacts.  Bristol Bay watershed supports the largest wild sockeye salmon runs on the planet with nearly 42 million salmon migrating to the watersheds headwaters every year.  The proposed Pebble Mine, containing gold, copper, and molybdenum has threatened the health of this watershed.  This project asks what effects the proposed Pebble Mine will have on water quality and quantity, and more specifi...

  9. Optimization of a radially cooled pebble bed reactor - HTR2008-58117

    International Nuclear Information System (INIS)

    By altering the coolant flow direction in a pebble bed reactor from axial to radial, the pressure drop can be reduced tremendously. In this case the coolant flows from the outer reflector through the pebble bed and finally to flow paths in the inner reflector. As a consequence, the fuel temperatures are elevated due to the reduced heat transfer of the coolant. However, the power profile and pebble size in a radially cooled pebble bed reactor can be optimized to achieve lower fuel temperatures than current axially cooled designs, while the low pressure drop can be maintained. The radial power profile in the core can be altered by adopting multi-pass fuel management using several radial fuel zones in the core. The optimal power profile yielding a flat temperature profile is derived analytically and is approximated by radial fuel zoning. In this case, the pebbles pass through the outer region of the core first and each consecutive pass is located in a fuel zone closer to the inner reflector. Thereby, the resulting radial distribution of the fissile material in the core is influenced and the temperature profile is close to optimal. The fuel temperature in the pebbles can be further reduced by reducing the standard pebble diameter from 6 cm to a value as low as I cm. An analytical investigation is used to demonstrate the effects on the fuel temperature and pressure drop for both radial and axial cooling. Finally, two-dimensional numerical calculations were performed, using codes for neutronics, thermal-hydraulics and fuel depletion analysis, in order to validate the results for the optimized design that were obtained from the analytical investigations. It was found that for a radially cooled design with an optimized power profile and reduced pebble diameter (below 3.5 cm) both a reduction in the pressure drop (Δp = -2.6 bar), which increases the reactor efficiency with several percent, and a reduction in the maximum fuel temperature (ΔT = -50 deg. C) can be achieved

  10. In-core fuel management optimization of a Very High Temperature pebble-bed Reactor

    International Nuclear Information System (INIS)

    A new calculation procedure was developed to reduce the power peak in the core of a Very High Temperature pebble-bed Reactor. The procedure consists in several coupled computational codes, which are used iteratively until convergence is reached. This procedure combines the fuel depletion and the neutronic behavior of the fuel in the reactor core, modeling once-through-then-out cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times, obtaining the asymptotic fuel-loading pattern directly, without any intermediate loading pattern. (Author)

  11. 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

  12. The design of the Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    India has a moderate uranium reserve and a large thorium reserve. The primary energy resource for electricity generation in the country is coal. The potential of other resources like gas, oil, wind, solar and biomass is very limited. The only viable and sustainable resource is the nuclear energy. Presently, Pressurised Heavy Water Reactors utilizing natural uranium are in operation/under construction and the plutonium generated from these reactors will be multiplied through breeding in fast breeder reactors. The successful construction, commissioning and operation of Fast Breeder Test Reactor at Kalpakkam has given confidence to embark on the construction of the Prototype Fast Breeder Reactor (PFBR). This paper describes the salient design features of PFBR including the design of the reactor core, reactor assembly, main heat transport systems, component handling, steam water system, electrical power systems, instrumentation and control, plant layout, safety and research and development

  13. 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.)

  14. 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.)

  15. 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

  16. 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

  17. 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.)

  18. Environmental durability of ceramics and ceramic composites

    Science.gov (United States)

    Fox, Dennis S.

    1992-01-01

    An account is given of the current understanding of the environmental durability of both monolithic ceramics and ceramic-matrix composites, with a view to the prospective development of methods for the characterization, prediction, and improvement of ceramics' environmental durability. Attention is given to the environmental degradation behaviors of SiC, Si3N4, Al2O3, and glass-ceramic matrix compositions. The focus of corrosion prevention in Si-based ceramics such as SiC and Si3N4 is on the high and low sulfur fuel combustion-product effects encountered in heat engine applications of these ceramics; sintering additives and raw material impurities are noted to play a decisive role in ceramics' high temperature environmental response.

  19. 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

  20. 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...

  1. 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.

  2. 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)

  3. 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

  4. 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

  5. Effect of low-density diets on broiler breeder and offspring performance

    OpenAIRE

    Enting, H.

    2005-01-01

    Restricted feeding of broiler breeders is required to obtain good reproductive performance. Current practical feed restriction levels can result in hunger feeling and chronic stress, particularly during the rearing period. On the basis of literature data, low-density diets might improve bird welfare. Recent findings also indicate that low-density broiler breeder diets can reduce offspring mortality. In this thesis, effects of low-density breeder diets on bird welfare and breeder and offspring...

  6. Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

    International Nuclear Information System (INIS)

    The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

  7. Separating gas-giant and ice-giant planets by halting pebble accretion

    CERN Document Server

    Lambrechts, Michiel; Morbidelli, Alessandro

    2014-01-01

    In the Solar System giant planets come in two flavours: 'gas giants' (Jupiter and Saturn) with massive gas envelopes and 'ice giants' (Uranus and Neptune) with much thinner envelopes around their cores. It is poorly understood how these two classes of planets formed. High solid accretion rates, necessary to form the cores of giant planets within the life-time of protoplanetary discs, heat the envelope and prevent rapid gas contraction onto the core, unless accretion is halted. We find that, in fact, accretion of pebbles (~ cm-sized particles) is self-limiting: when a core becomes massive enough it carves a gap in the pebble disc. This halt in pebble accretion subsequently triggers the rapid collapse of the super-critical gas envelope. As opposed to gas giants, ice giants do not reach this threshold mass and can only bind low-mass envelopes that are highly enriched by water vapour from sublimated icy pebbles. This offers an explanation for the compositional difference between gas giants and ice giants in the S...

  8. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  9. Catalyst functionalized buffer sorbent pebbles for rapid separation of carbon dioxide from gas mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Aines, Roger D

    2015-03-31

    A method for separating CO.sub.2 from gas mixtures uses a slurried media impregnated with buffer compounds and coating the solid media with a catalyst or enzyme that promotes the transformation of CO.sub.2 to carbonic acid. Buffer sorbent pebbles with a catalyst or enzyme coating are provided for rapid separation of CO.sub.2 from gas mixtures.

  10. Catalyst functionalized buffer sorbent pebbles for rapid separation of carbon dioxide from gas mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Aines, Roger D.

    2013-03-12

    A method for separating CO.sub.2 from gas mixtures uses a slurried media impregnated with buffer compounds and coating the solid media with a catalyst or enzyme that promotes the transformation of CO.sub.2 to carbonic acid. Buffer sorbent pebbles with a catalyst or enzyme coating are provided for rapid separation of CO.sub.2 from gas mixtures.

  11. Thermal ramp tritium release in COBRA-1A2 C03 beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, D.L. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    Tritium release kinetics, using the method of thermal ramp heating at three linear ramp rates, were measured on the COBRA-1A2 C03 1-mm beryllium pebbles. This report includes a brief discussion of the test, and the test data in graph format.

  12. 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

  13. 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.

  14. Discussion of the ERDA breeder program in the USA

    International Nuclear Information System (INIS)

    The ERDA breeder program in the United States has been drastically cut back by the Carter Administration. Because of its large reserves of uranium and coal the U.S. feels it is able to postpone considerably the development of commercial fast breeder power plants. This policy is also very much a consequence of the proliferation threat emphasized by the Carter Administration in connection with reprocessing and a plutonium economy. For the same reason, also the other industrialized nations have been requested to rearrange their fast breeder activities accordingly. However, the situation with respect to reserves of primary energy in the Federal Republic, in the whole of Western Europe and also in Japan compels these countries to make provision for reprocessing spent nuclear fuels, using plutonium as a nuclear fuel and last but not least, the determined development of commercial LMFBR power plants. In addition, an argument used against the new American nuclear policy, both in the United States and in other industrialized countries, is the contention that there is no such thing as an absolutely proliferation-proof fuel cycle and that this objective could not even be attained by a halt to fast breeder development. (orig.)

  15. Breeder Spent Fuel Handling Program multipurpose cask design basis document

    International Nuclear Information System (INIS)

    The Breeder Spent Fuel Handling (BSFH) Program multipurpose cask Design Basis Document defines the performance requirements essential to the development of a legal weight truck cask to transport FFTF spent fuel from reactor to a reprocessing facility and the resultant High Level Waste (HLW) to a repository. 1 ref

  16. 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

  17. 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

  18. 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

  19. Conceptual design study for a mirror fusion breeder

    International Nuclear Information System (INIS)

    A mirror fusion breeder, CHD, has been designed for providing plenty of nuclear fuel for light water reactors to meet the needs for rapid development of nuclear power in the first half of next century. The breeder is able to support the nuclear fuel needs for more than 10 LWRs of equal scale in power with fuel enriched directly in CHD without reprocessing. Measures are taken to flatten the power density distribution in the blanket so that fission is suppressed in the region close to the plasma, and by this way fuel production is enhanced for this direct enriched fusion breeder. In order to reduce the MHD pressure drop, LiPb flows in the blanket axially. Though the tritium inventory in the reactor is very low, special material and design have to be developed to reduce the permeation of tritium through the coolant pipes. The cost of electricity from the system, consisting of 11 LWR plants and one fusion breeder is predicted to be 1.05 times of that from a traditional LWR plant. This figure is insensitive both to the cost of CHD and its support ratio

  20. Dental ceramics: An update

    Directory of Open Access Journals (Sweden)

    Shenoy Arvind

    2010-01-01

    Full Text Available 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. Examples of the microstructure property relationships for these ceramic materials will also be addressed.

  1. Thermo-mechanical and neutron lifetime modelling and design of Be pebbles in the neutron multiplier for the LIFE engine

    International Nuclear Information System (INIS)

    Concept designs for the laser inertial fusion/fission energy (LIFE) engine include a neutron multiplication blanket containing Be pebbles flowing in a molten salt coolant. These pebbles must be designed to withstand the extreme irradiation and temperature conditions in the blanket to enable a reliable and cost-effective operation of LIFE. In this work, we develop design criteria for spherical Be pebbles on the basis of their thermo-mechanical behaviour under continued neutron exposure. We consider the effects of high fluence and fast fluxes on the elastic, thermal and mechanical properties of nuclear-grade Be. Our results suggest a maximum pebble diameter of 30 mm to avoid tensile failure, coated with an anti-corrosive, high-strength metallic shell to avoid failure by pebble contact. Moreover, we find that the operation temperature must always be kept above 450 deg. C to enable creep to relax the stresses induced by swelling. Under these circumstances, we estimate the pebble lifetime to be at least 16 months if uncoated, and up to six years when coated. We identify the sources of uncertainty on the properties used and discuss the advantages of new intermetallic beryllides and their use in LIFE's neutron multiplier. To establish Be-pebble lifetimes with improved confidence, reliable experiments to measure irradiation creep must be performed.

  2. 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

  3. Further adaptation of the European ceramic-B.I.T. blanket conceptual design to updated Demo specifications

    International Nuclear Information System (INIS)

    This paper presents the recent development studies on the adaptation of the European Ceramic Solid Breeder Inside Tube (BIT) Blanket to updated DEMO specifications. The adaptation work is in progress, since 1990, when a detailed comparison between two existing designs lead to the selection of an unique concept. The main new developments concern the separation in two parts of the inboard blanket segments at the level of the lower divertor, the consequent improvement of the blanket coverage, the simplification of maintenance operations, and finally the increased compactness of the blanket because of the inclusion of the shielding into the breeder assembly

  4. Growing the gas-giant planets by the gradual accumulation of pebbles.

    Science.gov (United States)

    Levison, Harold F; Kretke, Katherine A; Duncan, Martin J

    2015-08-20

    It is widely held that the first step in forming gas-giant planets, such as Jupiter and Saturn, was the production of solid 'cores' each with a mass roughly ten times that of the Earth. Getting the cores to form before the solar nebula dissipates (in about one to ten million years; ref. 3) has been a major challenge for planet formation models. Recently models have emerged in which 'pebbles' (centimetre-to-metre-sized objects) are first concentrated by aerodynamic drag and then gravitationally collapse to form objects 100 to 1,000 kilometres in size. These 'planetesimals' can then efficiently accrete left-over pebbles and directly form the cores of giant planets. This model is known as 'pebble accretion'; theoretically, it can produce cores of ten Earth masses in only a few thousand years. Unfortunately, full simulations of this process show that, rather than creating a few such cores, it produces a population of hundreds of Earth-mass objects that are inconsistent with the structure of the Solar System. Here we report that this difficulty can be overcome if pebbles form slowly enough to allow the planetesimals to gravitationally interact with one another. In this situation, the largest planetesimals have time to scatter their smaller siblings out of the disk of pebbles, thereby stifling their growth. Our models show that, for a large and physically reasonable region of parameter space, this typically leads to the formation of one to four gas giants between 5 and 15 astronomical units from the Sun, in agreement with the observed structure of the Solar System. PMID:26289203

  5. The growth of planets by pebble accretion in evolving protoplanetary discs

    Science.gov (United States)

    Bitsch, Bertram; Lambrechts, Michiel; Johansen, Anders

    2015-10-01

    The formation of planets depends on the underlying protoplanetary disc structure, which in turn influences both the accretion and migration rates of embedded planets. The disc itself evolves on time scales of several Myr, during which both temperature and density profiles change as matter accretes onto the central star. Here we used a detailed model of an evolving disc to determine the growth of planets by pebble accretion and their migration through the disc. Cores that reach their pebble isolation mass accrete gas to finally form giant planets with extensive gas envelopes, while planets that do not reach pebble isolation mass are stranded as ice giants and ice planets containing only minor amounts of gas in their envelopes. Unlike earlier population synthesis models, our model works without any artificial reductions in migration speed and for protoplanetary discs with gas and dust column densities similar to those inferred from observations. We find that in our nominal disc model, the emergence of planetary embryos preferably should occur after approximately 2 Myr in order to not exclusively form gas giants, but also ice giants and smaller planets. The high pebble accretion rates ensure that critical core masses for gas accretion can be reached at all orbital distances. Gas giant planets nevertheless experience significant reduction in semi-major axes by migration. Considering instead planetesimal accretion for planetary growth, we show that formation time scales are too long to compete with the migration time scales and the dissipation time of the protoplanetary disc. All in all, we find that pebble accretion overcomes many of the challenges in the formation of ice and gas giants in evolving protoplanetary discs. Appendices are available in electronic form at http://www.aanda.org

  6. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

    Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

  7. Quasi-direct numerical simulation of a pebble bed configuration, Part-II: Temperature field analysis

    International Nuclear Information System (INIS)

    Highlights: ► Quasi direct numerical simulations (q-DNSs) of a pebble bed configuration have been performed. ► This q-DNS database may serve as a reference for the validation of different turbulence modeling approaches. ► A wide range of qualitative and quantitative data throughout the computational domain has been generated. ► Results for mean, RMS of temperature and respective turbulent heat fluxes are extensively reported in this paper. -- Abstract: Good prediction of the flow and heat transfer phenomena in the pebble bed core of a high temperature reactor (HTR) is a challenge for available turbulence models, which still require to be validated. While experimental data are generally desirable in this validation process, due to the complex geometric configuration and measurement difficulties, a very limited amount of data is currently available. On the other hand, direct numerical simulation (DNS) is considered an accurate simulation technique, which may serve as an alternative for validating turbulence models. In the framework of the present study, quasi-direct numerical simulation (q-DNS) of a single face cubic centered pebble bed is performed, which will serve as a reference for the validation of different turbulence modeling approaches in order to perform calculations for a randomly arranged pebble bed. These simulations were performed at a Reynolds number of 3088, based on pebble diameter, with a porosity level of 0.42. Results related to flow field (mean, RMS and covariance of velocity) have been presented in Part-I, whereas, in the present article, we focus our attention to the analysis of the temperature field. A wide range of qualitative and quantitative data for the thermal field (mean, RMS and turbulent heat flux) has been generated

  8. Proliferation Resistant Fuel for Pebble Bed Modular Reactors

    International Nuclear Information System (INIS)

    Proliferation of nuclear weapons produced with power reactors plutonium has always been amajor problem of the nuclear energy industry. This includes the PebbleBed Modular Reactor(PBMR), which is a specific design of a GenIV High-Temperature Reactor (HTR), mainly due to its online refueling feature, which may be misused for the production of weapons gradeplutonium. A promising approach toward preventing the proliferation of power reactorplutoniumis to denaturate the plutonium by increasing the ratio of 238Pu to total Pu in the spentfuel(1). The 238Pu isotope is characterized by a high heat rate (approximately 567 W/kg) due to thealpha decay of the 238Pu with half-life of 87.74 yr, in addition to its high spontaneous fissionneutron emission, which is higher than that of 240Pu. Thus, the presence of 238Pu in Pu considerably complicates the design and construction of nuclear weapons based on Pu, owing tothese characteristics of 238Pu. Recent papers(2,3) show that a Pu mixture is proliferation resistant given that the weight ratio of 238Pu to Pu is larger than 6%. In this paper we have studied afeasible technique for ensuring that the 238Pu to Pu ratio, in the Pu produced in PBMR, is larger than 6% during the entire fuel cycle. Contamination of the spent fuel with 238Pu may be achieved by doping the nuclear fuel witheither 241Am or 237Np(4-13). The 238Pu isotope is obtained from both 241Am and 237Np by a neutron-capture reaction and the subsequent decay of the reaction products(13).The 237Np isotopeis by itself a potential weapons grade material. However, its large critical mass of 57±4 kg(14) andthe difficulty of extracting it from irradiated fuel elements make it impractical for weapons purposes. On the other hand, the critical mass of 241Am is smaller, i.e. 34 to 45 kg. However, withdecay heat production of 114W/kg, the critical mass becomes a heat source of 3.9 to 5.1 KW,which makes 241Am unsuitable for weapons applications(3). As a result, it is a non

  9. 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)

  10. Mechanical properties of LiAlO/sub 2/ ceramic breeder pellets

    International Nuclear Information System (INIS)

    The mechanical properties (compressive and bending strengths, sound velocity) have been determined for gamma-LiAlO/sub 2/ pellets produced by different processes at various porosities. A correlation between nondestructive measurements and mechanical strength has been obtained

  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. Tritium permeation through helium-heated steam generators of ceramic breeder blankets for DEMO

    International Nuclear Information System (INIS)

    The specifications of permeation barriers, tritium recovery process maintaining a very low tritium activity in the coolant, and control of the coolant chemistry, required the evaluation of the tritium losses through the steam generators and include the definition of its operating conditions by thermodynamic cycle calculations and its thermal-hydraulic design. For both tasks specific computer tools were developed. The obtained geometry, surface area, and temperature profiles along the heat exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam oxidized Incoloy 800 austenitic stainless steel was identified as the best suited existing material; in nominal steady-state operation, the tritium escape into the steam cycle could be restricted to less than 10 Ci/d. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) is identified to bear a large tritium release potential. Solutions are proposed. (from authors). 4 figs., 1 tab

  14. Recent progress in safety assessments of Japanese water cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Water Cooled Solid Breeder Test Blanket Module (WCSB TBM) is being designed by JAEA for the primary candidate TBM of Japan, and the safety evaluation of WCSB TBM has been performed. This reports presents summary of safety evaluation activities of the Japanese WCSB TBM, including nuclear analysis, source of RI, waste evaluation, occupational radiolysis exposure (ORE), failure mode effect analysis (FMEA) and postulated initiating event (PIE). For the purpose of basic evaluation of source terms on nuclear heating and radioactivity generation, two-dimensional nuclear analysis has been carried out. By the nuclear analysis, distributions of neutron flux, tritium breeding ratio (TBR), nuclear heat, decay heat and induced activity are calculated. Tritium production is calculated by the nuclear analysis by integrating distributions of TBR values, as about 0.2 g-T/FPD. With respect to the radioactive waste, the induced activity of the irradiated TBM is estimated. For the purpose of occupational radiolysis exposure (ORE), RI inventory is estimated. Tritium inventory in pebble bed of TBM is about 3 x 1012 Bq, and tritium in purge gas is about 3 x 1011 Bq. FMEA has been carried out to identify the PIEs that need safety evaluation. PIEs are summarized into three groups, i.e., heating, pressurization and release of RI. PIEs of local heating are converged without any special cares. With respect to heating of whole module, two PIEs are selected as the most severe events, i.e., loss of cooling of TBM during plasma operation and ingress of coolant into TBM during plasma operation. With respect to PIEs about pressurization, the PIEs of pressurization of the compartment nearby the pipes of cooling system are evaluated, because rupture of the pipes result pressurization of such compartments, i.e., box structure of TBM, purge gas loop, TRS, VV, port cell and TCWS vault. Box structure of TBM is designed to withstand the maximum pressure of the cooling system. At other compartments

  15. Kinetics of tritium release from irradiated Li2TiO3 pebbles in out-of-pile TPD tests

    International Nuclear Information System (INIS)

    The rate of tritium release from Li2TiO3 pebbles was examined by post irradiation thermal desorption spectroscopy (the Temperature Programmed Desorption (TPD) method). Pre-treatments before and even after irradiation were found useful to gain insight on the behavior of these pebbles at different temperatures, as good spectrum de-convolution is achieved and kinetic parameters for the rate determining pseudo-first-order steps can be estimated. We show the results concerning Li2TiO3 pebbles bed specimens developed in the frame of the European fusion technology program

  16. 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.)

  17. 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

  18. 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,...

  19. Sensitivity and Uncertainty Analyses of the Tritium Production in the HCPB Breeder Blanket Mock-up Experiment

    International Nuclear Information System (INIS)

    Dedicated computational methods, tools and data have been recently developed in the framework of the European Fusion Technology Programme to enable sensitivity and uncertainty analyses of fusion neutronics experiments. severely limited due to these two requirements. (author)er productgeneration and the associated uncertainties against the experimental data provided in the neutronics experiment at the Frascati Neutron Generator on a mock-up of the HCPB (Helium-Cooled Pebble Bed) breeder test blanket. This work is devoted to the computational analyses of this experiment comprising the following steps: (i) Calculation of the Tritium production rates (TPR) in the Li2CO3 pellets using a detailed 3D model of the experimental set-up; the Monte Carlo code MCNP and the discrete ordinates code TORT were applied for these calculations with EFF-3 and FENDL-2.0/2.1 nuclear data. (ii) Sensitivity calculations for the Li2CO3 pellets stacks to assess the sensitivity of the Tritium production to the reactions cross-sections of the involved nuclides Be, 6,7Li, C and O; the calculations were performed with the MCSEN Monte Carlo code using the track length estimator and, in parallel, with the deterministic SUSD3D code using neutron fluxes calculated by TORT in forward and adjoint mode. (iii) Calculations of the data related uncertainties of the TPR using co-variance data from EFF (9Be, 6Li, 12C), FENDL-2 (7Li) and JENDL-3.3 (16O); both probabilistic (MCNP/MCSEN) and deterministic (TORT/SUSD3D) approaches were applied. (iv) Assessment of the total uncertainties for the TPR including uncertainties of the measurements, the nuclear data and the calculations. The data related uncertainties of the calculated Tritium generation are in the order of 4 - 5 % (2 sigma). The main uncertainties are due to the Be cross-section data. The total uncertainties of the predicted TPR including data uncertainties, statistical uncertainties of the Monte Carlo calculation and the experimental uncertainties

  20. Proposal for an international experimental pebble bed reactor - HTR2008-58174

    International Nuclear Information System (INIS)

    HTRs, both prismatic block fuelled and pebble fuelled, feature a number of uniquely beneficial characteristics that will be discussed in this paper. In this paper the construction of an international experimental pebble bed reactor is proposed, possible experiments suggested and an invitation extended to interested partners for co-operation in the project. Experimental verification by nuclear regulators in order to facilitate licensing and the development of a new generation of reactors create a strong need for such a reactor. Suggested experiments include: Optimized incineration of waste Pu in a pebble bed reactor: The capability to incineration pure reactor grade plutonium by means of ultra high burn-up in pebble bed reactors will be presented at this conference in the track on fuel and fuel cycles. This will enable incineration of the global stockpile of separated reactor grade Pu within a relatively short time span. Testing of fuel sphere geometries, aimed at improving neutron moderation and a decrease in fuel temperatures. Th/Pu fuel cycles: Previous HTR programs demonstrated the viability of a Th-232 fuel-cycle, using highly enriched uranium (HEU) as driver material. However, considerations favoring proliferation resistance limit the enrichment level of uranium in commercial reactors to 20 %, thereby lowering the isotopic efficiency. Therefore, Pu driver material should be developed to replace the HEU component. Instead of deploying a (Th, Pu)O fuel concept, the proposal is to use the unique capability offered by pebble bed reactors in deploying separate Th- and Pu-containing pebbles, which can be cycled differently. Testing of carbon-fiber-carbon (CFC) structures for in-core or near-core applications, such as guide tubes for reserve shutdown systems, thus creating the possibility to safely shutdown reactors with increased diameter. Development of very high temperature reactor components for process heat applications. Advanced decay heat removal systems e

  1. Experimental study on single-phase convection heat transfer characteristics of pebble bed channels with internal heat generation

    International Nuclear Information System (INIS)

    The water-cooled pebble bed reactor core is the porous channels stacked with spherical fuel elements, having evident effect on enhancing heat transfer. Owing to the variability and randomness characteristics of it's interstice, pebble bed channels have a very complex heat transfer situation and have little correlative research. In order to research the heat transfer characters of pebble bed channels with internal heat source, electromagnetic induction heating method was adopted for overall heating the pebble bed which was composed of 8 mm diameter steel balls, and the internal heat transfer characteristics were researched. By comparing and analyzing the experimental data, the rule of power distribution and heat transfer coefficient with heat flux density, inlet temperature and working fluid's Re were got. According to the experimental data fitting, the dimensionless average heat transfer coefficient correlation criteria was got. The fitting results are good agreement with the experimental results within 12% difference. (authors)

  2. Characterization of constrained beryllium pebble beds after neutron irradiation at HFR at high temperatures up to helium production of 3000 appm

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V., E-mail: vladimir.chakin@kit.edu [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Plarz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R. [Institute for Applied Materials – Materials and Biomechanics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Plarz 1, 76344 Eggenstein-Leopoldshafen (Germany); Moeslang, A.; Vladimirov, P.; Kurinskiy, P. [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Plarz 1, 76344 Eggenstein-Leopoldshafen (Germany); Til, S. van; Magielsen, A.J. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/ Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2013-10-15

    Highlights: • Defragmentation of beryllium pebbles at irradiation temperatures of 873 and 948 K was detected. • Formation of brittle beryllium oxide layers on neutron irradiated beryllium pebbles was detected. • Strong interaction between beryllium pebbles and platinum foil under neutron irradiation was detected. • Strong interaction between beryllium pebbles and austenitic stainless steel under neutron irradiation was detected. -- Abstract: Small constrained beryllium pebble beds as well as unconstrained beryllium pebbles have been irradiated within HIDOBE-01 experiment at HFR, Petten, the Netherlands. Beryllium pebbles with 1 mm diameter produced by Rotating Electrode Method (REM) were investigated after irradiation at 630, 740, 873, and 948 K up to helium production of 3000 appm. Intensive pore and bubble formation occurs in beryllium after 873 K irradiation. In the contact zones of the pebbles enhanced pore formation takes place. Oxidation of beryllium pebble external surfaces is accompanied by partial destruction of oxide layers owing to their high brittleness. Strong interactions between beryllium pebbles and platinum foil, as well as between beryllium and stainless steel at contact zones occur at 873 and 948 K.

  3. 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.

  4. Automated control of a pebble bed core thermal flow test unit / by Jan H.J. Prinsloo

    OpenAIRE

    Prinsloo, Jan Hendrik Jacobus

    2006-01-01

    The HTTF (Heat Transfer Test Facility) is a unique project verifying the only pebble bed correlations currently used by PBMR (Pty) LTD. They are developing a new concept nuclear power station and are at present in the preparation phase of the conshuction of the worlds first PBMR (Pebble Bed Modular Reactor). The PBMR required the HTTF to be built at the North-West University in Potchefstroom. The HTTF consists of two separate test facilities: the H7TU (High Temperature Test Uni...

  5. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

    2002-11-01

    This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

  6. The preliminary analysis on the steady-state and kinetic features of the molten salt pebble-bed reactor

    International Nuclear Information System (INIS)

    A novel design concept of molten salt pebble-bed reactor with an ultra-simplified integral primary circuit called 'Nuclear Hot Spring' has been proposed, featured by horizontal coolant flow in a deep pool pebble-bed reactor, providing 'natural safety' features with natural circulation under full power operation and less expensive primary circuit arrangement. In this work, the steady-state physical properties of the equilibrium state of the molten salt pebble-bed reactor are calculated by using the VSOP code, and the steady-state thermo-hydraulic analysis is carried out based on the approximation of absolutely horizontal flow of the coolant through the core. A new concept of 2-dimensional, both axial and radial, multi-pass on-line fuelling scheme is presented. The result reveals that the radial multi-pass scheme provides more flattened power distribution and safer temperature distribution than the one-pass scheme. A parametric analysis is made corresponding to different pebble diameters, the key parameter of the core resistance and the temperature at the pebble center. It is verified that within a wide range of pebble diameters, the maximum pebble center temperatures are far below the safety limit of the fuel, and the core resistance is considerably less than the buoyant force, indicating that the natural circulation under full power operation is achievable and the ultra-simplified integral primary circuit without any pump is possible. For the kinetic properties, it is verified that the negative temperature coefficient is achieved in sufficient under-moderated condition through the preliminary analysis on the temperature coefficients of fuel, coolant and moderator. The requirement of reactivity compensation at the shutdown stages of the operation period is calculated for the further studies on the reactivity control. The molten salt pebble-bed reactor with horizontal coolant flow can provide enhanced safety and economical features. (authors)

  7. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  8. 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)

  9. Filbe molten salt research for tritium breeder applications

    International Nuclear Information System (INIS)

    This paper presents an overview of Flibe (2Lif·BeF2) molten salt research activities conducted at the INEEL as part of the Japan-US JUPITER-II joint research program. The research focuses on tritium/chemistry issues for self-cooled Flibe tritium breeder applications and includes the following activities: (1) Flibe preparation, purification, characterization and handling, (2) development and testing of REDOX strategies for containment material corrosion control, (3) tritium behavior and management in Flibe breeder systems, and (4) safety testing (e.g., mobilization of Flibe during accident scenarios). This paper describes the laboratory systems developed to support these research activities and summarizes key results of this work to date. (author)

  10. 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

  11. 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

  12. Feeding broiler breeders to improve their welfare whilst maintaining productivity

    DEFF Research Database (Denmark)

    Steenfeldt, Sanna; Nielsen, Birte Lindstrøm

    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......In the present experiment different types of fibre sources were used in high fibre diets to increase feeding quantity whilst limiting the growth of broiler breeders to industry recommended levels. Using scatter feeding, three diets (CON, commercial control diet; INF, high insoluble fibre content......; 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...

  13. 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...

  14. 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 ...

  15. 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

  16. 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

  17. 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

  18. The radial dependence of pebble accretion rates: a source of diversity in planetary systems I. Analytical formulation

    CERN Document Server

    Ida, Shigeru; Morbidelli, Alessandro

    2016-01-01

    Context. The classical "planetesimal" accretion scenario for the formation of planets has recently evolved with the idea that "pebbles", centimeter- to meter-sized icy grains migrating in protoplanetary disks, can control planetesimal and/or planetary growth. Aims. We investigate how pebble accretion depends on disk properties and affects the formation of planetary systems Methods. We construct analytical models of pebble accretion onto planetary embryos that consistently account for the mass and orbital evolution of the pebble flow and reflect disk structure. Results. We derive simple formulas for pebble accretion rates in the so-called "settling" regime for planetary embryos with more than 100 km in size. For relatively smaller embryos or in outer disk regions, the accretion mode is 3D, meaning that the thickness of the pebble flow must be taken into account, and resulting in an accretion rate that is independent of the embryo mass. For larger embryos or in inner regions, the accretion is in a 2D mode, i.e....

  19. Thermo-mechanical and neutron lifetime modeling and design of Be pebbles in the neutron multiplier for the LIFE engine

    Energy Technology Data Exchange (ETDEWEB)

    DeMange, P; Marian, J; de Caro, M S; Caro, A

    2009-03-16

    Concept designs for the laser-initiated fusion/fission engine (LIFE) include a neutron multiplication blanket containing Be pebbles flowing in a molten salt coolant. These pebbles must be designed to withstand the extreme irradiation and temperature conditions in the blanket to enable a safe and cost-effective operation of LIFE. In this work, we develop design criteria for spherical Be pebbles on the basis of their thermomechanical behavior under continued neutron exposure. We consider the effects of high fluence/fast flux on the elastic, thermal and mechanical properties of nuclear-grade Be. Our results suggest a maximum pebble diameter of 30 mm to avoid tensile failure, coated with an anti-corrosive, high-strength metallic shell to avoid failure by pebble contact. Moreover, we find that the operation temperature must always be kept above 450 C to enable creep to relax the stresses induced by swelling, which we estimate to be at least 16 months if uncoated and up to six years when coated. We identify the sources of uncertainty on the properties used and discuss the advantages of new intermetallic beryllides and their use in LIFE's neutron multiplier. To establish Be-pebble lifetimes with improved confidence, reliable experiments to measure irradiation creep must be performed.

  20. Thermo-mechanical and neutron lifetime modeling and design of Be pebbles in the neutron multiplier for the LIFE engine

    International Nuclear Information System (INIS)

    Concept designs for the laser-initiated fusion/fission engine (LIFE) include a neutron multiplication blanket containing Be pebbles flowing in a molten salt coolant. These pebbles must be designed to withstand the extreme irradiation and temperature conditions in the blanket to enable a safe and cost-effective operation of LIFE. In this work, we develop design criteria for spherical Be pebbles on the basis of their thermomechanical behavior under continued neutron exposure. We consider the effects of high fluence/fast flux on the elastic, thermal and mechanical properties of nuclear-grade Be. Our results suggest a maximum pebble diameter of 30 mm to avoid tensile failure, coated with an anti-corrosive, high-strength metallic shell to avoid failure by pebble contact. Moreover, we find that the operation temperature must always be kept above 450 C to enable creep to relax the stresses induced by swelling, which we estimate to be at least 16 months if uncoated and up to six years when coated. We identify the sources of uncertainty on the properties used and discuss the advantages of new intermetallic beryllides and their use in LIFE's neutron multiplier. To establish Be-pebble lifetimes with improved confidence, reliable experiments to measure irradiation creep must be performed

  1. Liquid metal fast breeder reactor: an environmental and economic critique

    International Nuclear Information System (INIS)

    Economic and environmental arguments made by the AEC and others for the liquid metal fast breeder reactor (LMFBR) as a central component of the U. S. electrical energy system are discussed. The LMFBR appears to have no environmental advantage over the currently operating light water reactor and especially not over the high temperature gas reactor. The principle environmental argument for the rapid introduction of LMFBRs is that they will provide a virtually inexhaustible fuel source, and reduce the demand for strip-mining the limited reserves of high grade U ore. A 20-yr delay in the construction of LMFBRs would result in an increase of only 50 mi2 of strip mining over the next 50 yr, and the cost of reclamation of this land would be about 0.1 mill/kw-hr. Uranium from which fuel has been extracted for use by nonbreeder reactors can still be used by breeders, thus breeders could still be introduced in the future, if fusion is not developed in time, and extract the same overall energy from a given supply of U as if they had been introduced earlier. Economic arguments in favor of the LMFBR are based on models highly sensitive to changes on some of the most critical input variables: nuclear power plant capital costs, fuel cycle costs, performance characteristics of LMFBR designs, electrical energy demand, and U ore costs. There is no basis for concluding that the LMFBR will be economical in the 1980s or early 1990s. (Pollut. Abstr.)

  2. A fast breeder reactor development scheme for Brazil

    International Nuclear Information System (INIS)

    Fast breeder reactors will be necessary in the next century in order to meet increasing demands for electricity resulting from industrialization and general improvement of standards of living. A scheme for the development of liquid metal fast breeder reactors in Brazil is proposed. Emphasis are placed on reactor safety in order to promote public acceptance, on utilization of thorium that is abundant in the country, and on consistency and smoothness of the development. The initial step is the construction and operation of a 5 MW experimental fast reactor in order to acquire basic experiences and technologies. The second step is the construction of a series of small power plants which should assure a ssound technological development. The reactor is designed with particular emphasis on safety and ease of operation. Demonstration of safety and reliability with small units would enhance public acceptance. In the final phase, when fast breeder reactors are to play a central role in electricity generation, large power plants that utilize both uranium and thorium fuel cycles will be built to establish a practically permanent power system. (Author)

  3. Gas Reactor International Cooperative Program: German Pebble Bed Reactor Technology review update

    International Nuclear Information System (INIS)

    This report provides a review of the German pebble bed reactor technology, and updates the information provided in the Gas Reactor International Cooperative Program Interim Report COO-4057-6, German Pebble Bed Reactor Design and Technology Review, dated September 1978. Most of the updated information is for the PNP-500 and the HHT-Prototype plants. The PNP-500 is a 500 MW(t) multi-purpose demonstration plant for coal conversion applications. The HHT-Prototype is a 1640 MWt reactor designed to produce 675 MWe of electricity using a direct cycle gas turbine. The report provides a description and evaluation of the overall plant and the nuclear reactor for both the PNP-500 and HHT-Prototype. A description and evaluation of the primary system components is presented for the process heat and gas turbine applications

  4. Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

    Directory of Open Access Journals (Sweden)

    SA Hosseini

    2013-09-01

    Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

  5. Transuranics elimination in an optimised pebble-bed sub-critical reactor

    International Nuclear Information System (INIS)

    In a nuclear energy economy the nuclear waste is a big burden to its further development and deployment. The possibility of eliminating the long-term part of the waste presents an appealing opportunity to the sustainability and acceptance of a better and cleaner source of energy. It is shown that the proposed pebble-bed transmutator has suitable characteristics to transmute most of the isotopes that contribute to the long-term radioactivity. This proposed reactor presents also inherent safety characteristics, which is a necessary element in a new reactor design to be accepted by the society. Throughout this paper, we will characterise the new reactor concept, and present some of the neutronics and safety characteristics of an accelerator driven pebble-bed reactor, (ADS) for transuranics elimination. (author)

  6. Measurement of the thermal conductivity and heat transfer coefficient of a binary bed of beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Donne, M.D.; Piazza, G. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Goraieb, A.; Sordon, G.

    1998-01-01

    The four ITER partners propose to use binary beryllium pebble bed as neutron multiplier. Recently this solution has been adopted for the ITER blanket as well. In order to study the heat transfer in the blanket the effective thermal conductivity and the wall heat transfer coefficient of the bed have to be known. Therefore at Forschungszentrum Karlsruhe heat transfer experiments have been performed with a binary bed of beryllium pebbles and the results have been correlated expressing thermal conductivity and wall heat transfer coefficients as a function of temperature in the bed and of the difference between the thermal expansion of the bed and of that of the confinement walls. The comparison of the obtained correlations with the data available from the literature show a quite good agreement. (author)

  7. Gas reactor international cooperative program interim report. Pebble bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, compare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  8. 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

  9. 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.)

  10. Medium voltage direct current (MVDC) converter for pebble bed modular reactor (PBMR) / Hendrik de Villiers Pretorius

    OpenAIRE

    Pretorius, Hendrik de Villiers

    2004-01-01

    Nuclear and renewable energy systems will probably be used more and more extensively in future due to high environmental demands regarding pollution and exhaustion of the world's gas and coal reserves. Because most types of renewable energy systems do not supply electric power at line frequency and voltage a converter is used to connect these sources to the existing power system. The Pebble Bed Modular Reactor (PBMR) is a nuclear power plant currently using a 50 Hz synchrono...

  11. Design of a power conversion system for an indirect cycle, helium cooled pebble bed reactor system

    International Nuclear Information System (INIS)

    A design is presented for the turbomachinery for an indirect cycle, closed, helium cooled modular pebble bed reactor system. The design makes use of current technology and will operate with an overall efficiency of 45%. The design uses an intermediate heat exchanger which isolated the reactor cycle from the turbomachinery. This design excludes radioactive fission products from the turbomachinery. This minimizes the probability of an air ingress accident and greatly simplifies maintenance. (author)

  12. Influence of surface condition on deuterium release from Li2TiO3 pebble

    International Nuclear Information System (INIS)

    Highlights: • Li2TiO3 pebbles were irradiated with deuterium ions and heated repeatedly for investigating the change in deuterium desorption behavior from the pebbles. • The composition of Li decreased and those of Ti and O increased with the increase of the number of the irradiation/heating cycle. • The desorption peaks of the gases contained deuterium atoms were shifted to higher temperature region because of the defects created by the irradiation. • The amount of deuterium desorbed in forms of water increase because of the increase of the composition ratio of O near the surface. - Abstract: Lithium titanate (Li2TiO3) pebbles were irradiated with D3+ ions with energy of 5.0 keV, and the amounts of retained deuterium in the pebbles were measured by thermal desorption spectroscopy. In this research the irradiation/heating cycles were carried out repeatedly in order to investigate the influence of surface condition on deuterium release from Li2TiO3. The composition ratio of Li decreased with the increase of the number of the irradiation/heating cycle. Then, the desorption peaks of the gases contained deuterium atoms were shifted to higher temperature region, and the amount of desorbed gases in forms of water tended to increase. In addition, we carried out other experiments for the comparison. Comparing these results, we considered that the increase of the defects created by the irradiation was more responsible for the change in the desorption behavior by the irradiation/heating cycles than the lithium depletion. These results suggest that the tritium recovery efficiency would decrease with the increase of the defects and the damages especially at the low temperature region during the operation

  13. Computational and experimental prediction of dust production in pebble bed reactors—Part I

    International Nuclear Information System (INIS)

    Highlights: • A nonlinear dimensionless wear coefficient is theoretically proposed. • A material constant for the relation of asperity height and wear is introduced. • A nonlinear modification of Archard wear formula is proposed. • The graphite wear dust production in a typical pebble bed reactor is predicted. • Experimental and computational wear results for graphite are presented. -- Abstract: This paper describes the computational modeling and simulation, and experimental testing of graphite moderators in frictional contacts as anticipated in a pebble bed reactor. The potential of carbonaceous particulate generation due to frictional contact at the surface of pebbles and the ensuing entrainment and transport into the gas coolant are safety concerns at elevated temperatures under accident scenarios such as air ingress in the high temperature gas-cooled reactor. The safety concerns are due to the documented ability of carbonaceous particulates to adsorb fission products and transport them in the primary circuit of the pebble bed reactor, thus potentially giving rise to a relevant source term under accident scenarios. Here, a finite element approach is implemented to develop a nonlinear wear model in air environment. In this model, material wear coefficient is related to the changes in asperity height during wear. The present work reports a comparison between the finite element simulations and the experimental results obtained using a custom-designed tribometer. The experimental and computational results are used to estimate the quantity of nuclear grade graphite dust produced from a typical anticipated configuration. In Part II, results from a helium environment at higher temperatures and pressures are experimentally studied

  14. Generic Investigations on Transport Theory Modelling of High Temperature Reactors of Pebble Bed Type

    OpenAIRE

    Sureda Sureda, Antonio Jaime

    2008-01-01

    The GRS (Gesellschaft fuer Anlagen und Reaktorsicherheit = Company for Plant and Reactor Safety) maintains and further develops the code system DORT-TD/HERMIX-DIREKT, which is a complex tool for the simulation of coupled neutronics/thermal-hydraulics transients and accident scenarios of high-temperature gas cooled reactors of pebble bed type. With this tool, GRS takes part in the international benchmark activity "OECD/NEA PBMR400 Transient Benchmark”, which aims at the simulation of transient...

  15. The study on fuel effect in discharge pipe of pebble bed reactor

    International Nuclear Information System (INIS)

    The simulation method of fuel loading in discharge pipe of pebble bed reactor is introduced. As an exemplary application case, the effect of fuel elements in the discharge pipe on reactor physics and thermohydraulic properties is calculated and analysed by CHTRP code in HTR-10 MW. The calculation gives the power and temperature distribution in the area of the discharge pipe, very useful for further analysis of reactor physics and safety

  16. Burnup performance of OTTO cycle pebble bed reactors with ROX fuel

    International Nuclear Information System (INIS)

    Highlights: • A 300 MWt Small Pebble Bed Reactor with Rock-like oxide fuel is proposed. • Using ROX fuel can achieve high discharged burnup of spent fuel. • High geological stability can be expected in direct disposal of the spent ROX fuel. • The Pebble Bed Reactor with ROX fuel can be critical at steady state operation. • All the reactor designs have a negative temperature coefficient. - Abstract: A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor

  17. Ceramic nuclear fuels development at BARC - recent activities

    International Nuclear Information System (INIS)

    The ceramic nuclear fuels activities at Bhabha Atomic Research Centre (BARC) in recent years have been in the following areas: - improvisation of process flowsheet for economic and efficient production of high density UO2 fuel pellets on industrial scale for pressurised heavy water reactors (PHWR), - fabrication, thermophysical property evaluation and irradiation-testing of (U, Pu)O2, (Th, Pu)O2 and (Th, U)O2 fuels for PHWRs, - fabrication and thermophysical property evaluation of (U, Pu)O2, (U, Pu)C and (U, Pu)N fueld pellets for liquid metal cooled fast breeder reactors (LMFBR). The conventional ''powder-pellet'' and the advanced ''sol-gel microsphere pelletisation'' (SGMP) processes have been optimised for fabrication of oxide, carbide and nitride fuel pellets of controlled density and microstructure. The SGMP route is dust-free and is ideally suitable for remote fabrication of highly radiotoxic Pu and Pu233 bearing ceramic fuels. For fabrication of UO2 and (U, Pu)O2 pellets, a process combining SGMP and low temperature oxidative sintering has been developed for minimising radioactive dust hazard and energy requirement during sintering. The thermal conductivity (laser flash method) and hot hardness (using Vickers diamond indentor) of oxide, carbide and nitride fuel pellets were evaluated upto 1800 K and 1500 K respectively. The out-of-pile experiments on chemical compatibility of mixed uranium plutonium monocarbide and mononitride fuel pellets with stainless steel (type SS 316) cladding have been carried out to find their suitability as fueld for fast breeder test reactor (FBTR) and prototype fast breeder reactors (PFBR) at Indira Gandhi Centre for Atomic Research (IGCAR) at Kalpakkam. (orig.)

  18. Development and testing of analytical models for the pebble bed type HTRs

    International Nuclear Information System (INIS)

    The pebble bed type gas cooled high temperature reactor (HTR) appears to be a good candidate for the next generation nuclear reactor technology. These reactors have unique characteristics in terms of the randomness in geometry, and require special techniques to analyze their systems. This study includes activities concerning the testing of computational tools and the qualification of models. Indeed, it is essential that the validated analytical tools be available to the research community. From this viewpoint codes like MCNP, ORIGEN and RELAP5, which have been used in nuclear industry for many years, are selected to identify and develop new capabilities needed to support HTR analysis. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP. The coupled MCNP-ORIGEN code is used to estimate the burnup and the refuelling scheme. Results obtained from Monte Carlo analysis are interfaced with RELAP5 to analyze the thermal hydraulics and safety characteristics of the reactor. New models and methodologies are developed for several past and present experimental and prototypical facilities that were based on HTR pebble bed concepts. The calculated results are compared with available experimental data and theoretical evaluations showing very good agreement. The ultimate goal of the validation of the computer codes for pebble bed HTR applications is to acquire and reinforce the capability of these general purpose computer codes for performing HTR core design and optimization studies

  19. Coupling of RMC and CFX for analysis of Pebble Bed-Advanced High Temperature Reactor core

    International Nuclear Information System (INIS)

    Highlights: ► The CFD code CFX is used for whole pebble bed reactor core calculation. ► The Monte Carlo Code RMC and CFX are used for the coupling of neutronics and T-H. ► Coupled calculations for steady-state problem can reach stable results. ► Increasing the number of neutron histories is effective to improve accuracy. - Abstract: This paper introduces a steady-state coupled calculation method using the Monte Carlo Code RMC (Reactor Monte Carlo) and the Computational Fluid Dynamic (CFD) code CFX for the analysis of a Pebble Bed-Advanced High Temperature Reactor (PB-AHTR) core. The RMC code is used for neutronics calculation while CFX is used for Thermal-Hydraulics (T-H) calculation. The porous media model is used in CFX modeling to simulate the pebble bed structure in PB-AHTR. The CFX model has also been validated against the RELAP5-3D model developed in the previous research. The script language PERL is used as a development tool to manipulate and control the entire coupled calculation. This research gives the conclusion that the steady-state coupled calculation using RMC and CFX is feasible and can obtain stable results within a few iterations. However, due to the statistical errors of Monte Carlo method, the fluctuation of results still occurs. For the purpose of improving the accuracy, the paper applies and discusses two methods, of which increasing the number of neutron histories is an effective method.

  20. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.