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Sample records for breeder blanket design

  1. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

    Sullivan, J.D.; Palmer, B.J.F.

    1987-11-01

    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

  2. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.A.

    1980-01-01

    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. Solid breeder test blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ying, A. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States)]. E-mail: ying@fusion.ucla.edu; Abdou, M. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Calderoni, P. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Sharafat, S. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Youssef, M. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); An, Z. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Abou-Sena, A. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Kim, E. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Reyes, S. [LANL, Livermore, CA (United States); Willms, S. [LANL, Los Alamos, NM (United States); Kurtz, R. [PNNL, Richland, WA (United States)

    2006-02-15

    This paper presents the design and analysis for the US ITER solid breeder blanket test articles. Objectives of solid breeder blanket testing during the first phase of the ITER operation focus on exploration of fusion break-in phenomena and configuration scoping. Specific emphasis is placed on first wall structural response, evaluation of neutronic parameters, assessment of thermomechanical behavior and characterization of tritium release. The tests will be conducted with three unit cell arrays/sub-modules. The development approach includes: (1) design the unit cell/sub-module for low temperature operations and (2) refer to a reactor blanket design and use engineering scaling to reproduce key parameters under ITER wall loading conditions, so that phenomena under investigation can be measured at a reactor-like level.

  4. Design and development of ceramic breeder demo blanket

    International Nuclear Information System (INIS)

    Enoeda, M.; Sato, S.; Hatano, T.

    2001-01-01

    Ceramic breeder blanket development has been widely conducted in Japan from fundamental researches to project-oriented engineering scaled development. A long term R and D program has been launched in JAERI since 1996 as a course of DEMO blanket development. The objectives of this program are to provide engineering data base and fabrication technologies of the DEMO blanket, aiming at module testing in ITER currently scheduled to start from the beginning of the ITER operation as a near-term target. Two types of DEMO blanket systems, water cooled blanket and helium cooled blanket, have been designed to be consistent with the SSTR (Steady State Tokamak Reactor) which is the reference DEMO reactor design in JAERI. Both of them utilize packed small pebbles of breeder Li 2 O or Li 2 TiO 3 as a candidate) and neutron multiplier (Be) and rely on the development of advanced structural materials (a reduced activation ferritic steel F82H) compatible with high temperature operation. (author)

  5. Ceramic sphere-pac breeder design for fusion blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.J.; Sullivan, J.D.

    1991-01-01

    Randomly packed beds of ceramic spheres are a practical approach to surrounding fusion plasmas with tritium-breeding material. This paper examines the general properties of sphere-pac beds for application in fusion breeder blankets. The design considerations and models are reviewed for packing, tritium breeding and recovery, thermal conductivity, purge-gas pressure drop, mechanical behavior and fabrication. The design correlations are compared against available fusion ceramic data. Specific conclusions are that ternary (three-size) beds are not attractive for fusion blankets, and that the fusion spheres should be as large as possible subject primarily to packing constraints. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  7. Design and R and D activities on ceramic breeder blanket for fusion experimental reactors in JAERI

    International Nuclear Information System (INIS)

    Kurasawa, T.; Takatsu, H.; Sato, S.; Nakahira, M.; Furuya, K.; Hashimoto, T.; Kawamura, H.; Kuroda, T.; Tsunematsu, T.; Seki, M.

    1995-01-01

    Design and R and D activities on ceramic breeder blanket of a fusion experimental reactor have been progressed in JAERI. A layered pebble bed type ceramic breeder blanket with water cooling is a prime candidate concept. Design activities have been concentrated on improvement of the design by conducting detailed analyses and also by fabrication procedure consideration based on the current technologies. A wide variety of R and Ds have also been conducted in accordance with the design activities. Development of fabrication technology of the blanket box structure and its mechanical testing, elementary testing on thermal performances of the pebble bed, and engineering-oriented material tests of breeder and beryllium pebbles are the main achievements during the last two years. (orig.)

  8. Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

    Science.gov (United States)

    Deepak, SHARMA; Paritosh, CHAUDHURI

    2018-04-01

    The Indian test blanket module (TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the R&D activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices (ITER relevant and DEMO). The Indian Lead–Lithium Cooled Ceramic Breeder (LLCB) blanket concept is one of the Indian DEMO relevant TBM, to be tested in ITER as a part of the TBM program. Helium-Cooled Ceramic Breeder (HCCB) is an alternative blanket concept that consists of lithium titanate (Li2TiO3) as ceramic breeder (CB) material in the form of packed pebble beds and beryllium as the neutron multiplier. Specifically, attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions. These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.

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

    Science.gov (United States)

    Enoeda, M.; Kosaku, Y.; Hatano, T.; Kuroda, T.; Miki, N.; Honma, T.; Akiba, M.; Konishi, S.; Nakamura, H.; Kawamura, Y.; Sato, S.; Furuya, K.; Asaoka, Y.; Okano, K.

    2003-12-01

    This paper presents results of conceptual design activities and associated R&D of a solid breeder blanket system for demonstration of power generation fusion reactors (DEMO blanket) cooled by supercritical water. The Fusion Council of Japan developed the long-term research and development programme of the blanket in 1999. To make the fusion DEMO reactor more attractive, a higher thermal efficiency of more than 40% was strongly recommended. To meet this requirement, the design of the DEMO fusion reactor was carried out. In conjunction with the reactor design, a new concept of a solid breeder blanket cooled by supercritical water was proposed and design and technology development of a solid breeder blanket cooled by supercritical water was performed. By thermo-mechanical analyses of the first wall, the tresca stress was evaluated to be 428 MPa, which clears the 3Sm value of F82H. By thermal and nuclear analyses of the breeder layers, it was shown that a net TBR of more than 1.05 can be achieved. By thermal analysis of the supercritical water power plant, it was shown that a thermal efficiency of more than 41% is achievable. The design work included design of the coolant flow pattern for blanket modules, module structure design, thermo-mechanical analysis and neutronics analysis of the blanket module, and analyses of the tritium inventory and permeation. Preliminary integration of the design of a solid breeder blanket cooled by supercritical water was achieved in this study. In parallel with the design activities, engineering R&D was conducted covering all necessary issues, such as development of structural materials, tritium breeding materials, and neutron multiplier materials; neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was

  10. Conceptual design and analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Li, Min; Lv, Zhongliang; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Wang, Xiaoliang; Zheng, Jie; Ye, Minyou

    2015-10-15

    Highlights: • A helium cooled solid blanket was proposed as a candidate blanket concept for CFETR. • Material selection, basic structure and gas flow scheme of the blanket were introduced. • A series of performance analyses for the blanket were summarized. - Abstract: To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and is being designed mainly to demonstrate 50–200 MW fusion power, 30–50% duty time factor, tritium self-sustained. Because of the high demand of tritium production and the realistic engineering consideration, the design of tritium breeding blanket for CFETR is a challenging work and getting special attention. As a blanket candidate, a helium cooled solid breeder blanket has been designed with the emphasis on conservative design and realistic blanket technology. This paper introduces the basic blanket scheme, including the material selection, structural design, cooling scheme and purge gas flow path. In addition, some results of neutronics, thermal-hydraulic and stress analysis are presented.

  11. Design study of blanket structure based on a water-cooled solid breeder for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Youji; Tobita, Kenji; Utoh, Hiroyasu; Tokunaga, Shinji; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Sakamoto, Yoshiteru

    2015-10-15

    Highlights: • Neutronics design of a water-cooled solid mixed breeder blanket was presented. • The blanket concept achieves a self-sufficient supply of tritium by neutronics analysis. • The overall outlet coolant temperature was 321 °C, which is in the acceptable range. - Abstract: Blanket concept with a simplified interior for mass production has been developed using a mixed bed of Li{sub 2}TiO{sub 3} and Be{sub 12}Ti pebbles, coolant conditions of 15.5 MPa and 290–325 °C and cooling pipes without any partitions. Considering the continuity with the ITER test blanket module option of Japan and the engineering feasibility in its fabrication, our design study focused on a water-cooled solid breeding blanket using the mixed pebbles bed. Herein, we propose blanket segmentation corresponding to the shape and dimension of the blanket and routing of the coolant flow. Moreover, we estimate the overall tritium breeding ratio (TBR) with a torus configuration, based on the segmentation using three-dimensional (3D) Monte Carlo N-particle calculations. As a result, the overall TBR is 1.15. Our 3D neutronics analysis for TBR ensures that the blanket concept can achieve a self-sufficient supply of tritium.

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

    International Nuclear Information System (INIS)

    Yuan Tao; Feng Kaiming; Chen Zhi; Wang Xiaoyu

    2007-01-01

    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 (Li 4 SiO 4 ) 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)

  13. Progress in design and analysis of the net water cooled liquid breeder blanket

    International Nuclear Information System (INIS)

    Danner, W.; Rieger, M.; Verschuur, K.A.; Vieider, G.; Casini, G.; Chazalon, M.; Libin, B.; Farfaletti-Casali, F.; Piana, R.

    1987-01-01

    The NET liquid breeder blanket was subjected to a major design revision and integrated in the new NET-DN machine configuration. In this paper briefly the most essential design features are summarized and some results from thermohydraulics and 1D as well as 3D neutronics analyses are presented. It is concluded that the performance meets well the requirements of NET but that the concept needs substantial improvement if applied to a reactor

  14. Conceptual design of two helium cooled fusion blankets (ceramic and liquid breeder) for INTOR

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Taczanowski, S.

    1983-08-01

    Neutronic and heat transfer calculations have been performed for two helium cooled blankets for the INTOR design. The neutronic calculations show that the local tritium breeding ratios, both for the ceramic blanket (Li 2 SiO 3 ) and for the liquid blanket (Li 17 Pb 83 ) solutions, are 1.34 for natural tritium and about 1.45 using 30% Li 6 enrichment. The heat transfer calculations show that it is possible to cool the divertor section of the torus (heat flux = 1.7 MW/m 2 ) with helium with an inlet pressure of 52 bar and an inlet temperature of 40 0 C. The temperature of the back face of the divertor can be kept at 130 0 C. With helium with the same inlet conditions it is possible to cool the first wall as well (heat flux = 0.136 MW/m 2 ) and keep the back-face of this wall at a temperature of 120 0 C. For the ceramic blanket we use helium with 52 bar inlet pressure and 400 0 C inlet temperature to ensure sufficiently high temperatures in the breeder material. The maximum temperature in the pressure tubes containing the blanket is 450 0 C, while the maximum breeder particle temperature is 476 0 C. (orig./RW) [de

  15. US-DOE Fusion-Breeder Program: blanket design and system performance

    International Nuclear Information System (INIS)

    Lee, J.D.

    1983-01-01

    Conceptual design studies are being used to assess the technical and economic feasibility of fusion's potential to produce fissile fuel. A reference design of a fission-suppressed blanket using conventional materials is under development. Theoretically, a fusion breeder that incorporates this fusion-suppressed blanket surrounding a 3000-MW tandem mirror fusion core produces its own tritium plus 5600 kg of 233 U per year. The 233 U could then provide fissile makeup for 21 GWe of light-water reactor (LWR) power using a denatured thorium fuel cycle with full recycle. This is 16 times the net electric power produced by the fusion breeder (1.3 GWe). The cost of electricity from this fusion-fission system is estimated to be only 23% higher than the cost from LWRs that have makeup from U 3 O 8 at present costs (55 $/kg). Nuclear performance, magnetohydrodynamics (MHD), radiation effects, and other issues concerning the fission-suppressed blanket are summarized, as are some of the present and future objectives of the fusion breeder program

  16. 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

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

    International Nuclear Information System (INIS)

    Enoeda, Mikio; Akiba, Masato; Ohara, Yoshihiro

    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 2 TiO 3 or Li 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 the energy

  18. Pre-conceptual design study on K-DEMO ceramic breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Sung, E-mail: jspark@nfri.re.kr [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kwon, Sungjin; Im, Kihak; Kim, Keeman [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Brown, Thomas; Neilson, George [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2015-11-15

    A pre-conceptual design study has been carried out for the Korean fusion demonstration reactor (K-DEMO) tokamak featured by high magnetic field (B{sub T0} = 7.4 T), R = 6.8 m, a = 2.1 m, and a steady-state operation. The design concepts of the K-DEMO blanket system considering the cooling in-vessel components with pressurized water and a solid pebble breeder are described herein. The structure of the K-DEMO blanket is toroidally subdivided into 16 inboard and 32 outboard sectors, in order to allow the vertical maintenance. Each blanket module is composed of plasma-facing first wall, layers of breeding parts, shielding and manifolds. A ceramic breeder using Li{sub 4}SiO{sub 4} pebbles with Be{sub 12}Ti as neuron multiplier is employed for study. MCNP neutronic simulations and thermo-hydraulic analyses are interactively performed in order to satisfy two key aspects: achieving a global Tritium Breeding Ratio (TBR) >1.05 and operating within the maximum allowable temperature ranges of materials.

  19. Design of the breeder units in the new HCPB modular blanket concept and material requirements

    International Nuclear Information System (INIS)

    Boccaccini, L.V.; Fischer, U.; Hermsmeyer, S.; Reimann, J.; Xu, Z.; Koehly, C.

    2004-01-01

    A major revision of the DEMO HCPB blanket concept took place in 2002-2003 as consequence of the results of the EU Power Plant Conceptual Study. In particular, it was decided to give up the previous maintenance schema based on segments in favour of a large module concept extrapolated from ITER. The adaptation of the HCPB concept to these modules (typical dimension at the FW of 2.0 x 2.0 m) required a complete revision of the box. The coolant flow scheme is based on a radial He flow (at 8 MPa) in order to have the entire manifold system in the rear part of the box. Furthermore, the requirement of a box capable of withstanding the coolant pressure of 8 MPa in case of an in-box LOCA led to a design of modules with an internal stiffening grid in toroidal and poloidal direction This grid results in cells open in the rear radial direction with toroidal-poloidal dimensions of about 20 cm x 20 cm that accommodate the breeder units. These units contain the ceramic breeder (CB) and the Beryllium in form of pebble beds and have to assure the main functions of the blanket, namely, a tritium breeding ratio significantly above one, heat removal with a temperature control in the beds and in the structure, mechanical stability of the beds and extraction of the produced tritium. Due to the relatively high quantity of steel necessary to assure the mechanical stability of the box, a strong requirement for the design of these units is to minimise the amount of steel to improve the neutronic performance. A satisfactory design has been achieved with a radial-toroidal bed configuration similar to the old DEMO design reaching the Tritium self-sufficiency with a radial depth of 47 cm, using monosized Beryllium and CB beds and, using Li 4 SiO 4 , a 6 Li enrichment of about 40%. This design allows a satisfactory control of the maximum acceptable temperatures in the CB and Be beds and the steel structure. The design of the breeder units has not been yet analysed thermo-mechanically in detail

  20. Status of EC solid breeder blanket designs and R and D for demo fusion reactors

    International Nuclear Information System (INIS)

    Proust, E.; Anzidei, L.; Moons, F.

    1994-01-01

    Within the European Community Fusion Technology Program two solid breeder blankets for a DEMO reactor are being developed. The two blankets have various features in common: helium as coolant and as tritium purge gas, the martensitic steel MANET as structural material and beryllium as neutron multiplier. The configurations of the two blankets are however different: in the B.I.T. (Breeder Inside Tube) concept the breeder materials are LiAlO 2 or Li 2 ZrO 3 in the form of annular pellets contained in tubes surrounded by beryllium blocks, the coolant helium being outside the tubes, whereas in the B.O.T. (Breeder out of Tube) the breeder and multiplier material are Li 4 SiO 4 and beryllium pebbles forming a mixed bed placed outside the tubes containing the coolant helium. The main critical issues for both blankets are the behavior of the breeder ceramics and of beryllium under irradiation and the tritium control. Other issues are the low temperature irradiation induced embrittlement of MANET, the mechanical effects caused by major plasma disruptions, and safety and reliability. The R and D work concentrate on these issues. The development of martensitic steels including MANET is part of a separate program. Breeder ceramics and beryllium irradiations have been so far performed for conditions which do not cover the peak values injected in the DEMO blankets. Further irradiations in thermal reactors and in fast reactors, especially for beryllium, are required. An effective tritium control requires the development of permeation barriers and/or of methods of oxidation of the tritium in the main helium cooling systems. First promising results have been obtained also in field of mechanical effects from plasma disruptions and safety and reliability, however further work is required in the reliability field and to validate the codes for the calculations of the plasma disruption effects. (authors). 8 figs., 2 tabs., 53 refs

  1. Preconceptual design and analysis of a solid-breeder blanket test in an existing fission reactor

    International Nuclear Information System (INIS)

    Deis, G.A.; Hsu, P.Y.; Watts, K.D.

    1983-01-01

    Preconceptual design and analysis have been performed to examine the capabilities of a proposed fission-based test of a water-cooled Li 2 O blanket concept. The mechanical configuration of the test piece is designed to simulate a unit cell of a breeder-outside-tube concept. This test piece will be placed in a fission test reactor, which provides an environment similar to that in a fusion reactor. The neutron/gamma flux from the reactor produces prototypical power density, tritium production rates, and operating temperatures and stresses. Steady-state tritium recovery from the test piece can be attained in short-duration (5-to-6-day) tests. The capabilities of this test indicate that fission-based testing can provide important near-term engineering information to support the development of fusion technology

  2. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1994-11-01

    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.) [de

  3. Manufacturing aspects in the design of the breeder unit for Helium Cooled Pebble Bed blankets

    International Nuclear Information System (INIS)

    Rey, J.; Ihli, T.; Filsinger, D.; Polixa, C.

    2007-01-01

    The breeding blanket programme has been in the focus of European fusion research for more than a decade. Recently, it has been driven by the EU Power Plant Conceptual Study (PPCS), investigating the potential of fusion energy in a future economic environment. On the way to the first commercial nuclear fusion reactor (DEMO) new studies for reactor in-vessel components have been initiated. One central focus is the design and manufacturing of the blankets that have to ensure the breeding process to maintain the fuel cycle and are also responsible for the extraction of the main part of the reactor heat for power generation. Two kinds are established: One is the Helium Cooled Pebble Bed (HCPB) and the other the Helium Cooled Liquid Lead (HCLL) blanket. Both designs employ three different cooling plate assemblies. The outer, cooled U-shaped shell, namely the First Wall (FW), with two caps builds the blanket box. The structural strength of the blanket box is realized by integrating Stiffening Grids (SG) that separate the equally spaced Breeder Unit (BU) and allow the box, in case of faulted conditions, to withstand an internal pressure of 8 MPa. The cooled SG constitute the side walls of the BU and are also cooled. The BU consists of a dedicated Cooling Plate (CP) assembly. In present studies about the fabrication of Cooling Plates two kinds of diffusion welding processes are focused on. One is based on a Hot Isostatic Gas Process (HIP). The second is a uni-axial Diffusion Welding Process (DWP). In both cases the bond between the two halves of the cooling plate structure is reached by controlled pressure and heat cycles. Approaching larger, realistic scaled components the uncertainty of ensuring uniform process parameters across the bonding zone increases the risk of defect sources and, therefore, makes it difficult to guarantee the required bonding penetration. This study presents an alternative manufacturing strategy. The premises for this strategy are the reduction of

  4. Materials data base and design equations for the UCLA solid breeder blanket

    International Nuclear Information System (INIS)

    Sharafat, S.; Amodeo, R.; Ghoniem, N.M.

    1986-02-01

    The materials and properties investigated for this blanket study are listed. The phenomenological equations and mathematical fits for all materials and properties considered are given. Efforts to develop a swelling equation based on the few experimental data points available for breeder materials are described. The sintering phenomena for ceramics is investigated

  5. Analysis of mechanical effects caused by plasma disruptions in the European breeder out of tube solid breeder blanket design with MANET as structural material

    International Nuclear Information System (INIS)

    Boccaccini, L.V.; Ruatto, P.

    1995-01-01

    In this paper we deal with some aspects related to the mechanical behaviour of the European breeder out of tube solid breeder blanket for the DEMO reactor during plasma disruptions. The first aspect regards the properties of the martensitic steel MANET which has been chosen as structural material. MANET is a magnetic material and its fracture toughness properties degrade considerably under irradiation. These two features have been taken into account in the calculation of magentic forces and in the assessment of conditions of unstable crack propagation respectively. As second aspect, a comparison between an electrically segmented and a continuous blanket design has been performed. The analysis reveals lower mechanical stresses for the second design during the DEMO reference disruption and in case of faster disruptions. (orig.)

  6. Materials data base and design equations for the UCLA solid breeder blanket

    International Nuclear Information System (INIS)

    Sharafat, S.; Amodeo, R.; Ghoniem, N.M.

    1986-01-01

    The need for a complete and coherent material data base for fusion reactor systems has been an important issue for some time now. Since the choices for materials used in fusion reactors are becoming more apparent, it is important to be able to quickly access this data to facilitate reactor design. The philosophy of a data base is one of expansion and modification. This will lead to a constantly growing collection of most recently acquired information. Based on this philosophy special care has been given to the structure, the accessibility and ease of modification. The data base is developed primarily for use on Personal Computers (PC's). In Section 10.2. materials and properties investigated for this blanket study are listed. Section 10.3. is a list of phenomenological equations and mathematical fits for all materials and properties considered. Section 10.4. describes the authors efforts to develop a swelling equations based on the few experimental data points available for breeder materials. In Section 10.5. the sintering phenomena for ceramics is investigated

  7. 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)

  8. 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)

    Norajitra, P.

    1995-06-01

    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 Li 4 SiO 4 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.) [de

  9. Engineering structure design and fabrication process of small sized China helium-cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Wang Zeming; Chen Lu; Hu Gang

    2014-01-01

    Preliminary design and analysis for china helium-cooled solid breeder (CHHC-SB) test blanket module (TBM) have been carried out recently. As partial verification that the original size module was reasonable and the development process was feasible, fabrication work of a small sized module was to be carried out targetedly. In this paper, detailed design and structure analysis of small sized TBM was carried out based on preliminary design work, fabrication process and integrated assembly process was proposed, so a fabrication for the trial engineering of TBM was layed successfully. (authors)

  10. ITER solid breeder blanket materials database

    International Nuclear Information System (INIS)

    Billone, M.C.; Dienst, W.; Noda, K.; Roux, N.

    1993-11-01

    The databases for solid breeder ceramics (Li 2 ,O, Li 4 SiO 4 , Li 2 ZrO 3 and LiAlO 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

  11. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Clement; Malang, S.; Sawan, M.; Dagher, Mohamad; Smolentsev, S.; Merrill, Brad; Youssef, M.; Reyes, Susanna; Sze, Dai Kai; Morley, Neil B.; Sharafat, Shahran; Calderoni, P.; Sviatoslavsky, G.; Kurtz, Richard J.; Fogarty, Paul J.; Zinkle, Steven J.; Abdou, Mohamed A.

    2006-02-01

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17LI is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Ph-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure.

  12. Evaluation of organic moderator/coolants for fusion breeder blankets

    International Nuclear Information System (INIS)

    Romero, J.B.

    1980-03-01

    Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process

  13. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States)]. E-mail: wongc@fusion.gat.com; Malang, S. [Fusion Nuclear Technology Consulting, Linkenheim (Germany); Sawan, M. [University of Wisconsin, Madison, WI (United States); Dagher, M. [University of California, Los Angeles, CA (United States); Smolentsev, S. [University of California, Los Angeles, CA (United States); Merrill, B. [INEEL, Idaho Falls, ID (United States); Youssef, M. [University of California, Los Angeles, CA (United States); Reyes, S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Sze, D.K. [University of California, San Diego, CA (United States); Morley, N.B. [University of California, Los Angeles, CA (United States); Sharafat, S. [University of California, Los Angeles, CA (United States); Calderoni, P. [University of California, Los Angeles, CA (United States); Sviatoslavsky, G. [University of Wisconsin, Madison, WI (United States); Kurtz, R. [Pacific Northwest Laboratory, Richland, WA (United States); Fogarty, P. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Zinkle, S. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Abdou, M. [University of California, Los Angeles, CA (United States)

    2006-02-15

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiC{sub f}/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 deg. C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R and D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

  14. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Clement; Malang, S.; Sawan, M.; Dagher, Mohamad; Smolentsev, S.; Merrill, Brad; Youssef, M.; Reyes, Susanna; Sze, Dai Kai; Morley, Neil B.; Sharafat, Shahran; Calderoni, P.; Sviatoslavsky, G.; Kurtz, Richard J.; Fogarty, Paul J.; Zinkle, Steven J.; Abdou, Mohamed A.

    2006-07-05

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperture of 700C. We have identified critical issues for the concept, some of which inlude the first wall design, the assessment of MHD effectrs with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time, we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

  15. Overview of EU activities on DEMO liquid metal breeder blanket

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Malang, S.; Reimann, J.; Perujo, A.

    1994-01-01

    The present paper gives an overview of both design and experimental activities within the European Union (EU) concerning the development of liquid metal breeder blankets for DEMO. After several years of studies on breeding blankets, two blanket concepts are presently considered, both using the eutectic Pb-17Li: the dual-coolant concept and the water-cooled concept. The analysis of such concepts has permitted to identify the experimental areas where further data are required. Tritium control and MHD-issues are, at present, the activities on which is devoted the greatest effort within the EU. (authors). 4 figs., 4 tabs., 39 refs

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

    International Nuclear Information System (INIS)

    Jackson, D.P.; Selander, W.N.; Townes, B.M.

    1985-01-01

    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

  17. Breeding blanket development. Tritium release from breeder

    International Nuclear Information System (INIS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nagao, Yoshiharu

    2006-01-01

    Engineering data on neutron irradiation performance of tritium breeders are needed to design the breeding blanket of fusion reactor. In this study, tritium release experiments of the breeders were carried out to examine the effects of various parameters (such as sweep gas flow rate, hydrogen content in sweep gas, irradiation temperature and thermal neutron flux) on tritium generation and release behavior. Lithium titanate (Li 2 TiO 3 ) is considered as a candidate tritium breeder in the blanket design of International Thermonuclear Experimental Reactor (ITER). As for the shape of the breeder material, a small spherical form is preferred to reduce the thermal stress induced in the breeder. Li 2 TiO 3 pebbles of about 170g in total weight and with 0.3 and 2 mm in diameter were manufactured by a wet process, and an assembly packed with the binary Li 2 TiO 3 pebbles was irradiated in Japan Materials Testing Reactor (JMTR). The tritium was generated in the Li 2 TiO 3 pebble bed and released from the pebble bed, and was swept downstream using the sweep gas for on-line analysis of tritium content. Concentration of total tritium and gaseous tritium (HT or T 2 gas) released from the Li 2 TiO 3 pebble bed were measured by ionization chambers, and the ratio of (gaseous tritium)/(total tritium) was evaluated. The sweep gas flow rate was changed from 100 to 900cm 3 /min, and hydrogen content in the sweep gas was changed from 100 to 10000 ppm. Furthermore, thermal neutron flux was changed using a window made of hafnium (Hf) neutron absorber. The irradiation temperature at an outer region of the Li 2 TiO 3 pebble bed was held between 200 and 400degC. The main results of this experiment are summarized as follows. 1) When the temperature at the outside edge of the Li 2 TiO 3 pebble bed exceeded 100degC, the tritium release from the Li 2 TiO 3 pebble bed started. The ratio of the tritium release rate and the tritium generation rate (normalized tritium release rate: R/G) reached

  18. Tritium behaviour in ceramic breeder blankets

    International Nuclear Information System (INIS)

    Miller, J.M.

    1989-01-01

    Tritium release from the candidate ceramic materials, Li 2 O, LiA10 2 , Li 2 SiO 3 , Li 4 SiO 4 and Li 2 ZrO 3 , is being investigated in many blanket programs. Factors that affect tritium release from the ceramic into the helium sweep gas stream include operating temperature, ceramic microstructure, tritium transport and solubility in the solid. A review is presented of the material properties studied and of the irradiation programs and the results are summarized. The ceramic breeder blanket concept is briefly reviewed

  19. An alternative high breeding radio design concept with liquid breeder for the NET/INTOR blanket

    International Nuclear Information System (INIS)

    Avanzini, P.G.; Cardella, A.; Raia, G.; Rosatelli, F.; Farfaletti-Casali, F.

    1984-01-01

    A liquid lithium tubolar breeding blanket concept has been studied which could be applied to NET/INTOR or other next generation Tokamak reactors. A high breeding ratio can be achieved using a moderator medium, without enriching lithium in the Li6 percentage. Preliminary neutron and gamma flux and thermohydraulics calculations have shown the feasibility and efficiency of our concept. (author)

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

    International Nuclear Information System (INIS)

    Tanigawa, Hisashi; Enoeda, Mikio

    2010-03-01

    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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, Hisashi; Enoeda, Mikio [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki (Japan)

    2010-03-15

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

  2. Processing and waste disposal representative for fusion breeder blanket systems

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1987-01-01

    This study is an evaluation of the waste handling concepts applicable to fusion breeder systems. Its goal is to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under US Nuclear Regulatory regulations. The radionuclides expected in the materials used in fusion reactor blankets are described, as are plans for reprocessing and disposal of the components of different breeder blankets. An estimate of the operating costs involved in waste disposal is made

  3. Processing and waste disposal needs for fusion breeder blankets system

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1988-01-01

    We evaluated the waste disposal and recycling requirements for two types of fusion breeder blanket (solid and liquid). The goal was to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under U.S. Nuclear Regulatory Commission regulations. Described in this paper are the radionuclides expected in fusion blanket materials, plans for reprocessing and disposal of blanket components, and estimates for the operating costs involved in waste disposal. (orig.)

  4. Tritium inventory and permeation in liquid breeder blankets

    International Nuclear Information System (INIS)

    Reiter, F.

    1990-01-01

    This report reviews studies of the transport of hydrogen isotopes in the DEMO relevant water-cooled Pb-17Li blanket to be tested in NET and in a self-cooled blanket which uses Pb-17Li or Flibe as a liquid breeder material and V or Fe as a first wall material. The time dependences of tritium inventory and permeation in these blankets and of deuterium and tritium recycling in the self-cooled blanket are presented and discussed

  5. Analysis of mechanical effects caused by plasma disruptions in the European BOT solid breeder blanket design with MANET as structural material

    International Nuclear Information System (INIS)

    Boccaccini, L.V.; Ruatto, P.

    1994-01-01

    The Karlsruhe Nuclear Center is developing, through design and experimental work, a BOT (Breeder Out of Tube) Helium Cooled Solid Breeder Blanket for a DEMO application. One of the crucial problems in the blanket design is to demonstrate the capability of the structure to withstand the mechanical effects of a major plasma disruption as extrapolated to DEMO from the experience of present machines. In this paper the results of the assessment work are presented; the acceptability of the design is discussed on the basis of a stress analysis of the structure under combined thermal and electromagnetic loads. The martensitic steel MANET has been chosen as structural material, because it is able to withstand the high neutron fluence in Demo (70 dpa) without appreciably swelling and has good thermal-mechanical properties - lower thermal expansion and higher strength - in comparison to AISI 316L steel. As far as it concerns the mechanical effects of plasma disruptions, MANET presents two important features which have been carefully investigated in the assessment: the magnetic properties of the material and the degradation of the fracture toughness behavior under irradiation

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

    International Nuclear Information System (INIS)

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A.

    1986-09-01

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

  7. Fusion blanket design and optimization techniques

    International Nuclear Information System (INIS)

    Gohar, Y.

    2005-01-01

    In fusion reactors, the blanket design and its characteristics have a major impact on the reactor performance, size, and economics. The selection and arrangement of the blanket materials, dimensions of the different blanket zones, and different requirements of the selected materials for a satisfactory performance are the main parameters, which define the blanket performance. These parameters translate to a large number of variables and design constraints, which need to be simultaneously considered in the blanket design process. This represents a major design challenge because of the lack of a comprehensive design tool capable of considering all these variables to define the optimum blanket design and satisfying all the design constraints for the adopted figure of merit and the blanket design criteria. The blanket design techniques of the First Wall/Blanket/Shield Design and Optimization System (BSDOS) have been developed to overcome this difficulty and to provide the state-of-the-art techniques and tools for performing blanket design and analysis. This report describes some of the BSDOS techniques and demonstrates its use. In addition, the use of the optimization technique of the BSDOS can result in a significant blanket performance enhancement and cost saving for the reactor design under consideration. In this report, examples are presented, which utilize an earlier version of the ITER solid breeder blanket design and a high power density self-cooled lithium blanket design for demonstrating some of the BSDOS blanket design techniques

  8. Reducing beryllium content in mixed bed solid-type breeder blankets

    Energy Technology Data Exchange (ETDEWEB)

    Shimwell, J., E-mail: mail@jshimwell.com [Department of Physics and Astronomy, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom); Lilley, S.; Morgan, L.; Packer, L.; Kovari, M.; Zheng, S. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); McMillan, J. [Department of Physics and Astronomy, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom)

    2016-11-01

    Highlights: • The ratio of breeder ceramic to neutron multiplier of breeder blankets was varied linearly with depth. • Blankets with varying composition were found to perform better than uniform composition breeder blankets. • It was also possible to reduce the amount of beryllium required by the blanket. - Abstract: Beryllium (Be) is a precious resource with many high value uses, the low energy threshold (n,2n) reaction makes Be an excellent neutron multiplier for use in fusion breeder blankets. Estimates of Be requirements and available resources suggest that this could represent a major supply difficulty for solid-type blanket concepts. Reducing the quantity of Be required by breeder blankets would help to alleviate the problem to some extent. In addition, it is important that the reduction in the Be quantity does not diminish the blanket's performance in key aspects such as the tritium breeding ratio (TBR), energy multiplication and peak nuclear heating. Mixed pebble bed designs allow for the multiplier fraction to be varied throughout the blanket. This neutronics study used MCNP 6 to investigate linear variations of the multiplier fraction in relation to blanket depth, in order to better utilise the important multiplying Be(n,2n) and breeding reactions. Blankets with a uniform multiplier fraction showed little scope for reduction in Be mass. Blankets with varying multiplier fractions were able to simultaneously use 10% less Be, increase the energy amplification by 1%, reduce the peak heating by 7% and maintaining a sufficient TBR when compared to the performance achievable using a uniform composition.

  9. Japanese contributions to ITER testing program of solid breeder blankets for DEMO

    International Nuclear Information System (INIS)

    Kuroda, Toshimasa; Yoshida, Hiroshi; Takatsu, Hideyuki; Maki, Koichi; Mori, Seiji; Kobayashi, Takeshi; Suzuki, Tatsushi; Hirata, Shingo; Miura, Hidenori.

    1991-04-01

    ITER Conceptual Design Activity (CDA), which has been conducted by four parties (Japan, EC, USA and USSR) since May 1988, has been finished on December 1990 with a great achievement of international design work of the integrated fusion experimental reactor. Numerous issues of physics and technology have been clarified for providing a framework of the next phase of ITER (Engineering Design Activity; EDA). Establishment of an ITER testing program, which includes technical test issues of neutronics, solid breeder blankets, liquid breeder blankets, plasma facing components, and materials, has been one of the goals of the CDA. This report describes Japanese proposal for the testing program of DEMO/power reactor blanket development. For two concepts of solid breeder blanket (helium-cooled and water-cooled), identification of technical issues, scheduling of test program, and conceptual design of test modules including required test facility such as cooling and tritium recovery systems have been carried out as the Japanese contribution to the CDA. (author)

  10. Investigation of tritium inventory and permeation behaviour in the liquid breeder blanket concept of Demo as a function of design and material parameters

    International Nuclear Information System (INIS)

    Tominetti, S.; Perujo, A.; Reiter, F.

    1991-01-01

    A numerical code has been used to estimate the time dependence of tritium inventory and of tritium transport into the coolant, into the first wall boxes and through the liquid breeder in the Pb-17Li blanket concept of DEMO. Several issues in both design and material parameters have been considered and the effect on inventory and permeation of coatings with low surface recombination coefficient and/or low diffusivity at various surfaces of the structural material has been studied. TiC has been chosen as reference material for these calculations and a general database on coating efficiency as a function of its properties has also been produced on the basis of TiC data

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

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2004-07-01

    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 Li 2 TiO 3 and so on, fabrication technology developments and characterization of the Li 2 TiO 3 and Li 4 SiO 4 pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li 2 TiO 3 and Li 4 SiO 4 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.)

  12. Stress analysis of the tokamak engineering test breeder blanket

    International Nuclear Information System (INIS)

    Huang Zhongqi

    1992-01-01

    The design features of the hybrid reactor blanket and main parameters are presented. The stress analysis is performed by using computer codes SAP5p and SAP6 with the three kinds of blanket module loadings, i.e, the pressure of coolant, the blanket weight and the thermal loading. Numerical calculation results indicate that the stresses of the blanket are smaller than the allowable ones of the material, the blanket design is therefore reasonable

  13. Fusion reactor blanket-main design aspects

    International Nuclear Information System (INIS)

    Strebkov, Yu.; Sidorov, A.; Danilov, I.

    1994-01-01

    The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm 2 and 7 MWa/m 2 accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket

  14. Ferritic steels for the first generation of breeder blankets

    International Nuclear Information System (INIS)

    Diegele, E.

    2009-01-01

    Materials development in nuclear fusion for in-vessel components, i.e. for breeder blankets and divertors, has a history of more than two decades. It is the specific in-service and loading conditions and the consequentially required properties in combination with safety standards and social-economic demands that create a unique set of specifications. Objectives of Fusion for Energy (F4E) include: 1) To provide Europe's contribution to the ITER international fusion energy project; 2) To implement the Broader Approach agreement between Euratom and Japan; 3) To prepare for the construction and demonstration of fusion reactors (DEMO). Consequently, activities in F4E focus on structural materials for the first generations of breeder blankets, i.e. ITER Test Blanket Modules (TBM) and DEMO, whereas a Fusion Materials Topical Group implemented under EFDA coordinates R and D on physically based modelling of irradiation effects and R and D in the longer term (new and /or higher risk materials). The paper focuses on martensitic-ferritic steels and (i) reviews briefly the challenges and the rationales for the decisions taken in the past, (ii) analyses the status of the main activities of development and qualification, (iii) indicates unresolved issues, and (iv) outlines future strategies and needs and their implications. Due to the exposure to intense high energy neutron flux, the main issue for breeder materials is high radiation resistance. The First Wall of a breeder blanket should survive 3-5 full power years or, respectively in terms of irradiation damage, typically 50-70 dpa for DEMO and double figures for a power plant. Even though the objective is to have the materials and key fabrication technologies needed for DEMO fully developed and qualified within the next two decades, a major part of the task has to be completed much earlier. Tritium breeding test blanket modules will be installed in ITER with the objective to test DEMO relevant technologies in fusion

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

    International Nuclear Information System (INIS)

    Noda, Kenji

    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 2 TiO 3 , tritium release behavior of Li 2 TiO 3 and Li 2 ZrO 3 including tritium diffusion, modeling of tritium release from Li 2 ZrO 3 in ITER condition, helium release behavior from Li 2 O, results of tritium release irradiation tests of Li 4 SiO 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.)

  16. Comparison of inventory of tritium in various ceramic breeder blankets

    International Nuclear Information System (INIS)

    Nishikawa, M.; Beloglazov, S.; Nakashima, N.; Hashimoto, K.; Enoeda, M.

    2002-01-01

    It has been pointed out by the present authors that it is essential to understand such mass transfer steps as diffusion of tritium in the grain of breeder material, absorption of water vapor into bulk of the grain, and adsorption of water on surface of the grain, together with the isotope exchange reaction between hydrogen in purge gas and tritium on surface of breeder material and the isotope exchange reaction between water vapor in purge gas and tritium on surface, for estimation of the tritium inventory in a uniform ceramic breeder blanket under the steady-state condition. It has been also pointed out by the present authors that the water formation reaction on the surface of ceramic breeder materials at introduction of hydrogen can give effect on behavior of bred tritium and lithium transfer in blanket. The tritium inventory for various ceramic breeder blankets are compared in this study basing on adsorption capacity, absorption capacity, isotope exchange capacity, and isotope exchange reactions on the Li 2 O, LiAlO 2 , Li 2 ZrO 3 , Li 4 SiO 4 and Li 2 TiO 3 surface experimentally obtained by the present authors. Effect of each mass transfer steps on the shape of release curve of bred tritium at change of the operational conditions is also discussed from the observation at out pile experiment in KUR. (orig.)

  17. Nuclear and thermal analyses of supercritical-water-cooled solid breeder blanket for fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yanagi, Yoshihiko; Sato, Satoshi; Enoeda, Mikio; Hatano, Toshihisa; Kikuchi, Shigeto; Kuroda, Toshimasa; Kosaku, Yasuo; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2001-11-01

    Within a design study of a fusion DEMO reactor aiming at demonstrating technologies of fusion power plant, supercritical water is applied as a coolant of solid breeder blanket to attain high thermal efficiency. The blanket has multi-layer composed of solid breeder pebbles (Li{sub 2}O) and neutron multiplier pebbles (Be) which are radially separated by cooling panels. The first wall and the breeding region are cooled by supercritical water below and above the pseudo-critical temperature, respectively. Temperature distribution and tritium breeding ratio (TBR) have been estimated by one-dimensional nuclear and thermal calculations. The local TBR as high as 1.47 has been obtained after optimization of temperature distribution in the breeder region under the following conditions: neutron wall loading of 5 MW/m{sup 2}, {sup 6}Li enrichment of 30% and coolant temperature at inlet of breeder region of 380degC. In the case of the higher coolant temperature 430degC of the breeder region the local TBR was reduced to be 1.40. This means that the net TBR higher than 1.0 could be expected with the supercritical-water-cooled blanket, whose temperature distribution in the breeder region would be optimized by following the coolant temperature, and where a coverage of the breeder region is assumed to be 70%. (author)

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

    International Nuclear Information System (INIS)

    Ishitsuka, E.

    2002-01-01

    Advanced solid breeding blanket design in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high dose of neutron irradiation. Therefore, the development of such advanced blanket materials is indispensable. In this paper, the cooperation activities among JAERI, universities and industries in Japan on the development of these advanced materials are reported. Advanced tritium breeding material to prevent the grain growth in high temperature had to be developed because the tritium release behavior degraded by the grain growth. As one of such materials, TiO 2 -doped Li 2 TiO 3 has been studied, and TiO 2 -doped Li 2 TiO 3 pebbles was successfully fabricated. For the advanced neutron multiplier, the beryllium intermetallic compounds that have high melting point and good chemical stability have been studied. Some characterization of Be 12 Ti was studied. The pebble fabrication study for Be 12 Ti was also performed and Be 12 Ti pebbles were successfully fabricated. From these activities, the bright prospect to realize the DEMO blanket by the application of TiO 2 -doped Li 2 TiO 3 and beryllium intermetallic compounds was obtained. (author)

  19. Assessment of First Wall and Blanket Options with the Use of Liquid Breeder

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Malang, S.; Sawan, M.

    2005-01-01

    As candidate blanket concepts for a U.S. advanced reactor power plant design, with consideration of the time frame for ITER development, we assessed first wall and blanket design concepts based on the use of reduced activation ferritic steel as structural material and liquid breeder as the coolant and tritium breeder. The liquid breeder choice includes the conventional molten salt Li 2 BeF 4 and the low melting point molten salts such as LiBeF 3 and LiNaBeF 4 (FLiNaBe). Both self-cooled and dual coolant molten salt options were evaluated. We have also included the dual coolant leadeutectic Pb-17Li design in our assessment. We take advantage of the molten salt low electrical and thermal conductivity to minimize impacts from the MHD effect and the heat losses from the breeder to the actively cooled steel structure. For the Pb-17Li breeder we employ flow channel inserts of SiC f /SiC composite with low electrical and thermal conductivity to perform respective insulation functions. We performed preliminary assessments of these design options in the areas of neutronics, thermal-hydraulics, safety, and power conversion system. Status of the R and D items of selected high performance blanket concepts is reported. Results from this study will form the technical basis for the formulation of the U.S. ITER test module program and corresponding test plan

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

    International Nuclear Information System (INIS)

    Paola Batistoni, P.; Angelone, M.; Bettinali, L.

    2006-01-01

    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 Li 2 CO 3 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 Li 2 CO 3 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

  1. Preliminary RAMI analysis of WCLL blanket and breeder systems

    International Nuclear Information System (INIS)

    Arroyo, Jose Manuel; Brown, Richard; Harman, Jon; Rosa, Elena; Ibarra, Angel

    2015-01-01

    Highlights: • Preliminary RAMI model for WCLL has been developed. • Critical parts and parameters influencing WCLL availability have been focused. • Necessary developments of tools/models to represent system performance have been identified. - Abstract: DEMO will be a prototype fusion reactor designed to prove the capability to produce electrical power in a commercially acceptable way. One of the key factors in that endeavor is the achievement of certain level of plant availability. Therefore, RAMI (Reliability, Availability, Maintainability and Inspectability) will be a key element in the engineering development of DEMO. Some studies have been started so as to develop the tools and models to assess different design alternatives from RAMI point of view. The main objective of these studies is to be able to evaluate the influence of different parameters on DEMO availability and to focus the critical parts that should be further researched and improved in order to develop a high-availability oriented DEMO design. A preliminary RAMI analysis of the Water Cooled Lithium-Lead (WCLL) blanket and breeder concept for DEMO has been developed. The amounts of single elements that may fail (e.g. more than 180,000 C-shaped tubes) and the mean down time associated to failures inside the vacuum vessel (around 3 months) have been highlighted as the critical parameters influencing the system availability. On the other hand, the necessary developments of tools/models to better represent the system performance have been identified and proposed for future work.

  2. Preliminary RAMI analysis of WCLL blanket and breeder systems

    Energy Technology Data Exchange (ETDEWEB)

    Arroyo, Jose Manuel, E-mail: josemanuel.arroyo@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, Madrid (Spain); Brown, Richard [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon (United Kingdom); Harman, Jon [EFDA Close Support Unit, Garching (Germany); Rosa, Elena; Ibarra, Angel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, Madrid (Spain)

    2015-10-15

    Highlights: • Preliminary RAMI model for WCLL has been developed. • Critical parts and parameters influencing WCLL availability have been focused. • Necessary developments of tools/models to represent system performance have been identified. - Abstract: DEMO will be a prototype fusion reactor designed to prove the capability to produce electrical power in a commercially acceptable way. One of the key factors in that endeavor is the achievement of certain level of plant availability. Therefore, RAMI (Reliability, Availability, Maintainability and Inspectability) will be a key element in the engineering development of DEMO. Some studies have been started so as to develop the tools and models to assess different design alternatives from RAMI point of view. The main objective of these studies is to be able to evaluate the influence of different parameters on DEMO availability and to focus the critical parts that should be further researched and improved in order to develop a high-availability oriented DEMO design. A preliminary RAMI analysis of the Water Cooled Lithium-Lead (WCLL) blanket and breeder concept for DEMO has been developed. The amounts of single elements that may fail (e.g. more than 180,000 C-shaped tubes) and the mean down time associated to failures inside the vacuum vessel (around 3 months) have been highlighted as the critical parameters influencing the system availability. On the other hand, the necessary developments of tools/models to better represent the system performance have been identified and proposed for future work.

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

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Y.; Tobita, K.; Utoh, H.; Hoshino, K.; Asakura, N.; Nakamura, M.; Tanigawa, H.; Mikio, E.; Tanigawa, H.; Nakamichi, M.; Hoshino, T., E-mail: someya.yoji@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan)

    2012-09-15

    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)

  4. Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR

    Science.gov (United States)

    Xiaokang, ZHANG; Songlin, LIU; Xia, LI; Qingjun, ZHU; Jia, LI

    2017-11-01

    The water cooled ceramic breeder (WCCB) blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor (CFETR). Some updating of neutronics analyses was needed, because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket, including the optimization of radial build-up and customized structure for each blanket module. A 22.5 degree toroidal symmetrical torus sector 3D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses. The tritium breeding capability, nuclear heating power, radiation damage, and decay heat were calculated by the MCNP and FISPACT code. The results show that the packing factor and 6Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency. The nuclear heating power of the blanket under 200 MW fusion power reaches 201.23 MW. The displacement per atom per full power year (FPY) of the plasma-facing component and first wall reach 0.90 and 2.60, respectively. The peak H production rate reaches 150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module #3. The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m-3 at that time. The decay heat density of the blanket modules slowly decreases to lower than 10 W m-3 in more than ten years.

  5. ITER blanket designs

    International Nuclear Information System (INIS)

    Gohar, Y.; Parker, R.; Rebut, P.H.

    1995-01-01

    The ITER first wall, blanket, and shield system is being designed to handle 1.5±0.3 GW of fusion power and 3 MWa m -2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design. (orig.)

  6. Self-cooled blanket concepts using Pb-17Li as liquid breeder and coolant

    International Nuclear Information System (INIS)

    Malang, S.; Deckers, H.; Fischer, U.; John, H.; Meyder, R.; Norajitra, P.; Reimann, J.; Reiser, H.; Rust, K.

    1991-01-01

    A blanket design concept using Pb-17Li eutectic alloy as both breeder material and coolant is described. Such a self-cooled blanket for the boundary conditions of a DEMO-reactor is under development at the Kernforschungszentrum Karlsruhe (KfK) in the frame of the European blanket development program. Results of investigations in the areas of design, neutronics, magneto-hydrodynamics, thermo-mechanics, ancillary loop systems, and safety are reported. Based on recent progress, it can be concluded that the boundary conditions of a DEMO-reactor can be met, tritium self-sufficiency can be obtained without using beryllium as an additional neutron multiplier, and tritium inventory and permeation are acceptably low. However, to complete judge the feasibility of the proposed concept, further studies are necessary to obtain a better understanding of the magneto-hydrodynamic phenomena and their effects on the thermal-hydraulic performance of a fusion reactor blanket. (orig.)

  7. 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.)

  8. Composite beryllium-ceramics breeder pin elements for a gas cooled solid blanket

    International Nuclear Information System (INIS)

    Carre, F.; Chevreau, G.; Gervaise, F.; Proust, E.

    1986-06-01

    Helium coolant have main advantages compared to water for solid blankets. But limitations exist too and the development of attractive helium cooled blankets based on breeder pin assemblies has been essentially made possible by the derivation from recent CEA neutronic studies of an optimized composite beryllium/ceramics breeder arrangement. Description of the proposed toroidal blanket layout for Net is made together with the analysis of its main performance. Merits of the considered composite Be/ceramics breeder elements are discussed

  9. Conceptual design on interface between ITER and tritium extraction system of Chinese helium-cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Zhang Long; Luo Tianyong; Feng Kaiming

    2010-01-01

    Tritium extraction system is essential for CN HCSB TBM for safety and technical reasons. Based on the assessments of system functions, integration issues and safety considerations, two main modifications of the system from previous design (Feng et al., 2007 ; Chen et al., 2008 ) are adopted: a)the TES has been split to 2 parts with one in port cell and another in tritium building. Q 2 O in the purge gas is reduced to Q 2 in a hot metal bed located in port cell; Q 2 is separated from the stream by a pair of cryogenic molecular sieve beds and a Pd/Ag diffuser located in tritium building. b)isotope separation process has been excluded. TES components sizes are estimated and space allocations are estimated. Required services and where and when they are needed are preliminary defined. Fluids delivered towards ITER tritium system are analyzed.

  10. 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.

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

    International Nuclear Information System (INIS)

    Patterson-Hine, F.A.

    1984-05-01

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

  12. A solid-breeder blanket and power conversion system for the Mirror Advanced Reactor Study (MARS)

    International Nuclear Information System (INIS)

    Bullis, R.; Clarkson, I.

    1983-01-01

    A solid-breeder blanket has been designed for a commercial fusion power reactor based on the tandem mirror concept (MARS). The design utilizes lithium oxide, cooled by helium which powers a conventional steam electric generating cycle. Maintenance and fabricability considerations led to a modular configuration 6 meters long which incorporates two magnets, shield, blanket and first wall. The modules are arranged to form the 150 meter long reactor central cell. Ferritic steel is used for the module primary structure. The lithium oxide is contained in thin-walled vanadium alloy tubes. A tritium breeding ratio of 1.25 and energy multiplication of 1.1 is predicted. The blanket design appears feasible with only a modest advance in current technology

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

    Energy Technology Data Exchange (ETDEWEB)

    Berwald, D.H.; Whitley, R.H.; Garner, J.K.; Gromada, R.J.; McCarville, T.J.; Moir, R.W.; Lee, J.D.; Bandini, B.R.; Fulton, F.J.; Wong, C.P.C.; Maya, I.; Hoot, C.G.; Schultz, K.R.; Miller, L.G.; Beeston, J.M.; Harris, B.L.; Westman, R.A.; Ghoniem, N.M.; Orient, G.; Wolfer, M.; DeVan, J.H.; Torterelli, P.

    1985-09-01

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

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

    International Nuclear Information System (INIS)

    Berwald, D.H.; Whitley, R.H.; Garner, J.K.

    1985-09-01

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

  15. Concept and nuclear performance of direct-enrichment fusion breeder blanket using UO2 powder

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Kasahara, Takayasu; An, Shigehiro

    1985-01-01

    A new concept is presented for direct enrichment of fissile fuel in the blanket of a fusion-fission hybrid reactor. The enriched fuel produced by this means can be used in fission reactors without reprocessing. The outstanding feature of the concept is the powdered form in which UO 2 fuel is placed in the reactor blanket, where it is irradiated to the requisite enrichment for use as fuel in burner reactor, e.g. 3%. After removal from blanket, the powder is mixed to homogenize the enrichment. Fuel pellets and assemblies are then fabricated from the powder without reprocessing. The concept of irradiating UO 2 in powder eliminates the problems of spatial nonuniformity in fissile enrichment, and of radiation damage to fuel clad, encountered in attempting to enrich prefabricated fuel. Powder mixing for homogenization brings the additional benefit of removing volatile fission products. Also burnable poison can be added, as necessary, after irradiation. An extensive neutronic parameter survey showed that the optimum blanket arrangement for this enrichment concept is one presenting a fission suppressing configuration and with beryllium adopted as moderator. By this arrangement, the average 239 Pu enrichment obtained on the natural UO 2 fuel in the blanket reaches 3% after only 0.56 MW.yr/m"2 exposure. A conceptual design is presented of the blanket, together with associated fusion breeder, from which, practical application of the concept is shown to be promising. (author)

  16. Novel blanket design for ICTR's

    International Nuclear Information System (INIS)

    Abdel-Khalik, S.I.; Conn, R.W.; Wolfer, W.G.; Larsen, E.N.; Sviatoslavsky, I.N.

    1978-01-01

    A novel blanket design for ICTRs is described. This blanket is used in SOLASE, the conceptual laser fusion reactor of the University of Wisconsin. The blanket to be described offers numerous advantages, including low cost, low weight, low induced radioactivity levels, the potential for hands-on maintenance, modular construction, low pressure, ability to decouple first wall and blanket coolant temperatures, adequate breeding, low tritium inventory and leakage, and sufficiently long life

  17. Progress of R&D on water cooled ceramic breeder for ITER test blanket system and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Yoshinori, E-mail: kawamura.yoshinori@jaea.go.jp [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Enoeda, Mikio [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Sato, Satoshi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Ochiai, Kentaro [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Konno, Chikara; Edao, Yuki; Hayashi, Takumi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Yamanishi, Toshihiko [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2016-11-01

    Highlights: • Thermo-hydraulic calculation in the TBM at the water ingress event has been done. • Shielding calculations for the ITER equatorial port #18 were conducted by using C-lite model. • Prototypic pebbles of Be{sub 17}Ti{sub 2} and Be{sub 12}V had a good oxidation property similar to Be{sub 12}Ti pebble. • Li rich Li{sub 2}TiO{sub 3} pebbles were successfully fabricated using the emulsion method by controlling sintering atmosphere. • New tritium production/recovery experiments at FNS have been started by using ionization chamber as on-line gas monitor. - Abstract: The development of a water cooled ceramic breeder (WCCB) test blanket module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and development of DEMO blanket, R&D has been performed on the module fabrication technology, breeder and multiplier pebble fabrication technology, tritium production rate evaluation, as well as structural and safety design activities. The fabrication of full-scale first wall, side walls, breeder pebble bed box and back wall was completed, and assembly of TBM with box structure was successfully achieved. Development of advanced breeder and multiplier pebbles for higher chemical stability was continued for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium transport simulation technology, investigation of the TBM neutron measurement technology and the evaluation of the tritium production and recovery test using D-T neutron in the fusion neutron source (FNS) facility has been performed. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

  18. Development of welding technologies for the manufacturing of European Tritium Breeder blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Poitevin, Y., E-mail: yves.poitevin@f4e.europa.eu [Fusion for Energy (F4E), Barcelona (Spain); Aubert, Ph. [CEA Saclay, DEN/DM2S and DEN/DMN, F-91191 Gif-sur-Yvette (France); Diegele, E. [Fusion for Energy (F4E), Barcelona (Spain); Dinechin, G. de [CEA Saclay, DEN/DM2S and DEN/DMN, F-91191 Gif-sur-Yvette (France); Rey, J. [Institut fuer Neutronenphysik und Reaktortechnik, FZK, Karlsruhe (Germany); Rieth, M. [Institut fuer Materialforschung I, FZK, Karlsruhe (Germany); Rigal, E. [CEA Grenoble, DRT/DTH, F-38000 Grenoble (France); Weth, A. von der [Institut fuer Neutronenphysik und Reaktortechnik, FZK, Karlsruhe (Germany); Boutard, J.-L. [European Fusion Development Agreement (EFDA), Garching (Germany); Tavassoli, F. [CEA Saclay, DEN/DM2S and DEN/DMN, F-91191 Gif-sur-Yvette (France)

    2011-10-01

    Europe has developed two reference Tritium Breeder Blankets concepts for a DEMO fusion reactor: the Helium-Cooled Lithium-Lead and the Helium-Cooled Pebble-Bed. Both are using the reduced-activation ferritic-martensitic EUROFER-97 steel as structural material and will be tested in ITER under the form of test blanket modules. The fabrication of their EUROFER structures requires developing welding processes like laser, TIG, EB and diffusion welding often beyond the state-of-the-art. The status of European achievements in this area is reviewed, illustrating the variety of processes and key issues behind retained options, in particular with respect to metallurgical aspects and mechanical properties. Fabrication of mock-ups is highlighted and their characterization and performances with respect to design requirements are reviewed.

  19. Use of nuclear data sensitivity and uncertainty analysis for the design preparation of the HCLL breeder blanket mock-up experiment for ITER

    International Nuclear Information System (INIS)

    Kodeli, I.

    2007-01-01

    An experiment on a mock-up of the Test Blanket module based on Helium Cooled Lithium Lead (HCLL) concept will be performed in 2007 in the FNG utility in Frascati in order to study neutronics characteristics of the module and the performance of the computational tools in the accurate prediction of the neutron transport. With the objective to prepare and optimise the design of the mock-up in the sense to provide maximum information on the state-of-the-art of the cross section data the mock-up was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR), their sensitivity to the underlying basic cross sections, as well as the corresponding uncertainty estimations were calculated using the deterministic transport codes (DOORS package), the sensitivity/uncertainty code package SUSD3D and the VITAMIN-J/COVA covariance matrix libraries. The cross section reactions with largest contribution to the uncertainty in the calculation of the TPR were identified to be (n,2n) and (n,3n) reactions on plumb. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross sections. (author)

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

    Directory of Open Access Journals (Sweden)

    I. Kodeli

    2008-01-01

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

  1. Impact of material system thermomechanics and thermofluid performance on He-cooled ceramic breeder blanket designs with SiCf/SiC

    International Nuclear Information System (INIS)

    Ying, Alice Y.; Yokomine, Takehiko; Shimizu, Akihiko; Abdou, Mohamed; Kohyama, Akira

    2004-01-01

    This paper presents results from a recent effort initiated under the JUPITER-II collaborative program for high temperature gas-cooled blanket systems using SiC f /SiC as a structural material. Current emphasis is to address issues associated with the function of the helium gas considered in the DREAM and ARIES-I concepts by performing thermomechanical and thermofluid analysis. The objective of the analysis is to guide future research focus for a task in the project. It is found that the DREAM concept has the advantage of achieving uniform temperature without threatening blanket pebble bed integrity by differential thermal stress. However, its superiority needs to be further justified by investigating the feasibility and economic issues involved in the tritium extraction technology

  2. Impact of material system thermomechanics and thermofluid performance on He-cooled ceramic breeder blanket designs with SiCf/SiC

    International Nuclear Information System (INIS)

    Ying, A.Y.; Abdou, M.; Yokomine, T.; Shimizu, A.; Kohyama, A.

    2008-01-01

    This paper presents results from a recent effort initiated under the JUPITER-II collaborative program for high temperature gas-cooled blanket systems using SiC/SiC as a structural material. Current emphasis is to address issues associated with the function of the helium gas considered in the DREAM and ARIES-I concepts by performing thermomechanical and thermofluid analysis. The objective of the analysis is to guide future research focus for a task in the project. It is found that the DREAM concept has the advantage of achieving uniform temperature without threatening blanket pebble bed integrity by differential thermal stress. However, its superiority needs to be further justified by investigating the feasibility and economic issues involved in the tritium extraction technology. (author)

  3. Development of an engineering-scale nuclear test of a solid-breeder fusion-blanket concept

    International Nuclear Information System (INIS)

    Deis, G.A.; Bohn, T.S.; Hsu, P.Y.; Miller, L.G.; Scott, A.J.; Watts, K.D.; Welch, E.C.

    1983-08-01

    As part of the Phase I effort on Program Element-II (PE-II) of the Office of Fusion Energy/Argonne National Laboratory First Wall/Blanket/Shield Engineering Technology Program, a study has been performed to develop preconceptual hardware designs and preliminary test program descriptions for two fission-reactor-based tests of a water-cooled, solid-breeder fusion reactor blanket concept. First, a list of potentially acceptable reactor facilities is developed, based on a list of required reactor characteristics. From this set of facilities, two facilities are selected for study: the Oak Ridge Research Reactor (ORR) and the Power Burst Facility (PBF). A test which employs a cylindrical unit cell of a solid-breeder fusion reactor blanket, with pressurized-water cooling is designed for each facility. The test design is adjusted to the particular characteristics of each reactor. These two test designs are then compared on the basis of technical issues and cost. Both tests can satisfy the PE-II mission: blanket thermal hydraulic and thermomechanical issues. In addition, both reactors will produce prototypical tritium production rates and profiles and release characteristics with little or no additional modifications

  4. European DEMO BOT Solid Breeder Blanket: the concept based on the use of cooling plates and beds of beryllium and Li4SiO4 pebbles

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Fischer, U.; Norajitra, P.; Reimann, G.; Reiser, H.

    1995-01-01

    The paper presents an important modification of the European DEMO BOT Solid Breeder Blanket. The new design uses cooling plates rather than tubes. This allows a considerable simplification of the blanket and the separation of the beryllium from the Li 4 SiO 4 pebbles. The neutronic, thermohydraulic and tritium performance of the new design is quite good and equivalent to that of the previous one. (orig.)

  5. Activation analysis and waste management of China ITER helium cooled solid breeder test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Han, J.R., E-mail: hanjingru@163.co [North China Electric Power University, School of Nuclear Science and Engineering, Zhu-Xin-Zhuang, De-Wai, Beijing 102206 (China); Chen, Y.X.; Han, R. [North China Electric Power University, School of Nuclear Science and Engineering, Zhu-Xin-Zhuang, De-Wai, Beijing 102206 (China); Feng, K.M. [Southwestern Institute of Physics, P.O.Box 432, Chengdu 610041 (China); Forrest, R.A. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom)

    2010-08-15

    Activation characteristics have been assessed for the ITER China helium cooled solid breeder (CH-HCSB) 3 x 6 test blanket module (TBM). Taking a representative irradiation scenario, the activation calculations were performed by FISPACT code. Neutron fluxes distributions in the TBM were provided by a preceding MCNP calculation. These fluxes were passed to FISPACT for the activation calculation. The main activation parameters of the HCSB-TBM were calculated and discussed, such as activity, afterheat and contact dose rate. Meanwhile, the dominant radioactivity nuclides and reaction channel pathways have been identified. According to the Safety and Environmental Assessment of Fusion Power (SEAFP) waste management strategy, the activated materials can be re-used following the remote handling recycling options. The results will provide useful indications for further optimization design and waste management of the TBM.

  6. The impact of tritium solubility and diffusivity on inventory and permeation in liquid breeder blankets

    International Nuclear Information System (INIS)

    Caorlin, M.; Gervasini, G.; Reiter, F.

    1988-01-01

    The authors reviewed hydrogen solubility and diffusivity data for liquid lithium-based compounds which are potential breeding blanket materials in NET-type fusion devices. These data have been used to assess tritium permeation and inventory in separately cooled NET blankets and in self cooled blankets with a vanadium first wall. The results for the separately cooled NET-liquid breeder show that tritium permeation is negligible for lithium, a serious problem for Pb-17Li and a critical one for Flibe. The total tritium inventory is lowest in lithium, high in Pb-17Li and very high in Flibe. The high tritium partial pressure for Flibe or Pb-17Li can be reduced in a self cooled blanket with a vanadium first wall. Permeation into the plasma reduces the blanket tritium inventory and permeation. Tritium recovery can be combined with the plasma exhaust

  7. Optimization of the breeder zone cooling tubes of the DEMO Water-Cooled Lithium Lead breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A.; Arena, P.; Bongiovì, G. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy); Chiovaro, P., E-mail: pierluigi.chiovaro@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy); Del Nevo, A. [ENEA Brasimone, Camugnano, BO (Italy); Forte, R. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy)

    2016-11-01

    Highlights: • Determination of an optimal configuration for the breeder zone cooling tubes. • Attention has been focused on the toroidal–radial breeder zone cooling tubes lay out. • A theoretical-computational approach based on the Finite Element Method (FEM) has been followed, adopting a qualified commercial FEM code. • Five different configurations have been investigated to optimize the breeder zone cooling tubes arrangement fulfilling all the rules prescribed by safety codes. - Abstract: The determination of an optimal configuration for the breeder zone (BZ) cooling tubes is one of the most important issues in the DEMO Water-Cooled Lithium Lead (WCLL) breeding blanket R&D activities, since BZ cooling tubes spatial distribution should ensure an efficient heat power removal from the breeder, avoiding hotspots occurrence in the thermal field. Within the framework of R&D activities supported by the HORIZON 2020 EUROfusion Consortium action on the DEMO WCLL breeding blanket design, a campaign of parametric analyses has been launched at the Department of Energy, Information Engineering and Mathematical Models of the University of Palermo (DEIM), in close cooperation with ENEA-Brasimone, in order to assess the potential influence of BZ cooling tubes number on the thermal performances of the DEMO WCLL outboard breeding blanket equatorial module under the nominal steady state operative conditions envisaged for it, optimizing their geometric configuration and taking also into account that a large number of cooling pipes can deteriorate the tritium breeding performances of the module. In particular, attention has been focused on the toroidal-radial option for the BZ tube bundles lay-out and a parametric study has been carried out taking into account different tube bundles arrangement within the module. The study has been carried out following a numerical approach, based on the finite element method (FEM), and adopting a qualified commercial FEM code. Results

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

    International Nuclear Information System (INIS)

    Amici, A.; Anzidei, L.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.; Zampaglione, V.; Petrizzi, L.

    1989-01-01

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

  9. Preliminary Analysis on Decay Heat Removal Capability of Helium Cooled Solid Breeder Test Blanket Module

    International Nuclear Information System (INIS)

    Ahn, Mu Young; Cho, Seung Yon; Kim, Duck Hoi; Lee, Eun Seok; Kim, Hyung Seok; Suh, Jae Seung; Yun, Sung Hwan; Cho, Nam Zin

    2007-01-01

    One of the main ITER goals is to test and validate design concepts of tritium breeding blankets relevant to DEMO or fusion power plants. Korea Helium-Cooled Solid Breeder (HCSB) Test Blanket Module (TBM) has been developed with overall objectives of achieving this goal. The TBM employs high pressure helium to cool down the First Wall (FW), Side Wall (SW) and Breeding Zone (BZ). Therefore, safety consideration is a part of the design process. Each ITER Party performing the TBM program is requested to reach a similar level of confidence in the TBM safety analysis. To meet ITER's request, Failure Mode and Effects Analysis (FMEA) studies have been performed on the TBM to identify the Postulated Initial Event (PIE). Although FMEA on the KO TBM has not been completed, in-vessel, in-box and ex-vessel Loss Of Coolant Accident (LOCA) are considered as enveloping cases of PIE in general. In this paper, accidental analyses for the three selected LOCA were performed to investigate the decay heat removal capability of the TBM. To simulate transient thermo-hydraulic behavior of the TBM for the selected scenarios, RELAP5/MOD3.2 code was used

  10. 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)

    Dalle Donne, M.; Boccaccini, L.V.; Bojarsky, E.; Deckers, H.; Dienst, W.; Doerr, L.; Fischer, U.; Giese, H.; Guenther, E.; Haefner, H.E.; Hofmann, P.; Kappler, F.; Knitter, R.; Kuechle, M.; Moellendorf, U. von; Norajitra, P.; Penzhorn, R.D.; Reimann, G.; Reiser, H.; Schulz, B.; Schumacher, G.; Schwenk-Ferrero, A.; Sordon, G.; Tsukiyama, T.; Wedemeyer, H.; Weimar, P.; Werle, H.; Wiegner, E.; Zimmermann, H.

    1991-10-01

    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.) [de

  11. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Cui, Shijie; Zhang, Dalin; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-01

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

  12. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-15

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

  13. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    1983-10-01

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

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

    International Nuclear Information System (INIS)

    Malang, S.; Reimann, J.; Sebening, H.; Barleon, L.; Bogusch, E.; Bojarsky, E.; Borgstedt, H.U.; Buehler, L.; Casal, V.; Deckers, H.; Feuerstein, H.; Fischer, U.; Frees, G.; Graebner, H.; John, H.; Jordan, T.; Kramer, W.; Krieg, R.; Lenhart, L.; Malang, S.; Meyder, R.; Norajitra, P.; Reimann, J.; Schwenk-Ferrero, A.; Schnauder, H.; Stieglitz, R.; Oschinski, J.; Wiegner, E.

    1991-12-01

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

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

    International Nuclear Information System (INIS)

    John, H.; Malang, S.; Sebening, H.

    1991-12-01

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

  16. Thermal-hydraulic analysis on the whole module of water cooled ceramic breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng; Ma, Xuebin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Lin, Shuang [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Huang, Kai [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China)

    2016-11-15

    Highlights: • The 3D thermal hydraulic analysis on the whole module of WCCB is performed by CFD method. • Temperature field and mass flow distribution have been obtained. • The design of WCCB is reasonable from the perspective of thermal-hydraulics. • The scheme for further optimization has been proposed. - Abstract: The Water Cooled Ceramic Breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). The thermal-hydraulic analysis is essential because the blanket should remove the high heat flux from the plasma and the volumetric heat generated by neutrons. In this paper, the detailed three dimensional (3D) thermal hydraulic analysis on the whole module of WCCB blanket has been performed by Computational Fluid Dynamics (CFD) method, which is capable of solving conjugate heat transfer between solid structure and fluid. The main results, including temperature field, distribution of mass flow rate and coolant pressure drop, have been calculated simultaneously. These provides beneficial guidance data for the further structural optimization and for the design arrangement of primary and secondary circuit. Under the total heat source of 1.23 MW, the coolant mass flow rate of 5.457 kg/s is required to make coolant water corresponding to the Pressurized Water Reactor (PWR) condition (15.5 MPa, 285 °C–325 °C), generating the total coolant pressure drop (△P) of 0.467 MPa. The results show that the present structural design can make all the materials effectively cooled to the allowable temperature range, except for a few small modifications on the both sides of FW. The main components, including the first wall (FW), cooling plates (CPs), side wall (SWs)&stiffening plates (SPs) and the manifold(1–4), dominate 4.7%/41.7%/13%/40.6% of the total pressure drop, respectively. Additionally, the mass flow rate of each channel has been obtained, showing the peak relative deviation of 3.4% and 2% from the average for the paratactic

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

    International Nuclear Information System (INIS)

    Clikeman, F.M.

    1982-07-01

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

  18. Experimental program for the Fast Breeder Blanket Facility, FBBF

    International Nuclear Information System (INIS)

    Ott, K.O.; Clikeman, F.M.; Johnson, R.H.; Borg, R.C.

    1976-01-01

    The work performed in the reporting period was primarily concerned with the development of the experimental program (Task A) and with the pre-analysis of future loadings and the impact upon the permanent loading of the two converter regions, which contain 4.8 percent enriched UO 2 rods. It appears necessary that a neutron poison (B 4 C) be placed in the converter (transformer) regions in order to hold, also for future loadings, the k/sub eff/ of a hypothetically flooded FBBF well below 1. Since it is planned to use the same welded converter regions for all experiments, the required B 4 C loading needs to be determined prior to the first blanket loading. Further the equipment needs have been identified (Task D), the 252 Cf-source has been requested on a loan basis (Task E). First discussions with ANL on blanket experiments have been initiated

  19. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-12-01

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

  20. Study on the thorium-based breeder with molten fluoride salt blanket in the Nuclear Hot Spring - 5420

    International Nuclear Information System (INIS)

    Bing, X.; Yingzhong, L.

    2015-01-01

    Nuclear Hot Spring (NHS) is an innovative reactor type featured by pool-type molten-salt-cooled pebble-bed reactor core with the capability of natural circulation under full power operation. Except for the potential applications in power generation and high temperature process heat, thorium-based breeding is also a promising feature of the NHS. In order to take advantage of both the highly inherent safety and the on-line processing capability of fluid thorium-based fuels, a breeder design of NHS equipped with a blanket of molten salt with thorium fluoride outside the pebble-bed core is proposed in this work. For the purpose of keeping cleanness of the primary loop and blanket loop, both loops are isolated physically from each other, and the rapid on-line extraction of converted 233 Pa and 233 U is employed for the processing of blanket salt. The conversion ratio, defined as the ratio of converted 233 Pa and 233 U to the consumed fissile uranium in seed fuels, is investigated by varying the relevant parameters such as the circulation flux of blanket salt and the discharge burn-up of seed fuels. It is found that breeding can be achieved for the pure 233 U seed scheme with relatively low discharge burn-up and low blanket salt flux. However, the reprocessing for the HTGR fuels with TRISO particles has to be taken into account to ensure the breeding. (authors)

  1. Advanced methods comparisons of reaction rates in the Purdue Fast Breeder Blanket Facility

    International Nuclear Information System (INIS)

    Hill, R.N.; Ott, K.O.

    1988-01-01

    A review of worldwide results revealed that reaction rates in the blanket region are generally underpredicted with the discrepancy increasing with penetration; however, these results vary widely. Experiments in the large uniform Purdue Fast Breeder Blanket Facility (FBBF) blanket yield an accurate quantification of this discrepancy. Using standard production code methods (diffusion theory with 50 group cross sections), a consistent Calculated/Experimental (C/E) drop-off was observed for various reaction rates. A 50% increase in the calculated results at the outer edge of the blanket is necessary for agreement with experiments. The usefulness of refined group constant generation utilizing specialized weighting spectra and transport theory methods in correcting this discrepancy was analyzed. Refined group constants reduce the discrepancy to half that observed using the standard method. The surprising result was that transport methods had no effect on the blanket deviations; thus, transport theory considerations do not constitute or even contribute to an explanation of the blanket discrepancies. The residual blanket C/E drop-off (about half the standard drop-off) using advanced methods must be caused by some approximations which are applied in all current methods. 27 refs., 3 figs., 1 tab

  2. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 2: BOT helium cooled solid breeder blanket. Vol. 1

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Boccaccini, L.V.; Bojarsky, E.; Deckers, H.; Dienst, W.; Doerr, L.; Fischer, U.; Giese, H.; Guenther, E.; Haefner, H.E.; Hofmann, P.; Kappler, F.; Knitter, R.; Kuechle, M.; Moellendorf, U. von; Norajitra, P.; Penzhorn, R.D.; Reimann, G.; Reiser, H.; Schulz, B.; Schumacher, G.; Schwenk-Ferrero, A.; Sordon, G.; Tsukiyama, T.; Wedemeyer, H.; Weimar, P.; Werle, H.; Wiegner, E.; Zimmermann, H.

    1991-10-01

    The BOT (Breeder Outside Tube) Helium Cooled 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 plankets 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 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.) [de

  3. Integration of test modules in the main blanket and vacuum vessel design

    International Nuclear Information System (INIS)

    Nakahira, Masataka; Kurasawa, Toshimasa; Sato, Satoshi; Furuya, Kazuyuki; Togami, Ikuhide; Hashimoto, Toshiyuki; Takatsu, Hideyuki; Kuroda, Toshimasa.

    1995-07-01

    Typical test modules for water-cooled and helium-cooled ceramic breeder blankets have been designed, and their major design parameters are summarized. Among various candidates studied in Japan at present, BOT (Breeder Out of Tube) type of blanket is exemplified here. The integration scheme of the test module into ITER basic machine is also shown. Even with other type of blanket, the integration scheme won't be affected. The composition and space requirement of cooling and tritium recovery systems for the test module have also been studied. (author)

  4. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

    Smith, D.L.; Sze, D.K.

    1986-02-01

    The recent Blanket Comparison and Selection Study (BCSS), which was a comprehensive evaluation of fusion reactor blanket design and the status of blanket technology, serves as an excellent basis for further development of blanket technology. This study provided an evaluation of over 130 blanket concepts for the reference case of electric power producing, DT fueled reactors in both Tokamak and Tandem Mirror (TMR) configurations. Based on a specific set of reactor operating parameters, the current understanding of materials and blanket technology, and a uniform evaluation methodology developed as part of the study, a limited number of concepts were identified that offer the greatest potential for making fusion an attractive energy source

  5. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Grief, Andrew; Merrill, Brad J.; Humrickhouse, Paul; Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon; Poitevin, Yves; Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard

    2016-01-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  6. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Grief, Andrew [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Merrill, Brad J.; Humrickhouse, Paul [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID (United States); Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom)

    2016-11-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

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

    International Nuclear Information System (INIS)

    Fuetterer, M.A.; Raepsaet, X.; Proust, E.; Leger, D.

    1994-01-01

    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

  8. Conceptual design study for a mirror fusion breeder

    International Nuclear Information System (INIS)

    Huang Jinhua; Deng Boquan; Li Guiqing

    1986-01-01

    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

  9. Activation analysis of Chinese ITER helium cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Han Jingru; Chen Yixue; Ma Xubo; Wang Shouhai; Forrest, R.A.

    2009-01-01

    Based on the Chinese ITER helium cooled solid breeder(CH-HCSB) test blanket module (TBM) of the 3 x 6 sub-modules options, the activation characteristics of the TBM were calculated. Three-dimensional neutronic calculations were performed using the Monte-Carlo code MCNP and the nuclear data library FENDL/2. Furthermore, the activation calculations of HCSB-TBM were carried out with the European activation system EASY-2007. At shutdown the total activity is 1.29 x 10 16 Bq, and the total afterheat is 2.46 kW. They are both dominated by the Eurofer steel. The activity and afterheat are both in the safe range of TBM design, and will not have a great impact on the environment. Meanwhile,on basis of the calculated contact dose rate, the activated materials can be re-used following the remote handling recycling options. The activation results demonstrate that the current HCSB-TBM design can satisfy the ITER safety design requirements from the activation point of view. (authors)

  10. Neutronic optimization of a LiAlO2 solid breeder blanket

    International Nuclear Information System (INIS)

    Levin, P.; Ghoniem, N.M.

    1986-02-01

    In this report, a pressurized lobular blanket configuration is neutronically optimized. Among the features of this blanket configuration are the use of beryllium and LiAlO 2 solid breeder pins in a cross-flow configuration in a helium coolant. One-dimensional neutronic optimization calculations are performed to maximize the tritium breeding ratio (TER). The procedure involves spatial allocations of Be, LiAlO 2 , 9-C (ferritic steel), and He; in such a way as to maximize the TBR subject to several material, engineering and geometrical constraints. A TBR of 1.17 is achieved for a relatively thin blanket (approx. = 43 cm depth), and consistency with all imposed constraints

  11. 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)

  12. Heat-pulse flowmeter for a liquid breeder blanket

    International Nuclear Information System (INIS)

    Kondo, Masatoshi; Shibano, Kyohei; Tanaka, Teruya; Muroga, Takeo

    2013-01-01

    Liquid metals Li, Pb-17Li and Sn-20Li are candidate liquid breeders in fusion reactors. The development of a flowmeter that can be applied to high-temperature liquid metals is an important issue. A heat-pulse flowmeter is proposed in the present study. Its basic performance was investigated by means of a loop experiment with Pb-17Li and a numerical simulation. The temperature distribution in flowing Pb-17Li was obtained by local transient heating of the outer surface of a loop tube. The temperature distribution gradually changed and resembled the movement of a hot spot, which had a higher temperature than its surroundings. This hot spot moved along the flow and passed through the tips of the thermocouples. The change in temperature distribution with the movement of the hot spot was monitored by three thermocouples exposed to the Pb-17Li flow. The results of the loop experiments were numerically simulated by assuming a certain flow rate, and the temperature profile obtained in the loop experiment was in agreement with the simulation results. The time taken by the hot spot to pass through the tips of the thermocouples was measured and simulated, and the correlation between this time and the average flow velocity was evaluated. The results indicated the average flow velocity can be obtained using the heat-pulse flowmeter proposed in this study. (author)

  13. Mechanical design of a light water breeder reactor

    International Nuclear Information System (INIS)

    Fauth, W.L. Jr.; Jones, D.S.; Kolsun, G.J.; Erbes, J.G.; Brennan, J.J.; Weissburg, J.A.; Sharbaugh, J.E.

    1976-01-01

    In a light water reactor system using the thorium-232--uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements. 4 claims, 24 drawing figures

  14. Updated conceptual design of helium cooling ceramic blanket for HCCB-DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Suhao [University of Science and Technology of China, Hefei, Anhui (China); Southwestern Institute of Physics, Chengdu, Sichuan (China); Cao, Qixiang; Wu, Xinghua; Wang, Xiaoyu; Zhang, Guoshu [Southwestern Institute of Physics, Chengdu, Sichuan (China); Feng, Kaiming, E-mail: fengkm@swip.ac.cn [Southwestern Institute of Physics, Chengdu, Sichuan (China)

    2016-11-15

    Highlights: • An updated design of Helium Cooled Ceramic breeder Blanket (HCCB) for HCCB-DEMO is proposed in this paper. • The Breeder Unit is transformed to TBM-like sub-modules, with double “banana” shape tritium breeder. Each sub-module is inserted in space formed by Stiffen Grids (SGs). • The performance analysis is performed based on the R&D development of material, fabrication technology and safety assessment in CN ITER TBM program. • Hot spots will be located at the FW bend side. - Abstract: The basic definition of the HCCB-DEMO plant and preliminary blanket designed by Southwestern Institution of Physics was proposed in 2009. The DEMO fusion power is 2550 MW and electric power is 800 MW. Based on development of R&D in breeding blanket, a conceptual design of helium cooled blanket with ceramic breeder in HCCB-DEMO was presented. The main design features of the HCCB-DEMO blanket were: (1) CLF-1 structure materials, Be multiplier and Li{sub 4}SiO{sub 4} breeder; (2) neutronic wall load is 2.3 MW/m{sup 2} and surface heat flux is 0.43 MW/m{sup 2} (2) TBR ≈ 1.15; (3) geometry of breeding units is ITER TBM-like segmentation; (4)Pressure of helium is 8 MPa and inlet/outlet temperature is 300/500 °C. On the basis of these design, some important analytical results are presented in aspects of (i) neutronic behavior of the blanket; (ii) design of 3D structure and thermal-hydraulic lay-out for breeding blanket module; (iii) structural-mechanical behavior of the blanket under pressurization. All of these assessments proved current stucture fulfill the design requirements.

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

    International Nuclear Information System (INIS)

    Kodeli, I.

    2005-01-01

    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 (Li 2 CO 3 powder). The reaction rate measurements include the Li 2 CO 3 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)

  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)

    Malang, S.; Mattas, R.

    1994-06-01

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

  17. ITER breeding blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

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

  18. Preliminary accident analysis of Loss of Off-Site Power and In-Box LOCA for the CFETR helium cooled solid breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Lian, Qiang; Cui, Shijie [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Zhang, Jing; Zhang, Dalin; Su, G.H. [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China)

    2017-05-15

    Highlights: • The CFETR HCSB blanket has been investigated using RELAP5. • Loss of Off-Site Power is investigated. • The parametric analyses during In-Box LOCA are investigated. • The HCSB blanket for CFETR is designed with sufficient decay heat removal capability. - Abstract: As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage.

  19. Design requirement on HYPER blanket fuel assembly

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, B. O.; Nam, C.; Ryu, W. S.; Lee, B. S.; Park, W. S.

    2000-07-01

    This document describes design requirements which are needed for designing the blanket assembly of the HYPER as design guidance. The blanket assembly of the HYPER consists of blanket fuel rods, mounting rail, spacer, upper nozzle with handling socket, bottom nozzle with mounting rail and skeleton structure. The blanket fuel rod consists of top end plug, bottom end plug with key way, blanket fuel slug, and cladding. In the assembly, the rods are in a triangular pitch array. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements for the blanket fuel assembly of the HYPER

  20. A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; Trow, Martin; Dillistone, Michael

    2016-01-01

    'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

  1. Status of the European R and D on beryllium as multiplier material for breeder blankets

    International Nuclear Information System (INIS)

    Moeslang, A.; Boccaccini, L.V.; Rabaglino, E.; Piazza, G.; Cardella, A.; Sannen, L.; Scibetta, M.; Laan, J. van der; Hegeman, J.B.J.W.

    2004-01-01

    Within the international fusion community a variety of breeding blanket concepts are being considered, ranging from more conservative concepts to higher-risk concepts for fusion power reactors. In Europe, the Helium Cooled Pebble Bed (HCPB) blanket is one of the two reference concepts which will also be tested as Test Blanket Module (TBM) in ITER. In addition to the R and D for structural parts of the HCPB blanket, a considerable effort is devoted to the production and qualification of ceramic breeder and neutron multiplier (beryllium or beryllide) pebble beds. Since in the HCPB blanket pebbles made of lithium ceramics are foreseen, a high volume fraction of beryllium as a neutron multiplier to Li-based ceramic of about 4: l is needed. The typical loading conditions for beryllium are, with a neutron wall load of ∼12.5 MWa/m 2 and in ∼5 years lifetime: T min ∼300degC, T max ∼600-900degC, displacement damage ∼80 dpa, peak 4 He production ∼26000 appm and peak 3 H production ∼700 appm at the End-Of-Life. The behaviour of beryllium under irradiation is considered to be a key issue of the HCPB blanket, because of swelling due to helium bubbles and tritium retention. A large R and D programme on beryllium has been implemented in Europe, aimed at characterising and predicting the material behaviour before and under irradiation. An overview on experimental and modelling activities performed during the past 2 years is given with typical results on non-irradiated and irradiated Beryllium materials and pebble beds and the relevance of major results on future beryllium R and D is addressed. (author)

  2. Overview of design and R and D of solid breeder TBM in China

    International Nuclear Information System (INIS)

    Feng, K.M.; Pan, C.H.; Zhang, G.S.; Yuan, T.; Chen, Z.; Zhao, Z.; Liu, H.B.; Li, Z.Q.; Hu, G.; Wang, X.Y.; Ye, X.F.; Luo, D.L.; Wang, H.Y.; Zhou, Z.W.; Gao, C.M.; Chen, Y.J.; Wang, P.H.; Cao, Q.X.; Wang, Q.J.

    2008-01-01

    Testing of breeding blanket module (TBM) is one of ITER's important objectives. China is performing design and technology development of ITER TBMs based on the development strategy of fusion DEMO in China. Solid breeder with helium-cooled test blanket module concept for test in ITER should be the basic option in China. The progress and status of China helium-cooled solid breeder (CH HCSB) TBM since 2004 are introduced briefly. Concept designs of HCSB TBM and ancillary systems, test strategy for their tests in ITER, key R and D issues are summarized in this paper. An international collaboration in R and D, development and testing of TBMs are proposed

  3. Evolution of actinides in ThO2 blanket of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Bachchan, Abhitab; Riyas, A.; Devan, K.; Puthiyavinayagam, P.

    2015-01-01

    The third stage of India's nuclear program focuses on fissile fuel production through Th- 233 U cycle in view of the better abundance and relative merits of thorium. For early introduction of Thorium into the nuclear energy system, several R and D program has started to find the best possible route of thorium utilization. Towards this, efforts were made to assess the feasibility of Th-U cycle in a fast spectrum reactor like Prototype Fast Breeder Reactor (PFBR). The effect on core neutronic parameters and actinide evolution with the replacement of depleted UO 2 in the PFBR blanket SA with thorium oxide has been studied using 3-D diffusion code FARCOB. Study shows that by the introduction of thorium blanket, core excess reactivity is coming down by ∼ 535 pcm and core breeding ratio is slightly lower than conventional oxide blanket. The distribution of region wise power production is slightly changed. Power from radial blanket is reduced from 3% to 2% while the core-1 power is increased from 49 % to 50 %. The estimated 233 U production is 7.6, 11.5 and 14.1 kg/t with 180, 360 and 540 days of irradiation respectively. (author)

  4. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    International Nuclear Information System (INIS)

    Beller, D.E.; Ott, K.O.; Terry, W.K.

    1987-01-01

    A new conceptual design of a fusion reactor blanket simulation facility has been developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), where experiments have resulted in the discovery of substantial deficiencies in neutronics predictions. With this design, discrepancies between calculation and experimental data can be nearly fully attributed to calculation methods because design deficiencies that could affect results are insignificant. The conceptual design of this FBBF analog, the Fusion Reactor Blanket Facility, is presented

  5. Neutronic analyses of design issues affecting the tritium breeding performance in different DEMO blanket concepts

    Energy Technology Data Exchange (ETDEWEB)

    Pereslavtsev, Pavel, E-mail: pavel.pereslavtsev@kit.edu [Karlsruhe Institute for Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, Christian [EUROfusion – Programme Management Unit, Boltzmannstrasse 2, 85748 Garching (Germany); Fischer, Ulrich [Karlsruhe Institute for Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-11-01

    Highlights: • Realistic 3D MCNP model based on the CAD engineering model of DEMO. • Automated procedure for the generation and arrangement of the blanket modules for different DEMO concepts: HCPB, HCLL, WCLL, DCLL. • Several parameters affecting tritium breeding ratio (TBR) were investigated. • A set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts. - Abstract: Neutronic analyses were performed to assess systematically the tritium breeding ratio (TBR) variations in the DEMO for the different blanket concepts HCPB, HCLL, WCLL and DCLL DEMOs due to modifications of the blanket configurations. A dedicated automated procedure was developed to fill the breeding modules in the common generic model in correspondence to the different concepts. The TBR calculations were carried out using the MCNP5 Monte Carlo code. The following parameters affecting the global TBR were investigated: TBR poloidal distribution, radial breeder zone depth, {sup 6}Li enrichment, steel content in the breeder modules, poloidal segmentation of the breeder blanket volume, size of gaps between blankets, thickness of the first wall and of the tungsten armour. Based on the results a set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts with the goal to achieve the required tritium breeding performance in DEMO.

  6. Neutronic analyses of design issues affecting the tritium breeding performance in different DEMO blanket concepts

    International Nuclear Information System (INIS)

    Pereslavtsev, Pavel; Bachmann, Christian; Fischer, Ulrich

    2016-01-01

    Highlights: • Realistic 3D MCNP model based on the CAD engineering model of DEMO. • Automated procedure for the generation and arrangement of the blanket modules for different DEMO concepts: HCPB, HCLL, WCLL, DCLL. • Several parameters affecting tritium breeding ratio (TBR) were investigated. • A set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts. - Abstract: Neutronic analyses were performed to assess systematically the tritium breeding ratio (TBR) variations in the DEMO for the different blanket concepts HCPB, HCLL, WCLL and DCLL DEMOs due to modifications of the blanket configurations. A dedicated automated procedure was developed to fill the breeding modules in the common generic model in correspondence to the different concepts. The TBR calculations were carried out using the MCNP5 Monte Carlo code. The following parameters affecting the global TBR were investigated: TBR poloidal distribution, radial breeder zone depth, "6Li enrichment, steel content in the breeder modules, poloidal segmentation of the breeder blanket volume, size of gaps between blankets, thickness of the first wall and of the tungsten armour. Based on the results a set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts with the goal to achieve the required tritium breeding performance in DEMO.

  7. NET test blanket design and remote maintenance

    International Nuclear Information System (INIS)

    Holloway, C.; Hubert, P.

    1991-01-01

    The NET machine has three horizontal ports reserved for testing tritium breeding blanket designs during the physics phase and possibly five during the technology phase. The design of the ports and test blankets are modular to accept a range of blanket options, provide radiation shielding and allow routine replacement. Radiation levels during replacement or maintenance require that all operations must be carried out remotely. The paper describes the problems overcome in providing a port design which includes attachment to the vacuum vessel with double vacuum seals, an integrated cooled first wall and support guides for the test blanket module. The method selected to remotely replace the test module whilst controlling the spread of contamination is also adressed. The paper concludes that the provisions of a test blanket facility based on the NET machine design is feasible. (orig.)

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

    International Nuclear Information System (INIS)

    1979-11-01

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

  9. Mechanical and thermal design of hybrid blankets

    International Nuclear Information System (INIS)

    Schultz, K.R.

    1978-01-01

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

  10. Design and analysis of ITER shield blanket

    Energy Technology Data Exchange (ETDEWEB)

    Ohmori, Junji; Hatano, Toshihisa; Ezato, Kouichiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-12-01

    This report includes electromagnetic analyses for ITER shielding blanket modules, fabrication methods for the blanket modules and the back plate, the design and the fabrication methods for port limiter have been investigated. Studies on the runaway electron impact for Be armor have been also performed. (J.P.N.)

  11. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Redler, K.; Reis, E.E.; Will, R.; Cheng, E.; Hasan, C.M.; Sharafat, S.

    1993-11-01

    The ARIES-IV Nested Shell Blanket (NSB) Design is an alternate blanket concept of the ARIES-IV low activation helium-cooled reactor design. The reference design has the coolant routed in the poloidal direction and the inlet and outlet plena are located at the top and bottom of the torus. The NSB design has the high velocity coolant routed in the toroidal direction and the plena are located behind the blanket. This is of significance since the selected structural material is SiC-composite. The NSB is designed to have key high performance components with characteristic dimensions of no larger than 2 m. These components can be brazed to form the blanket module. For the diverter design, we eliminated the use of W as the divertor coating material by relying on the successful development of the gaseous divertor concept. The neutronics and thermal-hydraulic performance of both blanket concepts are similar. The selected blanket and divertor configurations can also meet all the projected structural, neutronics and thermal-hydraulics design limits and requirements. With the selected blanket and divertor materials, the design has a level of safety assurance rate of I (LSA-1), which indicates an inherently safe design

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

    International Nuclear Information System (INIS)

    Patterson-Hine, F.A.; Davidson, J.W.; Klein, D.E.; Lee, J.D.

    1985-01-01

    Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs: fluorination only, fluorination plus reductive extraction, and fluorination, plus reductive extraction, plus metal transfer. The effects of processing on blanket performance have been assessed for these three levels of processing and various equilibrium uranium concentrations in the salt. A one-dimensional model of the blanket was used in the neutronics analysis, which incorporated transport calculations with time-dependent isotope generation and depletion calculations. The method of salt processing was found to have little affect on the level of radioactivity, toxicity, or the thermal behavior of the salt during operation of the reactor. The processing rates necessary to maintain the desired uranium concentrations in the suppressed-fission environment were quite low, which permitted only long-lived species to be removed from the salt. The effects of the processing therefore became apparent only after the radioactivity due to the short-lived species diminished. The effect of the additional processing (reductive extraction and metal transfer) could be seen after approximately 1 year of decay, but were not significant at times closer to shutdown. The reduced radioactivity and corresponding heat deposition were thus of no consequence in accident or maintenance situations. Net fissile production in the Be/MS blanket concept at a fusion power level of 3000 MW at 70% capacity ranged from 5100 kg/year to 5170 kg/year for uranium concentrations of 0.11% and 1.0% 233 U in thorium, respectively, with fluorination-only processing. The addition of processing by reductive extraction resulted in 5125 kg/year for the 0.11% 233 U case and 5225 kg/year for the 1.0% 233 U case

  13. First wall and blanket design for the STARFIRE commercial tokamak power reactor

    International Nuclear Information System (INIS)

    Morgan, G.D.; Trachsel, C.A.; Cramer, B.A.; Bowers, D.A.; Smith, D.L.

    1979-01-01

    The first wall and blanket design concepts being evaluated for the STARFIRE commercial tokamak reactor study are presented. The two concepts represent different approaches to the mechanical design of a tritium breeding blanket using the reference materials options. Each concept has a separate ferritic steel first wall cooled by heavy water (D 2 O), and a ferritic steel blanket with solid lithium oxide breeder cooled by helium. A separate helium purge system is used in both concepts to extract tritium. The two concepts are compared and relative advantages and disadvantages for each are discussed

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

    International Nuclear Information System (INIS)

    Merrill, B.J.

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    Tsuru, Daigo; Enoeda, Mikio; Akiba, Masato

    2007-01-01

    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 10 12 Bq, and tritium in purge gas is about 3 x 10 11 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

  16. Blanket design for imploding liner systems

    International Nuclear Information System (INIS)

    Schaffer, M. J.

    1980-01-01

    The blanket design comprises hot, molten, rotating liquid vortex systems suitable for rapidly compressing confined plasmas, in which stratified immiscible liquid layers having successively greater mass densities outwardly of the axis of rotation are provided

  17. Fast Breeder Blanket Facility. Quarterly progress report, July 1, 1978--September 30, 1978

    International Nuclear Information System (INIS)

    Clikeman, F.M.

    1978-09-01

    This quarterly progress report summarizes work done at Purdue University's Fast Breeder Blanket Facility for the Department of Energy during the months July to September 1978. The summary includes reports on the models and methods used to characterize the FBBF facility. Using the reported models and calculational methods and computer codes a new cross section set has been generated, self-shielded for 300 0 K for use in all FBBF calculations using the 2DB computer code. The summary includes reports of the reproducability of foil activation data and measurements of the azmuthal symmetry of the facility. The status of the development of technique for the experimental measurements and preliminary foil activation measurements are also reviewed

  18. Evaluation on the heat removal capacity of the first wall for water cooled breeder blanket of CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng, E-mail: jiangkecheng@ipp.ac.cn; Cheng, Xiaoman; Chen, Lei; Huang, Kai; Ma, Xuebin; Liu, Songlin

    2016-02-15

    Highlights: • Heat removal capacity of the FW is evaluated under BWR, PWR and He coolant inlet conditions. • Heat transfer property of the gas–liquid two phase and the two boiling crises are analyzed. • Heat removal capacity of water is larger than helium coolant. - Abstract: The water cooled ceramic breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). As an important component of the blanket, the FW should satisfy with the thermal requirements in any case. In this paper, three parameters including the heat removal capacity, coolant pressure drop as well as the temperature rise of the FW were investigated under different coolant velocity and heat flux from the plasma. Using the same first wall structure, two main water cooled schemes including Boiling Water Reactor (BWR, 7 MPa pressure and 265 °C temperature inlet) and Pressurized Water Reactor (PWR, 15 MPa pressure and 285 °C temperature inlet) conditions are discussed in the thermal hydraulic calculation. For further research, the thermal hydraulic characteristics of using helium as coolant (8 MPa pressure, 300 °C temperature inlet) are also explored to provide CFETR blanket design with more useful data supports. Without regard to the outlet coolant condition requirements of the blanket, the results indicate that the ultimate heat flux that the FW can resist is 2.2 MW/m{sup 2} at velocity of 5 m/s for BWR, 2.0 MW/m{sup 2} at velocity of 5 m/s for PWR and 0.87 MW/m{sup 2} for helium at velocity 100 m/s under the chosen operation condition. The detrimental departure from nucleate boiling (DNB) crisis would occur at the velocity of 1 m/s under the heat flux of 3 MW/m{sup 2} and dry out crisis appears at the velocity of less than 0.2 m/s with the heat flux of more than 1 MW/m{sup 2} for BWR. The further blanket/FW optimization design is provided with more useful data references according to the abundant calculation results.

  19. Progress in blanket designs using SiCf/SiC composites

    International Nuclear Information System (INIS)

    Giancarli, L.; Golfier, H.; Nishio, S.; Raffray, R.; Wong, C.; Yamada, R.

    2002-01-01

    This paper summarizes the most recent design activities concerning the use of SiC f /SiC composite as structural material for fusion power reactor breeding blanket. Several studies have been performed in the past. The most recent proposals are the TAURO blanket concept in the European Union, the ARIES-AT concept in the US, and DREAM concept in Japan. The first two concepts are self-cooled lithium-lead blankets, while DREAM is an helium-cooled beryllium/ceramic blanket. Both TAURO and ARIES-AT blankets are essentially formed by a SiC f /SiC box acting as a container for the lithium-lead which has the simultaneous functions of coolant, tritium breeder, neutron multiplier and, finally, tritium carrier. The DREAM blanket is characterized by small modules using pebble beds of Be as neutron multiplier material, of Li 2 O (or other lithium ceramics) as breeder material and of SiC as shielding material. The He coolant path includes a flow through the pebble beds and a porous partition wall. For each blanket, this paper describes the main design features and performances, the most recent design improvements, and the proposed manufacturing routes in order to identify specific issues and requirements for the future R and D on SiC f /SiC

  20. Review: BNL Tokamak graphite blanket design concepts

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.

    1976-01-01

    The BNL minimum activity graphite blanket designs are reviewed, and three are discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a 30 cm or thicker graphite screen. Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy, which is then radiated to a secondary blanket with coolant tubes, as in types A and B, or removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma. (Auth.)

  1. New concepts for controlled fusion reactor blanket design

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Avci, H.; El-Maghrabi, M.

    1975-01-01

    Several new concepts for fusion reactor blanket design based on the idea of shifting, or tailoring, the neutron spectrum incident on the first structural wall are presented. The spectral shifter is a nonstructural element which can be made of graphite, silicon carbide, or three dimensionally woven carbon fibers (and containing other materials as appropriate) placed between the neutron source and the first structural wall. The softened neutron spectrum incident on the structural components leads to lower gas production and atom displacement rates than in more standard fusion blanket designs. In turn, this results in longer anticipated lifetimes for the structural materials and can significantly reduce radioactivity and afterheat levels. In addition, the neutron spectrum in the first structural wall can be made to approach the flux shape in fast breeder reactors. Such spectral softening means that existing radiation facilities may be more profitably used to provide relevant materials radiation damage data for the structural materials in these fusion blanket designs. This general class of blanket concepts are referred to as internal spectral shifter and energy converter, or ISSEC concepts. These specific design concepts fall into three main categories: ISSEC/EB concepts based on utilizing existing designs which breed tritium behind the first structural wall; ISSEC/IB concepts based on breeding tritium inside the first vacuum wall; and ISSEC/Bu concepts based on using boron, carbon, and perhaps, beryllium to obtain an energy multiplier and converter design that does not attempt to breed tritium or utilize lithium. The detailed analyses relate specifically to the nuclear performance of ISSEC systems and to a discussion of materials radiation damage problems in the structural material.(U.S.)

  2. Advanced liquid metal fast breeder reactor designs

    International Nuclear Information System (INIS)

    Sayles, C.W.

    1978-01-01

    Fast Breeder reactor power plants in the 1000-1200 MW(e) range are being built overseas and are being designed in this country. While these reactors have many characteristics in common, a variety of different approaches have been adopted for some of the major features. Some of those alternatives are discussed

  3. Tritium system design studies of fusion experimental breeder

    International Nuclear Information System (INIS)

    Deng Baiquan; Huang Jinhua

    2003-01-01

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

  4. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Sato, Satoshi; Hatano, Toshihisa; Tokami, Ikuhide; Kitamura, Kazunori; Miura, Hidenori; Ito, Yutaka; Kuroda, Toshimasa; Takatsu, Hideyuki

    1997-05-01

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

  5. First results of the post-irradiation examination of the Ceramic Breeder materials from the Pebble Bed Assemblies Irradiation for the HCPB Blanket concept

    International Nuclear Information System (INIS)

    Hegeman, J.; Magielsen, A.J.; Peeters, M.; Stijkel, M.P.; Fokkens, J.H.; Laan, J.G. van der

    2006-01-01

    In the framework of developing the European Helium Cooled Pebble-Bed (HCPB) blanket an irradiation test of pebble-bed assemblies is performed in the HFR Petten. The experiment is focused on the thermo-mechanical behavior of the HCPB type breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. To achieve representative conditions a section of the HCPB is simulated by EUROFER-97 cylinders with a horizontal bed of ceramic breeder pebbles sandwiched between two beryllium beds. Floating Eurofer-97 steel plates separate the pebble-beds. The structural integrity of the ceramic breeder materials is an issue for the design of the Helium Cooled Pebble Bed concept. Therefore the objective of the post irradiation examination is to study deformation of pebbles and the pebble beds and to investigate the microstructure of the ceramic pebbles from the Pebble Bed Assemblies. This paper concentrates on the Post Irradiation Examination (PIE) of the four ceramic pebble beds that have been irradiated in the Pebble Bed Assembly experiment for the HCPB blanket concept. Two assemblies with Li 4 SiO 4 pebble-beds are operated at different maximum temperatures of approximately 600 o C and 800 o C. Post irradiation computational analysis has shown that both have different creep deformation. Two other assemblies have been loaded with a ceramic breeder bed of two types of Li 2 TiO 3 beds having different sintering temperatures and consequently different creep behavior. The irradiation maximum temperature of the Li 2 TiO 3 was 800 o C. To support the first PIE result, the post irradiation thermal analysis will be discussed because thermal gradients have influence on the pebble-bed thermo-mechanical behavior and as a result it may have impact on the structural integrity of the ceramic breeder materials. (author)

  6. Key achievements in elementary R&D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

    Science.gov (United States)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Hayashi, K.; Tanigawa, H.; Ochiai, K.; Nishitani, T.; Tobita, K.; Akiba, M.

    2006-02-01

    This paper presents the significant progress made in the research and development (R&D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li2TiO3 pebble bed under neutron-pulsed operation conditions are summarized. With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 °C followed by normalizing it at 930 °C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R&D on the breeder material, Li2TiO3, the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li2TiO3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li2TiO3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation.

  7. Key achievements in elementary R and D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

    International Nuclear Information System (INIS)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Hayashi, K.; Tanigawa, H.; Ochiai, K.; Nishitani, T.; Tobita, K.; Akiba, M.

    2006-01-01

    This paper presents the significant progress made in the research and development (R and D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li 2 TiO 3 pebble bed under neutron-pulsed operation conditions are summarized. With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 0 C followed by normalizing it at 930 0 C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R and D on the breeder material, Li 2 TiO 3 , the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li 2 TiO 3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li 2 TiO 3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation

  8. Key achievements in elementary R and Ds on water-cooled solid breeder blanket for ITER Test Blanket Module in JAERI

    International Nuclear Information System (INIS)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Tanigawa, H.; Tobita, K.; Akiba, M.; Hayashi, K.; Ochiai, K.; Nishitani, T.

    2005-01-01

    This paper presents significant progress in research and development (R and D) of key elementary technologies on the water-cooled solid breeder blanket for the ITER test blanket modules (TBMs) in JAERI. Development of module fabrication technology, bonding technology of armors, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup, and tritium release behavior from Li 2 TiO 3 pebble bed under neutron pulsed operation condition are summarized. By the improvement of heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H, can be obtained by homogenizing it at 1150 deg C followed by normalizing at 930 deg C after the Hot Isostatic Pressing (HIP) process. Moreover, a promising bonding process for a tungsten armor and an F82H structural material was developed by using a solid state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it was found that the thermal fatigue lifetime of F82H can be predicted by using Manson-Coffin's law. As for R and Ds on a breeder material, Li 2 TiO 3 , effective thermal conductivity of Li 2 TiO 3 pebble was measured under compressive force simulating the ITER TBM environment. The increase in the effective thermal conductivity of the pebble bed was about 2.5 % at the compressive strain of 0.9 % at 400 deg C. Neutronic performance of the blanket module mockup has been carried out by the 14 MeV neutron irradiation. It was confirmed that the measured tritium production rate agreed with the calculated values within about 10% difference. Also, tritium release from a Li 2 TiO 3 pebble bed was measured under pulsed neutron irradiation conditions simulating the ITER operation. (author)

  9. Detailed mechanical design and manufacturing study for the ITER reference breeding blanket

    International Nuclear Information System (INIS)

    Zacchia, F.; Daenner, W.; Stefanis, L. de; Ferrari, M.; Gerber, A.; Mustoe, J.

    1998-01-01

    This papers relates on the detailed mechanical design, manufacturing feasibility and assembly analysis of a water-cooled solid breeding blanket concept, selected as the ITER reference design. This breeding blanket design is characterised by: i) pressurised water flowing inside flat steel panels for cooling of the internals; each panel is welded along its contour onto the first wall structure and to the rear shield plate after closure of the module (last assembly step). ii) Beryllium (neutronic multiplier) in the form of micro-spheres filling the volume between parallel flat coolant panels. iii) Breeder pebbles enclosed in rods, which form bundles and are themselves embedded inside the Beryllium micro-spheres. (authors)

  10. Neutronic performance of two European breeder-inside-tube (BIT) blankets for DEMO: the helium-cooled ceramic LiAlO2 with Be multiplier and the water-cooled liquid Li17Pb

    International Nuclear Information System (INIS)

    Petrizzi, L.; Rado, V.

    1995-01-01

    In support of ENEA activity in the European Community Test Programme, neutron analysis has been performed on the two latest blanket designs: helium-cooled ceramic breeder-inside-tube (BIT) (with LiAlO 2 and Be multiplier) and water-cooled liquid Li 17 Pb in cylindrical modules (CM). The powerful MCNP Monte Carlo code was used (version 4.2). A detailed and accurate description of the geometrical model has been performed by inserting the main reactor details and avoiding breeder material dilution inside the modules. The tritium breeding ratio (TBR) performance is low for the solid breeder BIT blanket (with 10 ports 1.011) due mainly to low blanket coverage near the exhaust duct, and this solution should be revised. The CM Li 17 Pb blanket reaches a sufficient TBR (1.059, with ports) to rely on tritium self-sufficiency. Shielding properties, with respect to the toroidal field coils, have been estimated in a simplified model by means of the ANISN code, supplied with a nuclear data library consistent with that used by MCNP. The analysis suggests that a careful shield thickness/composition design should be used to ensure the shielding capability of the whole blanket plus shield system. (orig.)

  11. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

    Simbolotti, G.; Zampaglione, V.; Ferrari, M.; Gallina, M.; Mazzone, G.; Nardi, C.; Petrizzi, L.; Rado, V.; Violante, V.; Daenner, W.; Lorenzetto, P.; Gierszewski, P.; Grattarola, M.; Rosatelli, F.; Secolo, F.; Zacchia, F.; Caira, M.; Sorabella, L.

    1993-01-01

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

  12. Neutronic analyses of the preliminary design of a DCLL blanket for the EUROfusion DEMO power plant

    Energy Technology Data Exchange (ETDEWEB)

    Palermo, Iole, E-mail: iole.palermo@ciemat.es; Fernández, Iván; Rapisarda, David; Ibarra, Angel

    2016-11-01

    Highlights: • We perform neutronic calculations for the preliminary DCLL Blanket design. • We study the tritium breeding capability of the reactor. • We determine the nuclear heating in the main components. • We verify if the shielding of the TF coil is maintained. - Abstract: In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding.

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

    International Nuclear Information System (INIS)

    Cizelj, L.; Riesch-Oppermann, H.; Kernforschungszentrum Karlsruhe GmbH

    1994-10-01

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

  14. Design activities of a fusion experimental breeder

    International Nuclear Information System (INIS)

    Huang, J.; Feng, K.; Sheng, G.

    1999-01-01

    The fusion reactor design studies in China are under the support of a fusion-fission hybrid reactor research Program. The purpose of this program is to explore the potential near-term application of fusion energy to support the long-term fusion energy on the one hand and the fission energy development on the other. During 1992-1996 a detailed consistent and integral conceptual design of a Fusion Experimental Breeder, FEB was completed. Beginning from 1996, a further design study towards an Engineering Outline Design of the FEB, FEB-E, has started. The design activities are briefly given. (author)

  15. Design activities of a fusion experimental breeder

    International Nuclear Information System (INIS)

    Huang, J.; Feng, K.; Sheng, G.

    2001-01-01

    The fusion reactor design studies in China are under the support of a fusion-fission hybrid reactor research Program. The purpose of this program is to explore the potential near-term application of fusion energy to support the long-term fusion energy on the one hand and the fission energy development on the other. During 1992-1996 a detailed consistent and integral conceptual design of a Fusion Experimental Breeder, FEB was completed. Beginning from 1996, a further design study towards an Engineering Outline Design of the FEB, FEB-E, has started. The design activities are briefly given. (author)

  16. Experimental programme in support of the development of the European ceramic-breeder-inside-tube test-blanket: present status and future work

    International Nuclear Information System (INIS)

    Proust, E.; Roux, N.; Flament, T.; Anzidei, L.; ENEA, Frascati; Casadio, S.; Dell'orco, G.

    1992-01-01

    Four DEMO blanket classes are under investigation within the framework of the European Test-Blanket Development Programme. One of them is featured by the use of lithium ceramic breeder pellets contained inside externally helium cooled tubes. This paper summarizes the main achievements to date of the experimental programme supporting the development of this class of blanket. It also gives an outline of the areas of the breeder material, beryllium, tritium control, and thermomechanical tests, the future work envisaged for the 92-94 period. 53 refs

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

    International Nuclear Information System (INIS)

    Tsoulfanidis, N.; Jankhah, M.H.

    1979-01-01

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

  18. Conceptual design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Sato, Satoshi; Takatsu, Hideyuki; Kurasawa, Toshimasa

    1995-03-01

    The present report summarizes the design activities of the ITER first wall and shielding blanket conducted by the JA Home Team during this year (1994) in close contact with the JCT, and reported during the four Technical Meetings held at Garching ITER Co-center. These activities are based on the Task Agreement between the JCT and the JA Home Team. In the present report, a layered configuration composed of separate first walls, modular-type blanket modules and separate back plates has been proposed to realize reliable assembly and maintenance schemes as well as to realize reliable component designs under high surface heat loads, high neutron wall loading and electromagnetic loads during disruptions. Outline of the structural design, consideration on fabricability and maintainability, and the results of thermal, mechanical and electromagnetic analyses are described. (author)

  19. Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148, Yuseong-gu, Daejeon 305-806 (Korea, Republic of)

    2016-02-15

    Highlights: • A methodology to simulate the K-DEMO blanket system was proposed. • The results were compared with the CFD, to verify the prediction capability of MARS. • 46 Blankets in a single sector in K-DEMO were simulated using MARS-KS. • Supervisor program was devised to handle each blanket module individually. • The calculation results showed the flow rates, pressure drops, and temperatures. - Abstract: According to the conceptual design of the fusion DEMO reactor proposed by the National Fusion Research Institute of Korea, the water-cooled breeding blanket system incorporates a total of 736 blanket modules. The heat flux and neutron wall loading to each blanket module vary along their poloidal direction, and hence, thermal analysis for at least one blanket sector is required to confirm that the temperature limitations of the materials are satisfied in all the blanket modules. The present paper proposes a methodology of thermal analysis for multiple modules of the blanket system using a nuclear reactor thermal-hydraulic analysis code, MARS-KS. In order to overcome the limitations of the code, caused by the restriction on the number of computational nodes, a supervisor program was devised, which handles each blanket module separately at first, and then corrects the flow rate, considering pressure drops that occur in each module. For a feasibility test of the proposed methodology, 46 blankets in a single sector were simulated; the calculation results of the parameters, such as mass flow, pressure drops, and temperature distribution in the multiple blanket modules showed that the multi-module analysis method can be used for efficient thermal-hydraulic analysis of the fusion DEMO reactor.

  20. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-01-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-I through IV. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies. ARIES-1 investigated the use of SiC composite as the structural material to increase the blanket temperature and reduce the blanket activation. Li 2 ZrO 3 was used as the breeding material due to its high temperature stability and good tritium recovery characteristics. The ARIES-IV is a modification of ARIES-1. The plasma was in the second stability regime. Li 2 O was used as the breeding material to remove Zr. A gaseous divertor was used to replace the conventional divertor so that high Z divertor target is not required. The physics of ARIES-II was the same as ARIES-IV. The engineering design of the ARIES-II was based on a self-cooled lithium blanket with a V-alloy as the structural material. Even though it was assumed that the plasma was in the second stability regime, the plasma beta was still rather low (3.4%). The ARIES-III is an advanced fuel (D- 3 He) tokamak reactor. The reactor design assumed major advancement on the physics, with a plasma beta of 23.9%. A conventional structural material is acceptable due to the low neutron wall loading. From the radiation damage point of view, the first wall can last the life of the reactor, which is expected to be a major advantage from the engineering design and waste disposal point of view

  1. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m 2 and a particle heat flux of 1 MW/m 2 . Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma

  2. Tokamak blanket design study, final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m/sup 2/ and a particle heat flux of 1 MW/m/sup 2/. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma.

  3. A coupled systems code-CFD MHD solver for fusion blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Wolfendale, Michael J., E-mail: m.wolfendale11@imperial.ac.uk; Bluck, Michael J.

    2015-10-15

    Highlights: • A coupled systems code-CFD MHD solver for fusion blanket applications is proposed. • Development of a thermal hydraulic systems code with MHD capabilities is detailed. • A code coupling methodology based on the use of TCP socket communications is detailed. • Validation cases are briefly discussed for the systems code and coupled solver. - Abstract: The network of flow channels in a fusion blanket can be modelled using a 1D thermal hydraulic systems code. For more complex components such as junctions and manifolds, the simplifications employed in such codes can become invalid, requiring more detailed analyses. For magnetic confinement reactor blanket designs using a conducting fluid as coolant/breeder, the difficulties in flow modelling are particularly severe due to MHD effects. Blanket analysis is an ideal candidate for the application of a code coupling methodology, with a thermal hydraulic systems code modelling portions of the blanket amenable to 1D analysis, and CFD providing detail where necessary. A systems code, MHD-SYS, has been developed and validated against existing analyses. The code shows good agreement in the prediction of MHD pressure loss and the temperature profile in the fluid and wall regions of the blanket breeding zone. MHD-SYS has been coupled to an MHD solver developed in OpenFOAM and the coupled solver validated for test geometries in preparation for modelling blanket systems.

  4. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-06-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-1 through 4 and PULSAR 1 and 2. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. Also, the requirements of engineering and physics systems for a pulsed reactor were evaluated by the PULSAR design studies. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies

  5. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    International Nuclear Information System (INIS)

    Beller, D.E.

    1986-01-01

    A new conceptual design of a fusion reactor blanket simulation facility was developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBR), because experiments conducted in it have resulted in the discovery of deficiencies in neutronics prediction methods. With this design, discrepancies between calculation and experimental data can be fully attributed to calculation methods because design deficiencies that could affect results are insignificant. Inelastic scattering cross sections are identified as a major source of these discrepancies. The conceptual design of this FBBR analog, the fusion reactor blanket facility (FRBF), is presented. Essential features are a cylindrical geometry and a distributed, cosine-shaped line source of 14-MeV neutrons. This source can be created by sweeping a deuteron beam over an elongated titanium-tritide target. To demonstrate that the design of the FRBF will not contribute significant deviations in experimental results, neutronics analyses were performed: results of comparisons of 2-dimensional to 1-dimensional predictions are reported for two blanket compositions. Expected deviations from 1-D predictions which are due to source anisotropy and blanket asymmetry are minimal. Thus, design of the FRBF allows simple and straightforward interpretation of the experimental results, without a need for coarse 3-D calculations

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

    International Nuclear Information System (INIS)

    Chapin, D.L.; Green, L.; Lee, A.Y.; Culbert, M.E.; Kelly, J.L.

    1979-09-01

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

  7. Conceptual design of Blanket Remote Handling System for CFETR

    International Nuclear Information System (INIS)

    Wei, Jianghua; Song, Yuntao; Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong

    2015-01-01

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  8. Conceptual design of Blanket Remote Handling System for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Jianghua, E-mail: weijh@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2015-11-15

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  9. Neutronic analysis of graphite-moderated solid breeder design for INTOR

    International Nuclear Information System (INIS)

    Jung, J.; Abdou, M.A.

    1981-01-01

    An in-depth analysis of the INTOR tritium-production-blanket design is presented. A ternary system of solid silicate breeder, lead neutron multiplier, and graphite moderator is explored primary from safety and blanket tritium-inventory considerations. Lithium-silicate (Li 2 SiO 3 ) breeder systems are studied along with water (H 2 O/D 2 O) and Type 316 stainless steel as coolant and structural material, respectively. The analysis examines the neutronics effects on tritium-production regarding: (1) coolant choice; (2) moderator choice; (3) moderator location; (4) multiplier thickness; (5) 6 Li enrichment; and (6) 6 Li burnup. The tritium-breeding-blanket modules are located at the top, outboard, and bottom (outer) parts of the torus, resulting in a breeding coverage of approx. 60% at the first-wall surface. It is found that the reference INTOR design yields, based on a three-dimensional analysis, a net tritium breeding ratio (BR) of approx. 0.65 at the beginning of reactor operation, satisfying the design criterion of BR > 0.6

  10. Fusion breeder

    International Nuclear Information System (INIS)

    Moir, R.W.

    1982-01-01

    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

  11. Safety and environmental impact of the BOT helium cooled solid breeder blanket for DEMO. SEAL subtask 6.2, final report

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Dammel, F.; Gabel, K.

    1996-03-01

    The European Union has been engaged since 1989 in a programme to develop tritium breeding blankets for application in a fusion power reactor. There are four concepts under development, namely two of the solid breeder type and two of the liquid breeder type. At the Forschungszentrum Karlsruhe one blanket concept of each line has been pursued so far with the so-called breeder outside tube (BOT) type representing the solid breeder line. In the BOT concept, Li 4 SiO 4 is used as ceramic breeding material in the form of pebble beds in combination with beryllium pebbles serving as neutron multiplier. Breeder and multiplier materials are arranged in radial-toroidal layers, separated by cooling plates. The coolant is high pressure helium which is circulated in series, at first through the first wall structure and subsequently through the cooling plates. The safety and environmental impact of the BOT blanket concept has been assessed as part of the blanket concept selection exercise, a European concerted action aiming at selecting the two most promising concepts for further development. The topics investigated are: (a) Blanket materials and toxic materials inventory, (b) energy sources for mobilisation, (c) fault tolerance, (d) tritium and activation product release, and (e) waste generation. No insurmountable safety problems have been identified for the BOT concept. The results of the assessment are described in this report. The information collected is also intended to serve as input to the EU 'Safety and Environmental Assessment of Fusion long-term Programme' (SEAL). The unresolved issues pertaining to the BOT blanket which need further investigations in future programmes are outlined herein. (orig.) [de

  12. Ceramic breeder materials

    International Nuclear Information System (INIS)

    Johnson, C.E.

    1990-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear 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. 32 refs., 1 fig., 1 tab

  13. Overview of R and D at TLK for process and analytical issues on tritium management in breeder blankets of ITER and DEMO

    International Nuclear Information System (INIS)

    Demange, D.; Alecu, C.G.; Bekris, N.; Borisevich, O.; Bornschein, B.; Fischer, S.; Gramlich, N.; Köllö, Z.; Le, T.L.; Michling, R.; Priester, F.; Röllig, M.; Schlösser, M.; Stämmler, S.; Sturm, M.; Wagner, R.; Welte, S.

    2012-01-01

    Highlights: ► We present advanced processes and analytics to improve tritium management. ► Membranes and membrane reactors can minimise tritium residence time and inventory. ► Spectroscopic methods can ensure on-line and near to real time tritium measurement. - Abstract: Safe, reliable, and efficient tritium management in the breeder blanket will have to face unprecedented technological challenges. Beside the efficiency for tritium recovery from the breeder blanket (Tritium Extraction (TES) and Coolant Purification Systems (CPS)), the accuracy for tritium tracking between the inner and the outer fuel cycle must also be demonstrated. This paper focuses on the recent R and D carried out at the Tritium Laboratory Karlsruhe to tackle these issues. For ITER, the recently consolidated TES and CPS designs comprise adsorption columns and getter beds operated in semi-continuous mode. Different approaches for the tritium accountancy stage (TAS) have been evaluated. Balancing static (batch-wise gas collection at the TBM outlets and the tritium plant) or dynamic (in/on-line) approaches with respect to the expected analytical performances and integration issues, the first conceptual design of the TAS for EU TBMs is presented. For DEMO, the overall strategy for tritium recovery and tracking has been revisited. The necessity for on-line real-time tritium accountancy and improved process efficiency suggest the use of continuous processes such as permeator and catalytic membrane reactor. The main benefits combining the PERMCAT process with advanced membranes is discussed with respect to process improvements and facilitated accountancy using spectroscopic methods.

  14. Fast breeder physics and nuclear core design

    International Nuclear Information System (INIS)

    Marth, W.; Schroeder, R.

    1983-07-01

    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.) [de

  15. Using one hybrid 3D-1D-3D approach for the conceptual design of WCCB blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng; Zhang, Xiaokang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Li, Jia [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Ma, Xuebin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China)

    2017-01-15

    Highlights: • The Hybrid 3D-1D-3D approach is used for radial building design of WCCB. • Nuclear heat obtained by this method agrees well with 3D neutronics results. • The final results of temperature and TBR satisfy with the requirements. • All the results show that this approach is high efficiency and high reliability. - Abstract: A hybrid 3D-1D-3D approach is proposed for the conceptual design of a blanket. Firstly, the neutron wall loading (NWL) of each blanket module is obtained through a neutronics calculation employing a 3D model, which contains the geometry outline of in-vacuum vessel components and the exact neutron source distribution. Secondly, a 1D cylindrical model with the blanket module containing a detailed radial building is adopted for the neutronics analysis, with the aim of calculating the tritium breeding ratio (TBR) and nuclear heating. Being normalized to the NWL, the nuclear heating is transferred to a 2D model for thermal-hydraulics analysis using the FLUENT code. Through a series analysis of nuclear-thermal iterations that considers the tritium breeding ratio (TBR) and thermal performance as optimization objectives, the optimized radial building of each module surrounding plasma can be obtained. Thirdly, the 3D structural design of each module is established by adding side walls, cover plates, stiffening plates, and other components based on the radial building. The 3D neutronics and thermal-hydraulics using the detailed blanket modules are re-analyzed. This approach has been successfully applied to the design of a water-cooled ceramic breeder blanket for the Chinese Fusion Engineering Test Reactor (CFETR). The radial building of each blanket module surrounding plasma is optimized. The global tritium breeding ratio (TBR) calculated by the 3D neutronics analysis is 1.21, and the temperature of all materials in the 3D blanket structure is below the upper limits. As indicated by the comparison of the 1D and 3D neutronics and thermal

  16. Blanket materials for DT fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1981-01-01

    This paper presents an overview of the critical materials issues that must be considered in the development of a tritium breeding blanket for a tokamak fusion reactor that operates on the D-T-Li fuel cycle. The primary requirements of the blanket system are identified and the important criteria that must be considered in the development of blanket technology are summarized. The candidate materials are listed for the different blanket components, e.g., breeder, coolant, structure and neutron multiplier. Three blanket concepts that appear to offer the most potential are: (1) liquid-metal breeder/coolant, (2) liquid-metal breeder/separate coolant, and (3) solid breeder/separate coolant. The major uncertainties associated with each of the design concepts are discussed and the key materials R and D requirements for each concept are identified

  17. Surface condition effects on tritium permeation through the first wall of a water-cooled ceramic breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, H.-S. [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei (China); Xu, Y.-P.; Liu, H.-D. [Science Island Branch of Graduate School, University of Science and Technology of China, P.O. Box 1126, Hefei (China); Liu, F.; Li, X.-C.; Zhao, M.-Z.; Qi, Q.; Ding, F. [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei (China); Luo, G.-N., E-mail: gnluo@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei (China); Science Island Branch of Graduate School, University of Science and Technology of China, P.O. Box 1126, Hefei (China); Hefei Center for Physical Science and Technology, P.O. Box 1126, Hefei (China); Hefei Science Center of Chinese Academy of Science, P.O. Box 1126, Hefei (China)

    2016-11-01

    Highlights: • We investigate surface effects on T transport through the first wall. • We solve transport equations with various surface conditions. • The RAFMs walls w/and w/o W exhibit different T permeation behavior. • Diffusion in W has been found to be the rate-limiting step. - Abstract: Plasma-driven permeation of tritium (T) through the first wall of a water-cooled ceramic breeder (WCCB) blanket may raise safety and other issues. In the present work, surface effects on T transport through the first wall of a WCCB blanket have been investigated by theoretical calculation. Two types of wall structures, i.e., reduced activation ferritic/martensitic steels (RAFMs) walls with and without tungsten (W) armor, have been analyzed. Surface recombination is assumed to be the boundary condition for both the plasma-facing side and the coolant side. It has been found that surface conditions at both sides can affect T permeation flux and inventory. For the first wall using W as armor material, T permeation is not sensitive to the plasma-facing surface conditions. Contamination of the surfaces will lead to higher T inventory inside the first wall.

  18. A new combination of membranes and membrane reactors for improved tritium management in breeder blanket of fusion machines

    International Nuclear Information System (INIS)

    Demange, D.; Staemmler, S.; Kind, M.

    2011-01-01

    Tritium used as fuel in future fusion machines will be produced within the breeder blanket. The tritium extraction system recovers the tritium to be routed into the inner-fuel cycle of the machine. Accurate and precise tritium accountancy between both systems is mandatory to ensure a reliable operation. Handling in the blanket huge helium flow rates containing tritium as traces in molecular and oxide forms is challenging both for the process and the accountancy. Alternative tritium processes based on combinations of membranes and membrane reactors are proposed to facilitate the tritium management. The PERMCAT process is based on counter-current isotope swamping in a palladium membrane reactor. It allows recovering tritium efficiently from any chemical species. It produces a pure hydrogen stream enriched in tritium of advantage for integration upstream of the accountancy stage. A pre-separation and pre-concentration stage using new zeolite membranes has been studied to optimize the whole process. Such a combination could improve the tritium processes and facilitate accountancy in DEMO.

  19. Neutronic design for the TFTR lithium blanket module

    International Nuclear Information System (INIS)

    Cheng, E.T.; Engholm, B.A.; Su, S.D.

    1981-01-01

    The preliminary design of a lithium blanket module (LBM) to be installed and tested in the TFTR has been performed under subcontract to PPPL and EPRI. The objectives of the LBM program are calculation and measurement of neutron fluences and tritium production in a breeding blanket module using state of art techniques, comparison of calculations with measurements, and acquisition of operational experience with a fusion reactor blanket module. The neutronic design of the LBM is one of the key areas of this program in which the LBM composition and geometry are optimized and the boundary material effects on the tritium production in the blanket module are explored. The concept of employing sintered Li/sub 2/O pellets in tubes is proposed for the blanket design

  20. Numerical studies on the heat transfer and friction characteristics of the first wall inserted with the screw blade for water cooled ceramic breeder blanket of CFETR

    International Nuclear Information System (INIS)

    Jiang, Kecheng; Ma, Xuebin; Cheng, Xiaoman; Liu, Songlin

    2016-01-01

    Highlights: • Enhanced heat transfer and friction characteristics of the FW inserted with screw blade is investigated. • The screw blade structure optimization was done on the screw pitch and diameter. • Decreasing screw pitch and increasing screw diameter could further enhance heat transfer accompanied with increasing flow resistance. • Evaluate the overall enhanced heat performance by using the PEC value. - Abstract: The Water Cooled Ceramic Breeder (WCCB) blanket based on Pressurized Water Reactor (PWR) condition is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). The first wall (FW) which plays an important part in the blanket design must remove the high heat flux radiated from plasma and nuclear heat deposition on the structure in any operating conditions. In this paper, the characteristics of enhanced heat transfer and friction for the FW with the inserted screw blade are studied by the numerical method. After the comparison between the numerical and experimental results, the standard k–ε turbulent model is selected to do the numerical calculation. The numerical results show that the peak temperature of RAFM steel could be reduced by decreasing screw pitch or increasing screw diameter, while accompanied with ascending flow resistance. Besides, among all of the chosen calculation cases compared with the smooth channel, the maximum value of temperature reduction is 10 °C under the conditions of heat flux of 0.5 MW/m"2 as well as screw pitch of 18 mm and screw diameter of 6 mm. The maximum increment ratio of the friction factor is 257% under the conditions of screw pitch of 10 mm and screw diameter of 4 mm. Furthermore, screw blade of 74 mm pitch and 4 mm diameter presents the highest overall performance evaluation criterion (PEC) value of 0.93 under Reynolds number of 270 000 conditions, and shows the best overall heat transfer enhancement performance.

  1. Numerical studies on the heat transfer and friction characteristics of the first wall inserted with the screw blade for water cooled ceramic breeder blanket of CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); University of Science and Technology of China, Hefei, Anhui 230037 (China); Ma, Xuebin; Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2016-03-15

    Highlights: • Enhanced heat transfer and friction characteristics of the FW inserted with screw blade is investigated. • The screw blade structure optimization was done on the screw pitch and diameter. • Decreasing screw pitch and increasing screw diameter could further enhance heat transfer accompanied with increasing flow resistance. • Evaluate the overall enhanced heat performance by using the PEC value. - Abstract: The Water Cooled Ceramic Breeder (WCCB) blanket based on Pressurized Water Reactor (PWR) condition is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). The first wall (FW) which plays an important part in the blanket design must remove the high heat flux radiated from plasma and nuclear heat deposition on the structure in any operating conditions. In this paper, the characteristics of enhanced heat transfer and friction for the FW with the inserted screw blade are studied by the numerical method. After the comparison between the numerical and experimental results, the standard k–ε turbulent model is selected to do the numerical calculation. The numerical results show that the peak temperature of RAFM steel could be reduced by decreasing screw pitch or increasing screw diameter, while accompanied with ascending flow resistance. Besides, among all of the chosen calculation cases compared with the smooth channel, the maximum value of temperature reduction is 10 °C under the conditions of heat flux of 0.5 MW/m{sup 2} as well as screw pitch of 18 mm and screw diameter of 6 mm. The maximum increment ratio of the friction factor is 257% under the conditions of screw pitch of 10 mm and screw diameter of 4 mm. Furthermore, screw blade of 74 mm pitch and 4 mm diameter presents the highest overall performance evaluation criterion (PEC) value of 0.93 under Reynolds number of 270 000 conditions, and shows the best overall heat transfer enhancement performance.

  2. Nuclear performance optimization of the Be/Li/Th blanket for the fusion breeder

    International Nuclear Information System (INIS)

    Lee, J.D.; Bandini, B.R.

    1985-01-01

    More rigorous nuclear analysis, including treatment of resonance self-shielding effects coupled with an optimization procedure, has resulted in improved performance of the Be/Li/Th blanket. Net U-233 breeding ratio has increased 36% (to 0.84) while at an average U-233/Th ratio of 0.5 a/o average energy multiplication has increased only 12% (to 2.1) compared with earlier results

  3. Thermally induced outdiffusion studies of deuterium in ceramic breeder blanket materials after irradiation

    Energy Technology Data Exchange (ETDEWEB)

    González, Maria, E-mail: maria.gonzalez@ciemat.es [LNF-CIEMAT, Materials for Fusion Group, Madrid (Spain); Carella, Elisabetta; Moroño, Alejandro [LNF-CIEMAT, Materials for Fusion Group, Madrid (Spain); Kolb, Matthias H.H.; Knitter, Regina [Karlsruhe Institute of Technology, Institute for Applied Materials (IAM-WPT), Karlsruhe (Germany)

    2015-10-15

    Highlights: • Surface defects in Lithium-based ceramics are acting as trapping centres for deuterium. • Ionizing radiation affects the deuterium sorption and desorption processes. • By extension, the release of the tritium produced in a fusion breeder will be effective. - Abstract: Based on a KIT–CIEMAT collaboration on the radiation damage effects of light ions sorption/desorption in ceramic breeder materials, candidate materials for the ITER EU TBM were tested for their outgassing behavior as a function of temperature and radiation. Lithium orthosilicate based pebbles with different metatitanate contents and pellets of the individual oxide components were exposed to a deuterium atmosphere at room temperature. Then the thermally induced release of deuterium gas was registered up to 800 °C. This as-received behavior was studied in comparison with that after exposing the deuterium-treated samples to 4 MGy total dose of gamma radiation. The thermal desorption spectra reveal differences in deuterium sorption/desorption behavior depending on the composition and the induced ionizing damage. In these breeder candidates, strong desorption rate at approx. 300 °C takes place, which slightly increases with increasing amount of the titanate second phase. For all studied materials, ionizing radiation induces electronic changes disabling a number of trapping centers for D{sub 2} adsorption.

  4. Chemical compatibility study between ceramic breeder and EUROFER97 steel for HCPB-DEMO blanket

    Energy Technology Data Exchange (ETDEWEB)

    Mukai, Keisuke, E-mail: keisuke.mukai@kit.edu [Institute for Applied Materials (IAM), Karlsruhe Institute of Technology, 76021 Karlsruhe (Germany); Sanchez, Fernando [National Fusion Laboratory, Division of Fusion Technology, CIEMAT, 28040 Madrid (Spain); Knitter, Regina [Institute for Applied Materials (IAM), Karlsruhe Institute of Technology, 76021 Karlsruhe (Germany)

    2017-05-15

    Chemical compatibility between ceramic breeder (Li{sub 4}SiO{sub 4} + 20 mol% addition of Li{sub 2}TiO{sub 3}) and EUROFER97 steel was examined in this study. These materials were contacted and heated at 623, 823 and 1073 K under He + 0.1 vol.% H{sub 2} atmosphere for up to 12 weeks. Limited influence was found in the breeder specimens, although losses of the constituent elements appeared near the surface of the breeder pellets heated at 1073 K. For the EUROFER specimens with formation of a corrosion layer, element diffusivity was estimated based on diffusion kinetics. In the temperature range, effective diffusion coefficients of oxygen into EUROFER steel were in the range from 3.5 × 10{sup −134} to 2.5 × 10{sup −112} cm/s{sup 2} and found to be faster than that of Li. The coefficients yielded an activation energy of 0.93 eV for oxygen diffusion into EUROFER steel and predicted the possible thickness of the corrosion layer after operational periods.

  5. Conceptual design and testing strategy of a dual functional lithium-lead test blanket module in ITER and EAST

    International Nuclear Information System (INIS)

    Wu, Y.

    2007-01-01

    A dual functional lithium-lead (DFLL) test blanket module (TBM) concept has been proposed for testing in the International Thermonuclear Experimental Reactor (ITER) and the Experimental Advanced Superconducting Tokamak (EAST) in China to demonstrate the technologies of the liquid lithium-lead breeder blankets with emphasis on the balance between the risks and the potential attractiveness of blanket technology development. The design of DFLL-TBM concept has the flexibility of testing both the helium-cooled quasi-static lithium-lead (SLL) blanket concept and the He/PbLi dual-cooled lithium-lead (DLL) blanket concept. This paper presents an effective testing strategy proposed to achieve the testing target of SLL and DLL DEMO blankets relevant conditions, which includes three parts: materials R and D and small-scale out-of-pile mockups testing in loops, middle-scale TBMs pre-testing in EAST and full-scale consecutive TBMs testing corresponding to different operation phases of ITER during the first 10 years. The design of the DFLL-TBM concept and the testing strategy ability to test TBMs for both blanket concepts in sequence and or in parallel for both ITER and EAST are discussed

  6. Conceptual design of blanket structures for fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-03-01

    Conceptual design study for in-vessel components including tritium breeding blanket of FER has been carried out. The objective of this study is to obtain the engineering and technological data for selecting the reactor concept and for its construction by investigating fully and broadly. The design work covers in-vessel components (such as tritium breeding blanket, first wall, shield, divertor and blanket test module), remote handling system and tritium system. The designs of those components and systems are accomplished in consideration of their accomodation to whole reactor system and problems for furthur study are clarified. (author)

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

    International Nuclear Information System (INIS)

    1978-01-01

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

  8. Transmutation blanket design for a Tokamak system

    International Nuclear Information System (INIS)

    Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria A. Fortini; Costa, Antonella L.

    2011-01-01

    Sub-critical advanced reactor with a D-T fusion neutron source based on Tokamak technology is an innovative type of nuclear system. Due to the high quantity of neutrons produced by fusion reactions, it could be well spent in the transmutation process of the transuranic elements. Nevertheless, to achieve a successful transmutation, it is necessary to know the neutron fluence along the radial axis and its characteristics. In this work, it evaluated the neutron flux and interaction frequency along the radial axis changing the material of the first wall. W-alloy, beryllium and the combination of both were studied and regions more suitable to transmutation were determined. The results demonstrated that the better zone to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements W-alloy/W-alloy and W-alloy/Beryllium would be able to hold the requirements of high fluence and hardening spectrum needed to transuranic transmutation. The system was simulated using the MCNP5 code, the ITER Final Design Report, 2001, and the FENDL/MC-2.1 nuclear data library. (author)

  9. A blanket design, apparatus, and fabrication techniques for the mass production of multilayer insulation blankets for the Superconducting Super Collider

    International Nuclear Information System (INIS)

    Gonczy, J.D.; Boroski, W.N.; Niemann, R.C.; Otavka, J.G.; Ruschman, M.K.; Schoo, C.J.

    1989-09-01

    The multilayer insulation (MLI) system for the Superconducting Super Collider (SSC) consists of full cryostat length assemblies of aluminized polyester film fabricated in the form of blankets and installed as blankets to the 4.5K cold mass and the 20K and 80K thermal radiation shields. Approximately 40,000 MLI blankets will be required in the 10,000 cryogenic devices comprising the SSC accelerator. Each blanket is nearly 17 meters long and 1.8 meters wide. This paper reports the blanket design, an apparatus, and the fabrication method used to mass produce pre-fabricated MLI blankets. Incorporated in the blanket design are techniques which automate quality control during installation of the MLI blankets in the SSC cryostat. The apparatus and blanket fabrication method insure consistency in the mass produced blankets by providing positive control of the dimensional parameters which contribute to the thermal performance of the MLI blanket. By virtue of the fabrication process, the MLI blankets have inherent features of dimensional stability three-dimensional uniformity, controlled layer density, layer-to-layer registration, interlayer cleanliness, and interlayer material to accommodate thermal contraction differences. 11 refs., 6 figs., 1 tab

  10. Activation analysis and waste management for dual-cooled lithium lead breeder (DLL) blanket of the fusion power reactor FDS-II

    International Nuclear Information System (INIS)

    Chen Mingliang; Huang Qunying; Li Jingjing; Zeng Qin; Wu Yican

    2005-01-01

    The calculation and analysis on the activation levels of the different regions of dual-cooled lithium-lead (DLL) breeder blanket of FDS-II, including afterheat, dose rate, activity and biological hazard potential after shutdown, were carried out with the neutronics code system VisualBUS and multi-group working library HENDL1.0/MG. The safety and environment assessment of fusion power (SEAFP) strategy for the management of activated material is here applied to the DLL blanket, to define the suitable recycling (reuse of activated material) procedure and the possibility of clearance (declassification of the material with low activity level to non-active waste). (authors)

  11. A helium-cooled blanket design of the low aspect ratio reactor

    International Nuclear Information System (INIS)

    Wong, C.P.; Baxi, C.B.; Reis, E.E.; Cerbone, R.; Cheng, E.T.

    1998-03-01

    An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh

  12. Tritium-assisted fusion breeders

    International Nuclear Information System (INIS)

    Greenspan, E.; Miley, G.H.

    1983-08-01

    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 233 U 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 3 He 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 3 He 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

  13. Achievements in the development of the Water Cooled Solid Breeder Test Blanket Module of Japan to the milestones for installation in ITER

    International Nuclear Information System (INIS)

    Tsuru, Daigo; Tanigawa, Hisashi; Hirose, Takanori; Mohri, Kensuke; Seki, Yohji; Enoeda, Mikio; Ezato, Koichiro; Suzuki, Satoshi; Nishi, Hiroshi; Akiba, Masato

    2009-01-01

    As the primary candidate of ITER Test Blanket Module (TBM) to be tested under the leadership of Japan, a water cooled solid breeder (WCSB) TBM is being developed. This paper shows the recent achievements towards the milestones of ITER TBMs prior to the installation, which consist of design integration in ITER, module qualification and safety assessment. With respect to the design integration, targeting the detailed design final report in 2012, structure designs of the WCSB TBM and the interfacing components (common frame and backside shielding) that are placed in a test port of ITER and the layout of the cooling system are presented. As for the module qualification, a real-scale first wall mock-up fabricated by using the hot isostatic pressing method by structural material of reduced activation martensitic ferritic steel, F82H, and flow and irradiation test of the mock-up are presented. As for safety milestones, the contents of the preliminary safety report in 2008 consisting of source term identification, failure mode and effect analysis (FMEA) and identification of postulated initiating events (PIEs) and safety analyses are presented.

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

    Directory of Open Access Journals (Sweden)

    E. Carella

    2015-07-01

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

  15. Manufacturing experiment on a cooling plate for a blanket breeder unit

    International Nuclear Information System (INIS)

    Weth, A. von der; Aktaa, J.

    2008-01-01

    Plates with curved cooling channels will be used as structural elements in a breeding blanket of a future fusion power plant. Such power plants are a promising attempt for future electrical energy production. The central manufacturing process of such cooling plates is a diffusion welding process. Such a process has been 'available' on a laboratory scale for years. But this diffusion welding process has not yet been applied on an industrial scale. This contribution documents our first attempt to transfer this to industry, a so-called uniaxial diffusion welding setup. The industrial transfer was attempted in two steps: (1) On a small cooling plate mock-up and (2) On a true-scale cooling plate. The problems with the technical transfer of the diffusion welding process from the laboratory scale to the true scale were outlined

  16. Helium-cooled pebble bed test blanket module alternative design and fabrication routes

    International Nuclear Information System (INIS)

    Lux, M.

    2007-01-01

    According to first results of the recently started European DEMO study, a new blanket integration philosophy was developed applying so-called multi-module segments. These consist of a number of blanket modules flexibly mounted onto a common vertical manifold structure that can be used for replacing all modules in one segment at one time through vertical remote-handling ports. This principle gives new freedom in the design choices applied to the blanket modules itself. Based on the alternative design options considered for DEMO also the ITER test blanket module was newly analyzed. As a result of these activities it was decided to keep the major principles of the reference design like stiffening grid, breeder unit concept and perpendicular arrangement of pebble beds related to the First Wall because of the very positive results of thermo-mechanical and neutronics studies. The present paper gives an overview on possible further design optimization and alternative fabrication routes. One of the most significant improvements in terms of the hydraulic performance of the Helium cooled reactor can be reached with a new First Wall concept. That concept is based on an internal heat transfer enhancement technique and allows drastically reducing the flow velocity in the FW cooling channels. Small ribs perpendicular to the flow direction (transverse-rib roughness) are arranged on the inner surface of the First Wall cooling channels at the plasma side. In the breeder units cooling plates which are mostly parallel but bent into U-shape at the plasma-side are considered. In this design all flow channels are parallel and straight with the flow entering on one side of the parallel plate sections and exiting on the other side. The ceramic pebble beds are embedded between two pairs of such type of cooling plates. Different modifications could possibly be combined, whereby the most relevant discussed in this paper are (i) rib-cooled First Wall channels, (ii) U-bent cooling plates for

  17. Tritium extraction methods proposed for a solid breeder blanket. Subtask WP-B 6.1 of the European Blanket Program 1996

    International Nuclear Information System (INIS)

    Albrecht, H.

    1997-04-01

    Ten different methods for the extraction of tritium from the purge gas of a ceramic blanket are described and evaluated with respect to their applicability for ITER and DEMO. The methods are based on the conditions that the purge gas is composed of helium with an addition of up to 0.1% of H 2 or O 2 and H 2 O to facilitate the release of tritium, and that tritium occurs in the purge gas in two main chemical forms, i.e. HT and HTO. Individual process steps of many methods are identical; in particular, the application of cold traps, molecular sieve beds, and diffusors are proposed in several cases. Differences between the methods arise mainly from the ways in which various process steps are combined and from the operating conditions which are chosen with respect to temperature and pressure. Up to now, none of the methods has been demonstrated to be reliably applicable for the purge gas conditions foreseen for the operation of an ITER blanket test module (or larger ceramic blanket designs such as for DEMO). These conditions are characterized by very high gas flow rates and extremely low concentrations of HT and HTO. Therefore, a proposal has been made (FZK concept) which is expected to have the best potential for applicability to ITER and DEMO and to incorporate the smallest development risk. In this concept, the extraction of tritium and excess hydrogen is accomplished by using a cold trap for freezing out HTO/H 2 O and a 5A molecular sieve bed for the adsorption of HT/H 2 . (orig.) [de

  18. Irradiaiton facilities for testing solid and liquid blanket breeder materials with in-situ tritium release measurements in the HFR Petten

    International Nuclear Information System (INIS)

    Conrad, R.; Debarberis, L.

    1991-01-01

    Lithium-based tritium breeder materials for solid and liquid fusion reactor blanket concepts are being tested in the High Flux Reactor (HFR) Petten with in-situ tritium release measurements since 1985, within the European Fusion Technology Programme and the BEATRIX-I programme. Ceramic breeder materials are being tested in the EXOTIC and COMPLIMENT experimental programmes and the liquid breeder material, Pb-17Li, is being tested in the LIBRETTO experimental programme. The in-pile experiments are performed with irradiation facilities developed by the Joint Research Centre (JRC) Petten. The irradiation vehicles are multi-channel rigs. The sample holders consist of independent, fully instrumented and triple contained capsules. The out-of-pile experimental equipment consist of twelve independent circuits for on-line tritium release and tritium permeation measurements and eight independent circuits for temperature control. The experimental achievements obtained so far contribute to the selection of candidate tritium breeder materials for blanket concepts of near future machines like NET, ITER and DEMO. (orig.)

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

    International Nuclear Information System (INIS)

    Ciampichetti, A.; Nitti, F.S.; Aiello, A.; Ricapito, I.; Liger, K.; Demange, D.; Sedano, L.; Moreno, C.; Succi, M.

    2012-01-01

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

  20. U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li{sub 2}O) and lithium zirconate (Li{sub 2}ZrO{sub 3}) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers.

  1. U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

    International Nuclear Information System (INIS)

    1995-01-01

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li 2 O) and lithium zirconate (Li 2 ZrO 3 ) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers

  2. Status of blanket design for RTO/RC ITER

    International Nuclear Information System (INIS)

    Yamada, M.; Ioki, K.; Cardella, A.; Elio, F.; Miki, N.

    2000-01-01

    Design has progressed on the FW/blanket for the RTO/RC (reduced technical objective/ reduced cost) ITER. The basic functions and structures are the same as for the 1998 ITER design. However, design and fabrication methods of the FW/blanket have been improved to achieve ∝ 50% reduction of the construction cost compared to that for the 1998 ITER design. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the EDA (engineering design activity) is still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed. (orig.)

  3. Tritium breeding blanket

    International Nuclear Information System (INIS)

    Smith, D.; Billone, M.; Gohar, Y.; Baker, C.; Mori, S.; Kuroda, T.; Maki, K.; Takatsu, H.; Yoshida, H.; Raffray, A.; Sviatoslavsky, I.; Simbolotti, G.; Shatalov, G.

    1991-01-01

    The terms of reference for ITER provide for incorporation of a tritium breeding blanket with a breeding ratio as close to unity as practical. A breeding blanket is required to assure an adequate supply of tritium to meet the program objectives. Based on specified design criteria, a ceramic breeder concept with water coolant and an austenitic steel structure has been selected as the first option and lithium-lead blanket concept has been chosen as an alternate option. The first wall, blanket, and shield are integrated into a single unit with separate cooling systems. The design makes extensive use of beryllium to enhance the tritium breeding ratio. The design goals with a tritium breeding ratio of 0.8--0.9 have been achieved and the R ampersand D requirements to qualify the design have been identified. 4 refs., 8 figs., 2 tabs

  4. ITER blanket module shield block design and analysis

    International Nuclear Information System (INIS)

    Mitin, D.; Khomyakov, S.; Razmerov, A.; Strebkov, Yu.

    2008-01-01

    This paper presents the alternative design of the shield block cooling path for a typical ITER blanket module with a predominantly sequential flow circuit. A number of serious disadvantages have been observed for the reference design, where the parallel flow circuit is used, which is inherent in the majority of blanket modules. The paper discusses these disadvantages and demonstrates the benefit of the alternative design based on the detailed design and the technological, hydraulic, thermal, structural and strength analyses, conducted for module no. 17

  5. Liquid metal cooled blanket concept for NET

    International Nuclear Information System (INIS)

    Malang, S.; Casal, V.; Arheidt, K.; Fischer, U.; Link, W.; Rust, K.

    1986-01-01

    A blanket concept for NET using liquid lithium-lead both as breeder material and as coolant is described. The need for inboard breeding is avoided by using beryllium as neutron multiplier in the outboard blanket. Novel flow channel inserts are employed in all poloidal ducts to reduce the MHD pressure drop. The concept offers a simple mechanical design and a higher tritium breeding ratio compared to water- and gas-cooled blankets. (author)

  6. Design, Manufacture, and Experimental Serviceability Validation of ITER Blanket Components

    Science.gov (United States)

    Leshukov, A. Yu.; Strebkov, Yu. S.; Sviridenko, M. N.; Safronov, V. M.; Putrik, A. B.

    2017-12-01

    In 2014, the Russian Federation and the ITER International Organization signed two Procurement Arrangements (PAs) for ITER blanket components: 1.6.P1ARF.01 "Blanket First Wall" of February 14, 2014, and 1.6.P3.RF.01 "Blanket Module Connections" of December 19, 2014. The first PA stipulates development, manufacture, testing, and delivery to the ITER site of 179 Enhanced Heat Flux (EHF) First Wall (FW) Panels intended for withstanding the heat flux from the plasma up to 4.7MW/m2. Two Russian institutions, NIIEFA (Efremov Institute) and NIKIET, are responsible for the implementation of this PA. NIIEFA manufactures plasma-facing components (PFCs) of the EHF FW panels and performs the final assembly and testing of the panels, and NIKIET manufactures FW beam structures, load-bearing structures of PFCs, and all elements of the panel attachment system. As for the second PA, NIKIET is the sole official supplier of flexible blanket supports, electrical insulation key pads (EIKPs), and blanket module/vacuum vessel electrical connectors. Joint activities of NIKIET and NIIEFA for implementing PA 1.6.P1ARF.01 are briefly described, and information on implementation of PA 1.6.P3.RF.01 is given. Results of the engineering design and research efforts in the scope of the above PAs in 2015-2016 are reported, and results of developing the technology for manufacturing ITER blanket components are presented.

  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)

    McCarville, T.J.; Berwald, D.H.; Wong, C.P.C.

    1984-09-01

    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. Preliminary design of the Neutron Spectral Shifter that is dedicated to the IFMIF Liquid Breeder Validation Module

    Energy Technology Data Exchange (ETDEWEB)

    Mas, A., E-mail: amassanchez@gmail.com; Mota, F.; Casal, N.; García, A.; Rapisarda, D.; Arroyo, J.M.; Molla, J.; Ibarra, A.

    2014-10-15

    The International Fusion Materials Irradiation Facility (IFMIF) has a D-Li neutron stripping source that provides typical fusion irradiation conditions for material testing. The Liquid Breeder Validation Module (LBVM) is one of the medium flux test modules of the IFMIF that is used to account for some of the DEMO liquid breeder blanket R and D needs. Previous analyses have shown that the main irradiation parameters (He (appm)/dpa and H (appm)/dpa) in the medium flux area of the IFMIF can be improved to fit the expected parameters in the DEMO reactor for functional materials of liquid breeder blankets. Therefore, the design of an additional module, called the Neutron Spectral Shifter (NSS), has been considered to optimize the irradiation conditions of LBVM experiments. The proposed concept consists of supported tungsten plates working as a shifter material inside a steel structure. This design assures the mechanical integrity of the different components and it fulfills the neutronic requirements as well as the cooling capability. This present paper summarizes the work devoted to the design of the LBVM Neutron Spectral Shifter as well as the results of neutronic, thermo-hydraulic, mechanical and safety studies carried out to validate the design.

  9. Preliminary design of a Binary Breeder Reactor

    International Nuclear Information System (INIS)

    Garcia C, E. Y.; Francois, J. L.; Lopez S, R. C.

    2014-10-01

    A binary breeder reactor (BBR) is a reactor that by means of the transmutation and fission process can operates through the depleted uranium burning with a small quantity of fissile material. The advantages of a BBR with relation to other nuclear reactor types are numerous, taking into account their capacity to operate for a long time without requiring fuel reload or re-arrangement. In this work four different simulations are shown carried out with the MCNPX code with libraries Jeff-3.1 to 1200 K. The objective of this study is to compare two different models of BBR: a spherical reactor and a cylindrical one, using two fuel cycles for each one of them (U-Pu and Th-U) and different reflectors for the two different geometries. For all the models a super-criticality state was obtained at least 10.9 years without carrying out some fuel re-arrangement or reload. The plutonium-239 production was achieved in the models where natural uranium was used in the breeding area, while the production of uranium-233 was observed in the cases where thorium was used in the fertile area. Finally, a behavior of stationary wave reactor was observed inside the models of spherical reactor when contemplating the power uniform increment in the breeding area, while inside the cylindrical models was observed the behavior of a traveling wave reactor when registering the displacement of the burnt wave along the cylindrical model. (Author)

  10. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

    Gohar, Y.; Baker, C.C.; Smith, D.L.

    1987-10-01

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

  11. A study of sodium-cooled fast breeder reactor with thorium blanket for supply of U-233 to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Yoshida, H.; Nishimura, H.; Osugi, T.

    1978-08-01

    Symbiotic energy system between fast breeder reactor and thermal reactor would have a potential merit for nuclear proliferation problem. And when using HTGR as the thermal reactor in the system, the energy system appears to be promising as an energy system self-sufficient in fuels, which can generate both electricity and high temperature process heat. In the system the fast breeder reactor has to supply sufficient amount of fissile plutonium to keep the reactor going, and also produce U-233 necessary to the associated U-233 fuelled process heat production HTGR. Three types of LMFBR concepts with thorium blanket, conventional homogeneous core LMFBR, and axial and radial parfait heterogeneous core LMFBRs, have been investigated to find out suitable configurations of LMFBR for supply of U-233 to the HTGR with relatively high conversion ratio of 0.85, in the symbiotic energy system between LMFBR and HTGR. The investigation on LMFBR has been made on fuel sufficiency of the system, inherent safety such as sodium-void and Doppler coefficients, and fuel cycle cost. The followings were revealed; (1) Conventional homogeneous core LMFBR with thorium radial blanket well satisfies the condition of fuel sufficiency, if adequate radial blanket thickness is chosen. However, the sodium-void coefficient and fuel cycle cost are inferior to the other concepts. (2) Axial parfait heterogeneous core LMFBR can be regarded as one of the best LMFBR concepts installed in the symbiotic energy system, from the viewpoints of fuel sufficiency, inherent safety and fuel cycle cost. However, further investigations should be needed on reliability and operationability of the concept. (3) Radial parfait heterogeneous core LMFBR seems inadequate as the LMFBR in the system, because the configurations based on this concept does not satisfy plutonium and U-233 breedings, simultaneously. This LMFBR concept, however, has excellent breeding performance in the internal radial blanket. So further

  12. Neutronics design aspects of reference ARIES-I fusion blanket

    International Nuclear Information System (INIS)

    Cheng, E.T.

    1990-12-01

    A SiC composite blanket concept was recently conceived for a deuterium-tritium burning, 1000 MW(e) tokamak fusion reactor design, ARIES-I. SiC composite structural material was chosen due to its very low activation features. High blanket nuclear performance and thermal efficiency, adequate tritium breeding, and a low level of activation are important design requirements for the ARIES-I reactor. The major approaches, other than using SiC as structural material, in meeting these design requirements, are to employ beryllium, the only low activation neutron multiplying material, and isotopically tailored Li 2 ZrO 3 , a tritium breeding material stable at high temperature, as blanket materials. 5 refs., 4 figs., 2 tabs

  13. Conceptual design of Indian molten salt breeder reactor

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Basak, A.; Dulera, I.V.; Vaze, K.K.; Basu, S.; Sinha, R.K.

    2014-01-01

    The fuel in a molten salt breeder reactor is in the form of a continuously circulating molten salt. Fluoride based salts have been almost universally proposed. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. This constitutes a major technological challenge for this type of reactors. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian Molten Salt Breeder Reactor (IMSBR). Presently various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel fundamental studies as regards various molten salts have also been initiated. This paper would discuss conceptual design of these reactors, as well as associated issues and technologies

  14. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

    A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. 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. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

  15. Design requirement on KALIMER blanket fuel assembly duct

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang, H. Y.; Nam, C.; Kim, J. O.

    1998-03-01

    This document describes design requirements which are needed for designing the blanket fuel assembly duct of the KALIMER as design guidance. The blanket fuel assembly duct of the KALIMER consists of fuel rods, mounting rail, nosepiece, duct with pad, handling socket with pad. Blanket fuel rod consists of top end plug, bottom end plug with solid ferritic-martensitic steel rod and key way blanket fuel slug, cladding, and wire wrap. In the assembly, the rods are in a triangular pitch array, and the rod bundle is attached to the nosepiece with mounting rails. The bottom end of the assembly duct is formed by a long nosepiece which provides the lower restraint function and the paths for coolant inlet. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. (author). 20 refs., 4 figs

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

    International Nuclear Information System (INIS)

    Fuetterer, M.A.; Raepsaet, X.; Proust, E.

    1994-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Fuetterer, M A; Raepsaet, X; Proust, E

    1994-12-31

    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.

  18. Safety design study of fast breeder reactors in Japan

    International Nuclear Information System (INIS)

    Miura, M.; Inagaki, T.

    1992-01-01

    This paper reports on two fast breeder reactor (FBR) concepts, the tank type and the loop type, that have been studied as possible reactor designs to be used for a demonstration FBR (DFBR). The basic principle fo the DFBR design is to ensure plant safety through a defense-in-depth methodology. Improvements in the seismic and thermal stress designs have been attempted for both reactor concepts. The system design study strives to maximize the reliability of the safety-related systems and to rationalize commercialization of the plant

  19. Breeder design for enhanced performance and safety characteristics

    International Nuclear Information System (INIS)

    Fischer, G.J.; Atefi, B.; Yang, J.W.; Galperin, A.; Segev, M.

    1980-01-01

    A fast breeder reactor design has been created which offers a considerably extended fuel cycle and excellent performance characteristics. An example of a core designed to operate on a ten-year fuel cycle is described in some detail. Use of metal fuel along with a moderator such as beryllium oxide dispersed throughout the core provides both design flexibility and safety advantages such as a strong Doppler feedback and limited sodium void reactivity gain. Local power variations are small for the entire cycle; control requirements are also modest, and fuel cycle costs are low

  20. Liquid Metal Fast Breeder Reactor plant maintenance and equipment design

    International Nuclear Information System (INIS)

    Swannack, D.L.

    1982-01-01

    This paper provides a summary of maintenance equipment considerations and actual plant handling experiences from operation of a sodium-cooled reactor, the Fast Flux Test Facility (FFTF). Equipment areas relating to design, repair techniques, in-cell handling, logistics and facility services are discussed. Plant design must make provisions for handling and replacement of components within containment or allow for transport to an ex-containment area for repair. The modular cask assemblies and transporter systems developed for FFTF can service major plant components as well as smaller units. The plant and equipment designs for the Clinch River Breeder Reactor (CRBR) plant have been patterned after successful FFTF equipment

  1. Thermal and mechanical design of WITAMIR-I blanket

    International Nuclear Information System (INIS)

    Sze, D.K.; Sviatoslavsky, I.N.

    1980-10-01

    The design philosophy of WITAMIR-I, a Wisconsin Tandem Mirror Reactor design study, uses the experience obtained from our previous tokamak studies and combines it with the unique features of the tandem mirror to obtain an attractive design of a TM power reactor. It is aimed at maximizing the strengths of the tandem mirror while mitigating its weaknesses. The end product should be a safe, reliable, maintainable and a relatively economic power reactor. The general description of the reactor, the plasma calculations, the magnet design, the neutronic calculations and the maintenance considerations are presented elsewhere. This paper presents the blanket design of this reactor study

  2. Neutronic analysis of the European reference design of the water cooled lithium lead blanket for a DEMOnstration reactor

    International Nuclear Information System (INIS)

    Petrizzi, L.

    1994-01-01

    Water cooled lithium lead blankets, using liquid Pb-17Li eutectic both as breeder and neutron multiplier material, and martensitic steel as structural material, represent one of the four families under development in the European DEMO blanket programme. Two concepts were proposed, both reaching tritium breeding self-sufficiency: the 'box-shaped' and the 'cylindrical modules'. Also to this scope a new concept has been defined: 'the single box'. A neutronic analysis of the 'single box' is presented. A full 3-D model including the whole assembly and many of the reactor details (divertors, holes, gaps) has been defined, together with a 3-D neutron source. A tritium breeding ration (TBR) value of 1.19 confirms the tritium breeding self-sufficiency of the design. Selected power densities, calculated for the different materials and zones, are here presented. Some shielding capability considerations with respect to the toroidal field coil system are presented too. (author) 10 refs.; 3 figs.; 3 tabs

  3. Human factors engineering in Clinch River Breeder plant design

    International Nuclear Information System (INIS)

    Planchon, H.P. Jr.; Kaushal, N.N.; Snider, J.

    1982-01-01

    The Clinch River Breeder Reactor Plant (CRBRP) Project formed a Control Room Task Force to ensure that lessons learned from the Three Mile Island accident are incorporated into the design. The charter for the Control Room Task Force was to review plant operations from the control room. The focus was on the man-machine interface to ensure that the systems' designs and operator actions meshed to properly support plant operation during normal and off-normal conditions. Specific items included for review are described. This paper describes the methodology utilized to accomplish the Task Forces' objectives and the results of the review

  4. A Li-particulate blanket concept for ITER

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.

    1989-01-01

    The Li-particulate blanket design concept the authors proposed for the International Thermonuclear Experimental Reactor (ITER) uses a dilute suspension of fine solid breeder particles in a carrier gas as the combined coolant and lithium breeder stream. This blanket concept has a simple mechanical and hydraulic configuration, low inventory of bred tritium, and simple tritium extraction system. Existing technology can be used to implement the design for ITER. The concept has the potential to be a highly reliable shield and blanket design for ITER with relatively low development and capital costs

  5. Coated ceramic breeder materials

    Science.gov (United States)

    Tam, Shiu-Wing; Johnson, Carl E.

    1987-01-01

    A breeder material for use in a breeder blanket of a nuclear reactor is disclosed. The breeder material comprises a core material of lithium containing ceramic particles which has been coated with a neutron multiplier such as Be or BeO, which coating has a higher thermal conductivity than the core material.

  6. Moving ring field-reversed mirror blanket design considerations

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, L.; Kessel, C.; Norman, J.; Schultz, K.R.

    1981-01-01

    A blanket design for the Moving Ring Field-Reversed Mirror Reactor (MRFRM) is presented in this paper. The design emphasis is placed on minimizing the induced radioactivities in the first-wall, blanket and shield. To this end, aluminum-alloy was selected as the reference structural material, giving dose rates two weeks after shutdown that are 3 to 4 orders of magnitude lower than comparable steel structures. The aluminum first-wall is water-cooled and thermally insulated from the high temperature SiC-clad Li 2 O tritium breeding zone. A local tritium breeding ratio of 1.05 was obtained for the design. The tritium is extracted from the Li 2 O by the use of a small dry helium purge stream through the SiC tubes. About 1 ppM hydrogen is added to the helium purge stream to enhance the tritium recovery rate. Helium at 28 atmospheres pressure is circulated through the blanket and shield, with an outlet temperature of 850 0 C, which is coupled with an existing small size closed-cycle gas turbine (CCGT) power conversion system. The spatial and temporal variations of the first-wall temperature caused by the translational movement of the plasma rings along the axis of the cylindrical reactor were evaluated. The after-heat cooling problems of the first-wall were also considered

  7. Mirror hybrid reactor blanket and power conversion system conceptual design

    International Nuclear Information System (INIS)

    Schultz, K.R.; Backus, G.A.; Baxi, C.B.; Dee, J.B.; Estrine, E.A.; Rao, R.; Veca, A.R.

    1976-01-01

    The conceptual design of the blanket and power conversion system for a gas-cooled mirror hybrid fusion-fission reactor is presented. The designs of the fuel, blanket module and power conversion system are based on existing gas-cooled fission reactor technology that has been developed at General Atomic Company. The uranium silicide fuel is contained in Inconel-clad rods and is cooled by helium gas. The fuel is contained in 16 spherical segment modules which surround the fusion plasma. The hot helium is used to raise steam for a conventional steam cycle turbine generator. The details of the method of support for the massive blanket modules and helium ducts remain to be determined. Nevertheless, the conceptual design appears to be technically feasible with existing gas-cooled technology. A preliminary safety analysis shows that with the development of a satisfactory method of primary coolant circuit containment and support, the hybrid reactor could be licensed under existing Nuclear Regulatory Commission regulations

  8. Japanese contributions to the Japan-US workshop on blanket design/technology

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Seki, Yasushi; Minato, Akio; Kobayashi, Takeshi; Mori, Seiji; Kawasaki, Hiromitsu; Sumita, Kenji.

    1983-02-01

    This report describes Japanese papers presented at the Japan-US Workshop on Blanket Design/Technology which was held at Argonne National Laboratory, November 10 - 11, 1982. Overview of Fusion Experimental Reactor (FER), JAERI's activities related to first wall/blanket/shield, summary of FER blanket and its technology development issues and summary of activities at universities on fusion reactor blanket engineering are covered. (author)

  9. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    1983-10-01

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

  10. Objectives and status of EUROfusion DEMO blanket studies

    Energy Technology Data Exchange (ETDEWEB)

    Boccaccini, L.V., E-mail: lorenzo.boccaccini@kit.edu [Karlsruhe Institute of Technology (KIT) (Germany); Aiello, G.; Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Bachmann, C. [EUROfusion, PPPT, Garching (Germany); Barrett, T. [CCFE, Abingdon OX14 3DB (United Kingdom); Del Nevo, A. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Demange, D. [Karlsruhe Institute of Technology (KIT) (Germany); Forest, L. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Hernandez, F.; Norajitra, P. [Karlsruhe Institute of Technology (KIT) (Germany); Porempovic, G. [Fuziotech Engineering Ltd (Hungary); Rapisarda, D. [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Sardain, P. [CEA/IRFM, 13115 Saint-Paul-lès-Durance (France); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Vala, L. [Centrum výzkumu Řež, 250 68 Husinec-Řež (Czech Republic)

    2016-11-01

    Highlights: • Short description of the new Breeding Blanket Project in the EUROfusion consortium for the design of the EU PPPT DEMO: objectives. • Presentation of the design approach used in the development of the Breeding Blanket design: requirements. • Breeding Blanket design; in particular the four blanket concepts included in the study are presented, recent results highlighted and the status discussed. • Auxiliary systems and related R&D programme: in particular the work areas addressed in the Project (Tritium Technology, Pb-Li and Solid Breeders Technology, First Wall Design and R&D, Manufacturing) are presented, recent results highlighted and the status discussed. - Abstract: The design of a DEMO reactor requires the design of a blanket system suitable of reliable T production and heat extraction for electricity production. In the frame of the EUROfusion Consortium activities, the Breeding Blanket Project has been constituted in 2014 with the goal to develop concepts of Breeding Blankets for the EU PPPT DEMO; this includes an integrated design and R&D programme with the goal to select after 2020 concepts on fusion plants for the engineering phase. The design activities are presently focalized around a pool of solid and liquid breeder blanket with helium, water and PbLi cooling. Development of tritium extraction and control technology, as well manufacturing and development of solid and PbLi breeders are part of the programme.

  11. Neutronic performance issues of the breeding blanket options for the European DEMO fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion—Programme Management Unit, Boltzmannstr. 2, 85748 Garching (Germany); Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, SERMA, LPEC, 91191 Gif-sur-Yvette (France); Moro, F. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy)

    2016-11-01

    Highlights: • Breeder blanket concepts for DEMO—design features. • Neutronic characteristics of breeder blankets. • Evaluation of Tritium breeding potential. • Evaluation of shielding performance. - Abstract: This paper presents nuclear performance issues of the HCPB, HCLL, DCLL and WCLL breeder blankets, which are under development within the PPPT (Power Plant Physics and Technology) programme of EUROfusion, with the objective to assess the potential and suitability of the blankets for the application to DEMO. The assessment is based on the initial design versions of the blankets developed in 2014. The Tritium breeding potential is considered sufficient for all breeder blankets although the initial design versions of the HCPB, HCLL and DCLL blankets were shown to require further design improvements. Suitable measures have been proposed and proven to be sufficient to achieve the required Tritium Breeding Ratio (TBR) ≥ 1.10. The shielding performance was shown to be sufficient to protect the super-conducting toroidal field coil provided that efficient shielding material mixtures including WC or borated water are utilized. The WCLL blanket does not require the use of such shielding materials due to a very compact blanket support structure/manifold configuration which yet requires design verification. The vacuum vessel can be safely operated over the full anticipated DEMO lifetime of 6 full power years for all blanket concepts considered.

  12. Conceptual design of reactor assembly of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Selvaraj, A.; Balasubramaniyan, V.; Raghupathy, S.; Elango, D.; Sodhi, B.S.; Chetal, S.C.; Bhoje, S.B.

    1996-01-01

    The conceptual design of Reactor Assembly of 500 MWe Prototype Fast Breeder Reactor (as selected in 1985) was reviewed with the aim of 'simplification of design', 'Compactness of the reactor assembly' and 'ease in construction'. The reduction in size has been possible by incorporating concentric core arrangement, adoption of elastomer seals for Rotatable plugs, fuel handling with one transfer arm type mechanism, incorporation of mechanical sealing arrangement for IHX at the penetration in Inner vessel redan and reduction in number of components. The erection of the components has been made easier by adopting 'hanging' support for roof slab with associated changes in the safety vessel design. This paper presents the conceptual design of the reactor assembly components. (author). 8 figs, 2 tabs

  13. Pebble bed blanket design for deuterium burning tandem mirror reactors

    International Nuclear Information System (INIS)

    Grotz, S.P.; Dhir, V.K.

    1983-01-01

    The UCLA tandem mirror reactor, SATYR, was developed around the capability of tandem mirrors with thermal barriers to burn deuterium at reasonable efficiency levels. The pebble bed concept has been incorporated into our blanket design for the following reasons: 1) Large area-to-volume ratio for purposes of heat removal; 2) Large volume of structure for high thermal capacity thus increasing the safety margin during off-normal incidents; 3) Relatively inexpensive manufacturing costs because of large acceptable tolerances and lack of exotic materials (i.e., lithium). A simplified stress analysis of the blanket module was performed to optimize and simplify the design. The pre-specified stress intensity limitations used were based upon a 30-year predicted lifetime for each module. Along with stress analysis of the vessel a detailed thermal hydraulic analysis of the pebble bed has been completed. Parameters affecting the pebble bed design are fluidization velocity, pressure drop, heat transfer coefficient, thermally induced stress in the spheres and spatial variation of the power density. Although reasonable gross thermal efficiencies of the 2 designs has been achieved (28% for H 2 O and 39% for He) the high net recirculating power fraction for heating and neutral beams results in relatively low net plant efficiencies (21% and 27%). The results show that a blanket can be designed with good thermal efficiency and a relative-ly simple configuration. However, application of this concept to the high Q deuterium-tritium fuel cycle would have difficulties resulting from the need for continuous removal of the tritium. (orig./HP)

  14. Blanket Manufacturing Technologies : Thermomechanical Tests on HCLL Blanket Mocks Up

    International Nuclear Information System (INIS)

    Laffont, G.; Cachon, L.; Taraud, P.; Challet, F.; Rampal, G.; Salavy, J.F.

    2006-01-01

    In the Helium Cooled Lithium Lead (HCLL) Blanket concept, the lithium lead plays the double role of breeder and multiplier material, and the helium is used as coolant. The HCCL Blanket Module are made of steel boxes reinforced by stiffening plates. These stiffening plates form cells in which the breeder is slowly flowing. The power deposited in the breeder material is recovered by the breeder cooling units constituted by 5 parallel cooling plates. All the structures such as first wall, stiffening and cooling plates are cooled by helium. Due to the complex geometry of these parts and the high level of pressure and temperature loading, thermo-mechanical phenomena expected in the '' HCLL blanket concept '' have motivated the present study. The aim of this study, carried out in the frame of EFDA Work program, is to validate the manufacturing technologies of HCLL blanket module by testing small scale mock-up under breeder blanket representative operating conditions.The first step of this experimental program is the design and manufacturing of a relevant test section in the DIADEMO facility, which was recently upgraded with an He cooling loop (pressure of 80 bar, maximum temperature of 500 o C,flow rate of 30 g/s) taking the opportunity of synergies with the gas-cooled fission reactor R-and-D program. The second step will deal with the thermo-mechanical tests. This paper focuses on the program made to support the cooling plate mock up tests which will be carried out on the DIADEMO facility (CEA) by thermo-mechanical calculations in order to define the relevant test conditions and the experimental parameters to be monitored. (author)

  15. Design features of the Light Water Breeder Reactor (LWBR) which improve fuel utilization in light water reactors (LWBR development program)

    International Nuclear Information System (INIS)

    Hecker, H.C.; Freeman, L.B.

    1981-08-01

    This report surveys reactor core design features of the Light Water Breeder Reactor which make possible improved fuel utilization in light water reactor systems and breeding with the uranium-thorium fuel cycle. The impact of developing the uranium-thorium fuel cycle on utilization of nuclear fuel resources is discussed. The specific core design features related to improved fuel utilization and breeding which have been implemented in the Shippingport LWBR core are presented. These design features include a seed-blanket module with movable fuel for reactivity control, radial and axial reflcetor regions, low hafnium Zircaloy for fuel element cladding and structurals, and a closely spaced fuel rod lattice. Also included is a discussion of several design modifications which could further improve fuel utilization in future light water reactor systems. These include further development of movable fuel control, use of Zircaloy fuel rod support grids, and fuel element design modifications

  16. Fuel cycles and advanced core designs for the Gas-Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Simon, R.H.; Hamilton, C.J.; Hunter, R.S.

    1982-01-01

    Studies indicate that a 1200 MW(e) Gas-Cooled Fast Breeder Reactor could achieve compound system doubling times of under ten years when using advanced oxide or carbide fuels. In addition, when thorium is used in the breeding blankets, enough U-233 can be generated in each GCFR to supply several advanced converter reactors with fissionable material and this symbiotic relationship could provide energy for the world for centuries. (author)

  17. On the use of tin-lithium alloys as breeder material for blankets of fusion power plants

    International Nuclear Information System (INIS)

    Fuetterer, M.A.; Aiello, G.; Barbier, F.; Giancarli, L.; Poitevin, Y.; Sardain, P.; Szczepanski, J.; Li Puma, A.; Ruvutuso, G.; Vella, G.

    2000-01-01

    Tin-lithium alloys have several attractive thermo-physical properties, in particular high thermal conductivity and heat capacity, that make them potentially interesting candidates for use in liquid metal blankets. This paper presents an evaluation of the advantages and drawbacks caused by the substitution of the currently employed alloy lead-lithium (Pb-17Li) by a suitable tin-lithium alloy: (i) for the European water-cooled Pb-17Li (WCLL) blanket concept with reduced activation ferritic-martensitic steel as the structural material; (ii) for the European self-cooled TAURO blanket with SiC f /SiC as the structural material. It was found that in none of these blankets Sn-Li alloys would lead to significant advantages, in particular due to the low tritium breeding capability. Only in forced convection cooled divertors with W-alloy structure, Sn-Li alloys would be slightly more favorable. It is concluded that Sn-Li alloys are only advantageous in free surface cooled reactor internals, as this would make maximum use of the principal advantage of Sn-Li, i.e., the low vapor pressure

  18. Design development and manufacturing sequence of the European water-cooled Pb-17Li test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Futterer, M.A.; Bielak, B.; Deffain, J.P.; Giancarli, L.; Li Puma, A.; Salavy, J.F.; Szczepanski, J. [CEA Saclay, Gif-sur-Yvette (France). FDRN/DMT/SERMA; Dellis, C. [CEA Grenoble, DTA-CEREM/SGM, Grenoble (France); Nardi, C. [ENEA Frascati, ERG-FUS-TECN-MEC, Frascati (Italy); Schleisiek, K. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Reaktorsicherheit

    1998-09-01

    In 1996, the European Community started the development of a water-cooled Pb17Li blanket test module for ITER. First tests are currently scheduled to start with the beginning of the basic performance phase prior to D-T operation. The test module is designed to be a representative for a DEMO breeding blanket and relies on the liquid alloy Pb-17Li as both tritium breeder and neutron multiplier material, and water at PWR pressure and temperature as coolant. The structural material is martensitic steel. The straight, box-like structure of this blanket confines a pool of liquid Pb-17Li which is slowly circulated for ex-situ tritium extraction and lithium adjustment. The box and the Pb-17Li pool are separately cooled, the former with toroido-radial tubes, the latter with a bundle of double-walled U-tubes, equally made of martensitic steel and equipped with a permeation barrier. This paper presents the latest design and three manufacturing schemes with different degrees of technology. Advanced techniques such as solid or powder HIP are proposed to provide design flexibility. With a 3D neutronics analysis, the power and tritium generation were determined. (orig.) 11 refs.

  19. Deterministic 3D transport, sensitivity and uncertainty analysis of TPR and reaction rate measurements in HCPB Breeder Blanket mock-up benchmark

    International Nuclear Information System (INIS)

    Kodeli, I.

    2006-01-01

    The Helium-Cooled Pebble Bed (HCPB) Breeder Blanket mock-up benchmark experiment was analysed using the deterministic transport, sensitivity and uncertainty code system in order to determine the Tritium Production Rate (TPR) in the ceramic breeder and the neutron reaction rates in beryllium, both nominal values and the corresponding uncertainties. The experiment, performed in 2005 to validate the HCPB concept, consists of a metallic beryllium set-up with two double layers of breeder material (Li 2 CO 3 powder). The reaction rate measurements include the Li 2 CO 3 pellets for the tritium breeding monitoring and activation foils, inserted at several axial and lateral locations in the block. In addition to the well established and validated procedure based on the 2-dimensional (2D) code DORT, a new approach for the 3D modelling was validated based on the TORT/GRTUNCL3D transport codes. The SUSD3D code, also in 3D geometry, was used for the cross-section sensitivity and uncertainty calculations. These studies are useful for the interpretation of the experimental measurements, in particular to assess the uncertainties linked to the basic nuclear data. The TPR, the neutron activation rates and the associated uncertainties were determined using the EFF-3.0 9 Be nuclear cross section and covariance data, and compared with those from other evaluations, like FENDL-2.1. Sensitivity profiles and nuclear data uncertainties of the TPR and detector reaction rates with respect to the cross-sections of 9 Be, 6 Li, 7 Li, O and C were determined at different positions in the experimental block. (author)

  20. Preliminary conceptual design of the blanket and power conversion system for the Mirror Hybrid Reactor

    International Nuclear Information System (INIS)

    Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

    1978-01-01

    A conceptual design of a commercial Mirror Hybrid Reactor, optimized for 239 Pu production, has been completed. This design is the product of a joint effort by Lawrence Livermore Laboratory and General Atomic Company, and follows directly from earlier work on the Mirror Hybrid. This paper describes the blanket and power conversion system of the reactor design. Included are descriptions of the prestressed concrete reactor vessel that supports the magnets and contains the blanket and power conversion system components, the blanket module design, the blanket fuel design, and the power conversion system

  1. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K. E-mail: iokik@itereu.de; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H

    2001-11-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable.

  2. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H.

    2001-01-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable

  3. First wall/blanket/shield design and power conversion for the ARIES-IV tokamak fusion reactor

    International Nuclear Information System (INIS)

    Hasan, M.Z.; Conn, R.W.; Najmabadi, F.

    1994-01-01

    ARIES-IV is a conceptual, D-T burning, steady-state tokamak fusion reactor producing 1000 MWe net. It operates in the second plasma stability regime. The structural material is SiC composite and the primary coolant is helium at 10 MPa base pressure. The coolant flows poloidally in two loops, one inboard and one outboard. The coolant channels are circular tubes that form shells and are placed between two purge plates; the space between two adjacent tubes and the plate is purge gas flow area. The solid breeder is Li 2 O, and Be is used as neutron multiplier to ensure adequate TBR. Beryllium and Li 2 O are placed in between the adjacent tube shells. A computer code was developed to perform and optimize thermal-hydraulic design. Minimization of blanket thickness and the amount of Be, and the maximization of breeder zone thickness were done by iteration with neutronics. The gross thermal efficiency is 49%. The cost of electricity is 68 mills/kWh. The use of low activation SiC composite as the structural material, Li 2 O as the solid breeder, and avoidance of tungsten in the divertor has resulted in a good safety performance, and LSA rating of 1. Overall, SiC/He/Li 2 O ARIES-IV design is expected to have attractive economic and safety advantages

  4. Development of three-dimensional nuclear design program for large fast breeder reactor

    International Nuclear Information System (INIS)

    Inoue, Kohtaro

    1987-01-01

    The report describes a calculation program for core design, called HICOM, and its calculation accuracy. HICOM is designed for three-dimensional neutron diffusion calculation and combustion calculation for large fast breeder reactors to be conducted according to a control rod plan and fuel replacement plan. The improved coarse mesh technique is applied to neutron diffusion calculation. It is demostrated that HICOM permits rapid and accurate operation. For the evaluation of the applicability of HICOM, three-dimensional six-group neutron diffusion calculation is conducted for a 1,000 MWe axial heterogeneous FBR core. Results demonstrate that the program can perform numerical calculation in a time period shorter than 1-40 that for calculation by CITATION (triangle mesh method). This is achieved by using the improved coarse mesh method and carrying out the operation by a vectorial procedure. For the evaluation of the nuclear calculation accuracy of HICOM, analysis is made of reactivity, output distribution and B 4 C control rod worth emasured in an FCA criticality experiment carried out by the Japan Atomic Energy Research Institute. Calculations are found to agree with measurements within a permissible error. The same level of calculation accuracy is obtained for homogneous core, axial heterogeneous core and cores with internal blankets with different forms. (Nogami, K.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  6. Seismic design criteria for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Morrone, A.; Bitner, J.L.; Sigal, G.B.

    1975-01-01

    The general criteria for seismic resistant design for structures, systems and components of the Clinch River Breeder Reactor Plant (CRBRP) are presented and discussed. Site dependency of the maximum ground accelerations for the Operating Basis Earthquake and the Safe Shutdown Earthquake is described from the viewpoint of historical records and geological and seismological studies for the CRBRP site. The respective ground response spectra are derived by normalization of the latest AEC Regulatory standard shapes to these maximum ground accelerations. Modeling and analytical techniques and requirements are given. In addition, loading conditions and categories, loading combinations, earthquake direction effects and allowable damping values are defined. A discussion of the testing criteria which considers both single and multiple frequency test motions, and basic test procedures for single frequency sine beat testing is presented. (U.S.)

  7. Pulsed activation analyses of the ITER blanket design options considered in the blanket trade-off study

    International Nuclear Information System (INIS)

    Wang, Q.; Henderson, D.L.

    1995-01-01

    Pulsed activation calculations have been performed on two blanket options considered as part of the ITER home team blanket trade-off study. The objective was to compare the activity, afterheat and waste disposal rating (WDR) results of a composite blanket-shield design for the continuous operation approximation to a pulsed operation case to determine whether the differences are at most the duty factor as predicted by the two nuclide chain model. Up to a cooling period of 100 years, the pulsed activity and afterheat values were below the continuous oepration results and well within (except for one afterheat value) the maximum deviation predicted by the two nuclide chain model. No differences in the WDR values were noted as they are, to a large extent, based on long-lived nuclides which are insensitive to short-term changes in the operation history. (orig.)

  8. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Yun, Sunghwan; Kim, Sang Ji

    2015-01-01

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

  9. Operability design review of prototype large breeder reactor (PLBR) designs. Final report, September 1981

    International Nuclear Information System (INIS)

    Beakes, J.H.; Ehman, J.R.; Jones, H.M.; Kinne, B.V.T.; Price, C.M.; Shores, S.P.; Welch, J.K.

    1981-09-01

    Prototype Large Breeder Reactor (PLBR) designs were reviewed by personnel with extensive power plant operations experience. Fourteen normal and off-normal events, such as startup, shutdown, refueling, reactor scram and loss of feedwater, were evaluated using an operational evaluation methodology which is designed to facilitate talk-through sessions on operational events. Human factors engineers participated in the review and assisted in developing and refining the review methodologies. Operating experience at breeder reactor facilities such as Experimental Breeder Reactor-II (EBR-II), Enrico Fermi Atomic Power Plant - Unit 1, and the Fast Flux Test Facility (FFTF) was gathered, analyzed, and used to determine whether lessons learned from operational experience had been incorporated into the PLBR designs. This eighteen month effort resulted in approximately one hundred specific recommendations for improving the operability of PLBR designs

  10. Magnetoconvection in HCLL blankets

    International Nuclear Information System (INIS)

    Mistrangelo, C.; Buehler, L.

    2014-01-01

    In the present work we consider magneto-convective flows in one of the proposed European liquid metal blankets that will be tested in the experimental fusion reactor ITER. Here the PbLi alloy is used as breeder material and helium as coolant. In order to finalize the design of the helium cooled lead lithium (HCLL) blanket, studies are still required to fully understand the behavior of the electrically conducting breeder under the influence of the intense magnetic field that confines the fusion plasma and in case of non-uniform thermal conditions. Liquid metal HCLL blanket flows are expected to be mainly driven by buoyancy forces caused by non-isothermal operating conditions due to neutron volumetric heating and cooling of walls, since only a weak forced ow is foreseen for tritium extraction in external ancillary systems. Buoyancy can therefore become very important and modify the velocity distribution and related heat transfer performance of the blanket. The present numerical study aims at clarifying the influence of electromagnetic and thermal coupling of neighboring fluid domains on magneto-convective flows in geometries relevant for the HCLL blanket concept. According to the last design review two internal cooling plates subdivide the fluid domain into three slender flow regions, which are thermally and electrically coupled through common walls. First a uniform volumetric heat source is considered to identify the basic convective patterns that establish in the liquid metal. Results are then compared with those obtained by applying a realistic radial distribution of the power density as obtained from a neutronic analysis. Velocity and temperature distributions are discussed for various volumetric heat sources and magnetic field strengths.

  11. Structural design study of tritium breeding blanket with a lead layer as a neutron multiplier

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Kitamura, Kazunori; Minato, Akio; Sakamoto, Hiroki; Yamamoto, Takashi

    1980-12-01

    Thermal and structural design study of a tritium breeding blanket with a lead layer for a International Tokamak Reactor (INTOR) is carried out. Tube in shell type blanket with a lead layer is found to be promising. The volume fraction of structural material in the lead layer can be small enough to keep the neutron multiplication effect of lead. Reasonable value of shell effect is attainable due to lead layer in the front part of the blanket. (author)

  12. Mechanical design

    Science.gov (United States)

    1976-01-01

    Design concepts for a 1000 mw thermal stationary power plant employing the UF6 fueled gas core breeder reactor are examined. Three design combinations-gaseous UF6 core with a solid matrix blanket, gaseous UF6 core with a liquid blanket, and gaseous UF6 core with a circulating blanket were considered. Results show the gaseous UF6 core with a circulating blanket was best suited to the power plant concept.

  13. Overview of pool hydraulic design of Indian prototype fast breeder ...

    Indian Academy of Sciences (India)

    Flow sheet of prototype fast breeder reactor. ... over, the main vessel that houses radioactive primary sodium is free of any ..... with superficial velocity components in porous media. ..... The attenuation within thermal boundary layer was found.

  14. Tokamak blanket design study: FY 78 summary report

    International Nuclear Information System (INIS)

    1979-06-01

    A tokamak blanket cylindrical module concept was designed, developed, and analyzed after review of several existing generic concepts. The design is based on use of state-of-the-art structural materials (20% cold worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders and features direct wall cooling by flowing helium between the outer (first wall) cylinder and the inner lithium containing cylinder. Each cylinder is capable of withstanding full coolant pressure for enhanced reliability. Results show that stainless steel is a viable material for a first wall subjected to 4 MW/m 2 neutron and 1 MW/m 2 particle heat flux. A lifetime analysis showed that the first wall design meets the goal of operating at 20 minute cycles with 95% duty for 10 5 cycles. The design is attractive for further development, and additional work and supporting experiments are identified to reduce analytical uncertainties and enhance the design reliability

  15. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Waganer, L.M.

    1985-01-01

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

  16. Design of self-cooled, liquid-metal blankets for tokamak and tandem mirror reactors

    International Nuclear Information System (INIS)

    Cha, Y.S.; Gohar, Y.; Hassanein, A.M.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.; Szo, D.K.

    1985-01-01

    Results of the self-cooled, liquid-metal blanket design from the Blanket Comparison and Selection Study (BCSS) are summarized. The objectives of the BCSS project are to define a small number (about three) of blanket concepts that should be the focus of the blanket research and development (RandD) program, identify and prioritize the critical issues for the leading blanket concepts, and provide technical input necessary to develop a blanket RandD program plan. Two liquid metals (lithium and lithium-lead (17Li-83Pb)) and three structural materials (primary candidate alloy (PCA), ferritic steel (FS) (HT-9), and vanadium alloy (V-15 Cr-5 Ti)) are included in the evaluations for both tokamaks and tandem mirror reactors (TMRs). TMR is of the tube configuration similar to the Mirror Advanced Reactor Study design. Analyses were performed in the following generic areas for each blanket concept: MHD, thermal hydraulics, stress, neutronics, and tritium recovery. Integral analyses were performed to determine the design window for each blanket design. The Li/Li/V blanket for tokamak and the Li/Li/V, LiPb/LiPb/V, and Li/Li/HT-9 blankets for the TMR are judged to be top-rated concepts. Because of its better thermophysical properties and more uniform nuclear heating profile, liquid lithium is a better coolant than liquid 17Li83Pb. From an engineering point of view, vanadium alloy is a better structural material than either FS or PCA since the former has both a higher allowable structural temperature and a higher allowable coolant/structure interface temperature than the latter. Critical feasibility issues and design constraints for the self-cooled, liquid-metal blanket concepts are identified and discussed

  17. Fast breeder reactor-block antiseismic design and verification

    International Nuclear Information System (INIS)

    Martelli, A.; Forni, M.

    1988-02-01

    The Specialists' Meeting on ''Fast Breeder Reactor-Block Antiseismic Design and Verification'' was organized by the ENEA Fast Reactor Department in co-operation with the International Working Group (IWGFR) of the International Atomic Energy Agency (IAEA), according to the recommendations of the 19th IAEA/IWGFR Meeting. It was held in Bologna, at the Headquarters of the ENEA Fast Reactor Department, on October 12-15, 1987, in the framework of the Celebrations for the Ninth Centenary of the Bologna University. The proceedings of the meeting consists of three parts. Part 1 contains the introduction and general comments, the agenda of the meeting, session summaries, conclusions and recommendations and the list of participants. Part 2 contains 8 status reports of Member States participating in the Working Group. Contributed papers were published in Part 3 and were further subdivided into 5 sessions as follows: whole reactor-block analysis (4 papers); whole reactor-block analysis (sloshing and buckling, seismic isolation effects) (8 papers); detailed core analysis (6 papers); shutdown systems and core structural and functional verifications (6 papers); component and piping analysis (7 papers). A separate abstract was prepared for each of the 8 status reports and 31 contributed papers. Refs, figs and tabs

  18. Study on advanced structural design for commercialized fast breeder reactors

    International Nuclear Information System (INIS)

    Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai

    2002-01-01

    Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) launched joint research programs on structural design and three-dimensional seismic isolation technologies, as part of the supporting R and D activities for the feasibility studies on commercialized fast breeder reactor cycle systems. A research project by JAPC under the auspices of the Ministry of Economy, Trade, and Industry (METI) with technical support by JNC is included in this joint study. This report contains the results of the research on the structural design technology. The research scope was identified as (1) System Based Code for Integrity, (2) FDS (FBR Design Standard), and (3) Standardization of new material, and the results of this year's studies are summarized as follows. (1) System Based Code for Integrity. Requirements that a structural design standard must fulfill for enhanced reliability and economy were clarified. Based on this, the authors began to develop the system-based code for integrity. A structural reliability based design approach was proposed as a basic concept for an integrated evaluation of structural integrity. A system consisting of a supreme code and partial codes was proposed. Technologies and engineering tools that are necessary to materialize this code were clarified and research and development was begun. (2) FDS(FBR Design Standard). A rational design approach against thermal loads was proposed. Applicable area of inelastic analysis methods was investigated to develop inelastic analysis guide. A new design system which realizes feedback from structural to thermal hydraulic designs was proposed with a total analysis method of thermal hydraulic and mechanical behaviors. (3) Standardization of new material. Current status of development of high-chromium ferritic steels was investigated. Those steels have excellent high temperature strength and thermal properties. The authors proposed material specifications to apply those steels to structures

  19. Study on advanced structural design for commercialized fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Sagayama, Yutaka; Dozaki, Koji; Tanaka, Yoshihiko [Japan Atomic Power Co., Research and Development Department, Tokyo (Japan)

    2002-01-01

    Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) launched joint research programs on structural design and three-dimensional seismic isolation technologies, as part of the supporting R and D activities for the feasibility studies on commercialized fast breeder reactor cycle systems. A research project by JAPC under the auspices of the Ministry of Economy, Trade, and Industry (METI) with technical support by JNC is included in this joint study. This report contains the results of the research on the structural design technology. The research scope was identified as (1) System Based Code for Integrity, (2) FDS (FBR Design Standard), and (3) Standardization of new material, and the results of this year's studies are summarized as follows. (1) System Based Code for Integrity. Requirements that a structural design standard must fulfill for enhanced reliability and economy were clarified. Based on this, the authors began to develop the system-based code for integrity. A structural reliability based design approach was proposed as a basic concept for an integrated evaluation of structural integrity. A system consisting of a supreme code and partial codes was proposed. Technologies and engineering tools that are necessary to materialize this code were clarified and research and development was begun. (2) FDS(FBR Design Standard). A rational design approach against thermal loads was proposed. Applicable area of inelastic analysis methods was investigated to develop inelastic analysis guide. A new design system which realizes feedback from structural to thermal hydraulic designs was proposed with a total analysis method of thermal hydraulic and mechanical behaviors. (3) Standardization of new material. Current status of development of high-chromium ferritic steels was investigated. Those steels have excellent high temperature strength and thermal properties. The authors proposed material specifications to apply those steels to

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

    International Nuclear Information System (INIS)

    Bond, A.

    1984-01-01

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

  1. Neutronic studies of a 233U breeder

    International Nuclear Information System (INIS)

    Hansen, L.F.; Maniscalco, J.A.

    1978-09-01

    Neutronic calculations have been carried out to design a laser fusion driven hybrid blanket which maximizes 233 U production per unit of thermal energy (>1 kg/MW/sub T/-year) with acceptable fusion energy multiplication (M/sub F/ approx. 4). Two hybrid blankets, a thorium and a uranium--thorium blanket, are discussed in detail and their performance is evaluated by incorporating them into an existing hybrid design (the LLL/Bechtel design). The performance of these two blankets is discussed in terms of their energy multiplication, tritium breeding and fissile fuel production. The neutronic calculations have been done for two neutron libraries, the ENDF/B-IV and the ENDL with differences no larger than 10% in the results. An estimate is given of the number of equivalent thermal power fission reactors (LWR, HWR, SSCR, and HTGR) that these fusion breeders can fuel

  2. Study on advanced structural design for commercialized fast breeder reactors

    International Nuclear Information System (INIS)

    Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai

    2003-05-01

    Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) launched joint research programs on structural design and three-dimensional seismic isolation technologies, as part of the supporting R and D activities for the feasibility studies on commercialized fast breeder reactor cycle systems. A research project by JAPC under the auspices of the Ministry of Economy, Trade, and Industry (METI) with technical support by JNC is included in this joint study. This report contains the results of the research on the structural design technology. The research scope was identified as (1) FDS (FBR Design Standard), (2) Standardization of new material, and (3) System Based Code for Integrity, and the results of this year's studies are summarized as follows. (1) FDS (FBR Design Standard). R and D policy of Phase II study considering to the newest needs demanded for structural design were clarified, and R and D items were settled concretely. As for failure criteria, preliminary ratcheting-fatigue tests were conducted with rational methods, and Negligible Creep curves that settle creep designing region rationally were expanded their applying condition. R and D policy and items of guideline for inelastic analysis were clarified, and analyzing methodologies were studied. Summering up exemplification of thermal load prediction methodologies were progressed. To predict thermal striping loads, advanced frequency response function of thermal stress, and fatigue evaluation methodology were studied. (2) Standardization of new material. As for candidate 12-chromium stainless steel (added tungsten, non-added tungsten), that is expected to improve strength of components of commercialized FR, short and medium-term material tests, and creep-fatigue tests at small strain range were conducted. As for above candidate steel, TIG (Tungsten Inert Gas) welding method was recommended. Requirements to standardize new materials and index to select new materials were studied

  3. Final report for fuel acquisition and design of a fast subcritical blanket facility

    International Nuclear Information System (INIS)

    Clikeman, F.M.; Ott, K.O.

    1976-01-01

    A summary is presented of work leading to the design of a subcritical facility for the study of fast reactor blankets. Included are activities related to fuel acquisition, design of the facility, and experiment planning

  4. Lithium ceramics as the solid breeder material in fusion reactors

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Reuther, T.C.; Johnson, C.E.

    1982-03-01

    Fusion blanket designs have for almost a decade considered the use of a solid breeder relying on available data and assumed performance. The conclusion from these studies is that acceptable neutronic and thermal hydraulic performance can be achieved. In the future, it will be necessary to establish that a particular material can tolerate the thermal and irradiation environment of the fusion blanket while still providing the required functions of tritium recovery, power production and neutron shielding

  5. Assessment of alternative vessel and blanket design on ITER operation

    Energy Technology Data Exchange (ETDEWEB)

    Cavinato, M., E-mail: mario.cavinato@f4e.europa.e [FUSION FOR ENERGY Joint Undertaking, 08019 Barcelona (Spain); Portone, A.; Saibene, G.; Sartori, R. [FUSION FOR ENERGY Joint Undertaking, 08019 Barcelona (Spain); Albanese, R.; Ambrosino, G.; Ariola, M. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Napoli (Italy); Artaserse, G. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Reggio Calabria (Italy); Mattei, M. [Associazione Euratom-ENEA-CREATE, DIAM, Seconda Universita di Napoli, Via Roma 29, Aversa, CE 81031 Italy (Italy); Pironti, A. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Napoli (Italy); Villone, F. [Associazione Euratom-ENEA-CREATE, DIMET, Universita degli Studi di Cassino (Italy)

    2010-12-15

    In the framework of the ITER project, an investigation has been conducted on an alternative vessel and blanket design, aimed at reducing cost and production risk. The modifications proposed have a strong impact on plasma control since they affect the main conducting structures surrounding the plasma column, providing passive stabilization but at the same time shielding the field generated by the active coils to control the plasma motion and shape. An extensive analysis was performed to assess the plasma vertical controllability and the modified requirements to the in-vessel vertical stability coils system as well as to the external Poloidal Field coils system. A similar analysis was aimed at assessing the performance of the shape control system in presence of the modified structures. The effect on plasma breakdown was also evaluated in terms of maximum initial loop voltage, quality of magnetic null and the flux loss related to the breakdown delay that was quantified under the same hypothesis employed by ITER for the baseline design. Furthermore, the modified design presents issues for the magnetic diagnostic system, related to the shielding of the probes by the eddy currents, which were analysed with a 3D model. The results of the analyses performed have some general interest in particular regarding the influence on plasma stability of 3D structures with close proximity to the plasma. The present paper aims at giving an overview of the analyses that have been carried out and a summary of the results in terms of impact of the modified design on plasma control and scenario, and in general an evaluation of the role of passive structure in plasma vertical stability and shape control.

  6. Comparison study on neutronic analysis of the K-DEMO water cooled ceramic breeder blanket using MCNP and ATTILA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Sung, E-mail: jspark@nfri.re.kr; Kwon, Sungjin; Im, Kihak

    2016-11-01

    Highlights: • A comparison study of main parameter calculations: neutron wall loading (NWL), tritium breeding ratio (TBR), and nuclear heating, on a Korean fusion demonstration reactor (K-DEMO) neutronic analysis model using MCNP and ATTILA was performed to investigate the feasibility of using ATTILA. • The calculation results of this study indicates that ATTILA showed close agreement with MCNP within ranges (3.3–28%). • Partly high discrepancy (17–28%) results between two codes existed to the nuclear heating calculation in high attenuating materials and radially thick structure regions. • The rest of the results showed small differences of NWL calculation (3.3%) and TBR distribution (3.9%). • ATTILA could be acceptable for K-DEMO neutronic analysis considering discrepancy (3.3–28%). - Abstract: A comparison study of main parameter calculations: neutron wall loading (NWL), tritium breeding ratio (TBR), and nuclear heating, on a Korean fusion demonstration reactor (K-DEMO) neutronic analysis model using MCNP and ATTILA was performed to investigate the feasibility of using ATTILA for the main parameter calculations. The model was created by commercial CAD program (Pro-Engineer™) as a 22.5° sector of tokamak consisting of major components such as blankets, shields, divertors, vacuum vessels (VV), toroidal field (TF) coils, and others, which was directly imported into ATTILA by Parasolid file. The discretizing in space, angle, and energy variables were refined for application of the K-DEMO neutronic analysis model through an iterative process since these variables greatly impact on accuracy, solution times, and memory consumptions in ATTILA. The main parameter calculations using ATTILA and the result of comparison studies indicate that the NWL distributions by two codes were almost agreed within discrepancy of 3.3%; the TBR distribution using ATTILA was slightly bigger than MCNP with a difference 3.9%; the nuclear heating values on TF coils and VV

  7. Fission-suppressed hybrid reactor: the fusion breeder

    International Nuclear Information System (INIS)

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

    1982-12-01

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

  8. An aqueous lithium salt blanket option for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Steiner, D.; Varsamis, G. (Rensselaer Polytechnic Inst., Troy, NY (USA). Dept. of Nuclear Engineering and Engineering Physics); Deutsch, L.; Rathke, J. (Grumman Corp., Bethpage, NY (USA). Advanced Energy Systems); Gierszewski, P. (Canadian Fusion Fuels Technology Project (CFFTP), Mississauga, ON (Canada))

    1989-04-01

    An aqueous lithium salt blanket (ALSB) concept is proposed which could be the basis for either a power reactor blanket or a test module in an engineering test reactor. The design is based on an austenitic stainless steel structure, a beryllium multiplier, and a salt breeder concentration of about 32 g LiNO/sub 3/ per 100 cm/sup 3/ of H/sub 2/O. To limit tritium release rates, the salt breeder solution is separated from the water coolant circuit. The overall tritium system cost for a 2400 MW (fusion power) reactor is estimated to be 180 million Dollar US87 installed. (orig.).

  9. Thermal-hydraulic analysis of low activity fusion blanket designs

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.; Yu, W.S.

    1977-01-01

    The heat transfer aspects of fusion blankets are considered where: (a) conduction and (b) boiling and condensation are the dominant heat transfer mechanisms. In some cases, unique heat transfer problems arise and additional heat transfer data and analyses may be required

  10. Neutronics analysis for aqueous self-cooled fusion reactor blankets

    International Nuclear Information System (INIS)

    Varsamis, G.; Embrechts, M.J.; Jaffa, R.; Steiner, D.; Deutsch, L.; Gierszewski, P.

    1986-06-01

    The tritium breeding performance of several Aqueous Self-Cooled Blanket (ASCB) configurations for fusion reactors has been evaluated. The ASCB concept employs small amounts of lithium compound dissolved in light or heavy water to serve as both coolant and breeding medium. The inherent simplicity of this concept allows the development of blankets with minimal technological risk. The tritium breeding performance of the ASCB concept is a critical issue for this family of blankets. Contrary to conventional blanket designs there will be a significant contribution to the tritium breeding ratio (TBR) in the water coolant/breeder of duct shields, and the 3-D TBR will therefore be similar to the 1-D TBR. The tritium breeding performance of an ASCB for a MARS-like-tandem reactor and an ASCB based breeding-shield for the Next European Torus (NET) are assessed. Two design options for the MARS-like blanket are discussed. One design employs a vanadium first wall, and zircaloy for the structural material. The trade-offs between light water and heavy water cooling options for this zircaloy blanket are discussed. The second design option for MARS relies on the use of a vanadium alloy as the stuctural material, and heavy water as the coolant. It is demonstrated that both design options lead to low-activation blankets that allow class C burial. The breeder-shield for NET consists of a water-cooled stainless steel shield

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

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-01-01

    A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of this study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. The primary results of the study are as follows: a) the lithium-lead blanket achieves a higher TBR with a smaller blanket thickness relative to the lithium blanket; b) the lithium blanket generates more energy per fusion neutron relative to the lithium-lead blanket; c) among the possible reflector materials, the carbon reflector produces the highest TBR; d) the high-Z reflector materials (Mo, Cu, W, or steel) generate more energy per fusion neutron and produce smaller TBRs relative to the carbon reflector; e) lithium-6 enrichment is required for the lithium-lead blanket to reduce the total blanket thickness; and f) the energy deposition per fusion neutron reaches a saturation as the blanket thickness, the fraction of the high-Z material in the reflector, or the reflector zone thickness increases (this allows one to design the blanket for a specific TBR without reducing the energy production)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-08-01

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

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

    International Nuclear Information System (INIS)

    Ulrickson, M.A.; Manly, W.D.; Dombrowski, D.E.

    1995-08-01

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

  14. Materials issues in the design of the ITER first wall, blanket, and divertor

    International Nuclear Information System (INIS)

    Mattas, R.F.; Smith, D.L.; Wu, C.H.; Shatalov, G.

    1992-01-01

    During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R ampersand D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented

  15. Thermal-hydraulics design comparisons for the tandem mirror hybrid reactor blanket

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Yang, Y.S.; Schultz, K.R.

    1980-09-01

    The Tandem Mirror Hybrid Reactor (TMHR) is a cylindrical reactor, and the fertile materials and tritium breeding fuel elements can be arranged with radial or axial orientation in the blanket module. Thermal-hydraulics performance comparisons were made between plate, axial rod and radial rod fuel geometrices. The three configurations result in different coolant/void fractions and different clad/structure fractions. The higher void fraction in the two rod designs means that these blankets will have to be thicker than the plate design blanket in order to achieve the same level of nuclear interactions. Their higher structural fractions will degrade the uranium breeding ratio and energy multiplication factor of the design. One difficulty in the thermal-hydraulics analysis of the plate design was caused by the varying energy multiplication of the blanket during the lifetime of the plate which forced the use of designs that operated in the transition flow regime at some point during life. To account for this, an approach was adopted from Gas Cooled Fast Reactor (GCFR) experience for the pressure drop calculation and the corresponding heat transfer coefficient that was used for the film drop thermal calculation. Because of the superior nuclear performance, the acceptable thermal-hydraulic characteristics and the mechanical design feasibility, the plate geometry concept was chosen for the reference gas-cooled TMHR blanket design

  16. The EC conceptual design proposal of a water-cooled convertible blanket for ITER

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Baraer, L.; Bielak, B.; Raepsaet, X.; Salavy, J.F.; Sedano, L.; Szczepanski, J.; Quintric-Bossy, J.; Severi, Y.

    1993-01-01

    For several years the EC laboratories have developed breeding blankets for DEMO. From this experience, it has been derived a proposal of tritium breeding blanket for the Extended Performance Phase (EPP) of ITER. The general basic ideas are the following: (i) the switch from the shielding blanket used during the BPP to the breeding blanket for the EPP should not require segments replacement ('convertible' blanket): (ii) its use should not have significant impact on the Basic Performance Phase (BPP); (iii) design and used materials should assure good safety standards and acceptable public perception; (iv) the blanket coolant should be compatible with the coolant required in the high heat-flux components (e.g. divertor, etc.; (v) the required R and D should fit with the ITER time schedule; (vi) the blanket should be able to withstand large power excursions and to accept long downtimes. The proposed design consists of a water-cooled liquid metal blanket, using the eutectic Pb-17Li during the EPP and a non-breeding Pb-alloy (Pb-18Mg or Pb-50Bi) during the BPP. Each segment is basically formed by a box containing the alloy, cooled by an array of poloidal hairpin-type cooling tubes and reinforced by toroidal and radial stiffeners. The coolant tubes are double-walled tubes allowing leak detections. The selected First Wall (FW) is a toroidally-drilled steel plate with brazed water-cooling U-tube. The structural material is austenitic stainless steel (316L(N)) which limits the maximum acceptable neutron fluence to about 1 MWa/m 2 . The advantages of using other structural materials requiring longer leadtimes, such as ferritic/martensitic steels, are also briefly discussed

  17. Thermostructural design of the first wall/blanket for the TITAN-RFP fusion reactor

    International Nuclear Information System (INIS)

    Orient, G.E.; Blanchard, J.P.; Ghoniem, N.M.

    1987-01-01

    The mass power density, which is defined as the average power per unit mass within the magnet boundary, is a rough and general measure of economic competitiveness. Conn et al. (1985) have identified a target value of 100 kW(e)/tonne as a reasonable threshold for 'compact' commercial fusion systems. In pursuit of this goal, Hagenson et al. (1984) and Najmabadi et al. (1987) have pointed out the inherent characteristics of the RFP toroidal confinement concept which allow it to exceed this target value. It is inevitable that the compactness of the fusion power core will introduce a unique set of design issues. The special design concerns stem from high thermal surface fluxes, high bulk energy deposition by neutrons, and a relatively short blanket structural lifetime. In the TITAN-RFP, study Najmabadi et al. (1987) investigate a number of blanket (B) and first wall (FW) options suitable for high power density fusion reactors. Final choices were made for two designs: A high pressure aqueous blanket and a vanadium/lithium self-cooled blanket. The first design utilizes a pressurized aqueous loop containing a lithium compound dissolved in water, while the second design is based upon a self-cooled lithium-vanadium blanket. In this paper, we consider the beginning-of-life (BOL) thermostructural design and analysis of only the second concept. (orig./GL)

  18. Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

    International Nuclear Information System (INIS)

    Li, Jia; Jiang, Kecheng; Zhang, Xiaokang; Nie, Xingchen; Zhu, Qinjun; Liu, Songlin

    2016-01-01

    Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

  19. Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jia, E-mail: lijia@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Jiang, Kecheng; Zhang, Xiaokang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China); Nie, Xingchen [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Zhu, Qinjun; Liu, Songlin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China)

    2016-12-15

    Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

  20. Neutronic performance optimization study of Indian fusion demo reactor first wall and breeding blanket

    International Nuclear Information System (INIS)

    Swami, H.L.; Danani, C.

    2015-01-01

    In frame of design studies of Indian Nuclear Fusion DEMO Reactor, neutronic performance optimization of first wall and breeding blanket are carried out. The study mainly focuses on tritium breeding ratio (TBR) and power density responses estimation of breeding blanket. Apart from neutronic efficiency of existing breeding blanket concepts for Indian DEMO i.e. lead lithium ceramic breeder and helium cooled solid breeder concept other concepts like helium cooled lead lithium and helium-cooled Li_8PbO_6 with reflector are also explored. The aim of study is to establish a neutronically efficient breeding blanket concept for DEMO. Effect of first wall materials and thickness on breeding blanket neutronic performance is also evaluated. For this study 1 D cylindrical neutronic model of DEMO has been constructed according to the preliminary radial build up of Indian DEMO. The assessment is being done using Monte Carlo based radiation transport code and nuclear cross section data file ENDF/B- VII. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

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

    Science.gov (United States)

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

    1998-10-01

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

  3. Tritium transport analysis for CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Yang, Wanli; Li, Yuanjie; Ge, Zhihao; Nie, Xingchen; Gao, Zhongping

    2017-01-15

    Highlights: • A simplified tritium transport model for CFETR WCSB blanket was developed. • Tritium transport process in CFETR WCSB blanket was analyzed. • Sensitivity analyses of tritium transport parameters were carried out. - Abstract: Water Cooled Solid Breeder (WCSB) blanket was put forward as one of the breeding blanket candidate schemes for Chinese Fusion Engineering Test Reactor (CFETR). In this study, a simplified tritium transport model was developed. Based on the conceptual engineering design, neutronics and thermal-hydraulic analyses of CFETR WCSB blanket, tritium transport process was analyzed. The results show that high tritium concentration and inventory exist in primary water loop and total tritium losses exceed CFETR limits under current conditions. Conducted were sensitivity analyses of influential parameters, including tritium source, temperature, flow-rate capacity and surface condition. Tritium performance of WCSB blanket can be significantly improved under a smaller tritium impinging rate, a larger flow-rate capacity or a better surface condition. This work provides valuable reference for the enhancement of tritium transport behavior in CFETR WCSB blanket.

  4. Progress in fusion reactors blanket analysis and evaluation at CEA

    International Nuclear Information System (INIS)

    Proust, E.; Gervaise, F.; Carre, F.; Chevereau, G.; Doutriaux, D.

    1986-09-01

    In the frame of the recent CEA studies aiming at the development, evaluation and comparison of solid breeder blanket concepts in view of their adaptation to NET, the evaluation of specific questions related to the first wall design, the present paper examines first the performances of a helium cooled toroidal blanket design for NET, based on innovative Beryllium/Ceramics breeder rod elements. Neutronic and thermo-mechanical optimisation converges on a concept featured by a breeding capability in excess of 1.2, a reasonnable pumping power of 1% and a narrow breeder temperature range (470+-30 deg C of the breeder), the latter being largely independent of the power level. This design proves naturally adapted to ceramic breeder assigned to very strict working conditions, and provides for any change in the thermal and heat transfer characteristics over the blanket lifetime. The final section of the paper is devoted to the evaluation of the heat load poloidal distribution and to the irradiation effects on first wall structural materials

  5. Thermomechanical analysis of solid breeders in sphere-pac, plate, and pellet configurations

    International Nuclear Information System (INIS)

    Blanchard, J.P.; Ghoniem, N.M.

    1986-02-01

    The first configuration studied is called sphere-pac. It features small breeder spheres of three different diameters, thus allowing efficient packing and minimal void fraction. The concept originated as an attempt to minimize thermal stresses in the breeder and improve the predictability of the breeder-structure interface heat conduction. In general the breeder is made as thin as possible, to maximize the breeding ratio, so the cladding's integrity will likely be the life-limiting issue of this concept. The third breeder configuration is in the form of pellets cladded by steel tubes. The major thermomechanical issue of the pin-type designs is cracking, which would impair the thermal performance of the blanket. Fortunately, the pins can be sized to prevent cracking under normal operation. In this report we have treated each blanket generically, dealing with basic issues rather than design specifics. Our basic philosophy is to avoid cracking of the breeder if at all possible. It can be argued that cracking could be allowed, but this would sacrifice predictability of the blanket thermal performance and tritium release characteristics. Proper design can and should minimize breeder cracking

  6. Liquid metal blanket module testing and design for ITER/TIBER II

    International Nuclear Information System (INIS)

    Mattas, R.F.; Cha, Y.; Finn, P.A.; Majumdar, S.; Picologlou, B.; Stevens, H.; Turner, L.

    1988-05-01

    A major goal for ITER is the testing of nuclear components to demonstrate the integrated performance of the most attractive concepts that can lead to a commercial fusion reactor. As part of the ITER/TIBER II study, the test program and design of test models were examined for a number of blanket concepts. The work at Argonne National Laboratory focused on self-cooled liquid metal blankets. A test program for liquid metal blankets was developed based upon the ITER/TIBER II operating schedule and the specific data needs to resolve the key issues for liquid metals. Testing can begin early in reactor operation with liquid metal MHD tests to confirm predictive capability. Combined heat transfer/MHD tests can be performed during initial plasma operation. After acceptable heat transfer performance is verified, tests to determine the integrated high temperature performance in a neutron environment can begin. During the high availability phase operation, long term performance and reliability tests will be performed. It is envisioned that a companion test program will be conducted outside ITER to determine behavior under severe accident conditions and upper performance limits. A detailed design of a liquid metal test module and auxiliary equipment was also developed. The module followed the design of the TPSS blanket. Detailed analysis of the heat transfer and tritium systems were performed, and the overall layout of the systems was determined. In general, the blanket module appears to be capable of addressing most of the testing needs. 8 refs., 27 figs., 11 tabs

  7. Self-cooled liquid-metal blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Arheidt, K.; Barleon, L.

    1988-01-01

    A blanket concept for the Next European Torus (NET) where 83Pb-17Li serves both as breeder material and as coolant is described. The concept is based on the use of novel flow channel inserts for a decisive reduction of the magnetohydrodynamic (MHD) pressure drop and employs beryllium as neutron multiplier in order to avoid the need for breeding blankets at the inboard side of the torus. This study includes the design, neutronics, thermal hydraulics, stresses, MHDs, corrosion, tritium recovery, and safety of a self-cooled liquid-metal blanket. The results of the investigations indicate that the self-cooled blanket is an attractive alternative to other driver blanket concepts for NET and that it can be extrapolated to the conditions of a DEMO reactor

  8. Neutronic Parametric Study on a Conceptual Design for a Transmutation Fusion Blanket

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2011-01-01

    Fusion energy may be the one of options of future energy. In all over the world, researchers are putting their efforts for its commercial and economical availability. Fusion-fission hybrid reactors have been studied for various applications in China. First milestone of fusion energy is expected to be the fusion fission hybrid reactors. In fusion-fission hybrid reactor the blanket design is of second prime importance after fusion source. In this study conceptual design of a fusion blanket is initiated for calculation of tritium production, transmutation of minor actinides (MA) and fission products (FP) and energy multiplication calculations

  9. Low activity blanket designs and heat transfer for experimental power reactors

    International Nuclear Information System (INIS)

    Fillo, J.; Tichler, P.; Lazareth, O.; Powell, J.

    1976-01-01

    Two minimum activity blanket designs are described, based on the ANL TEPR circular design parameters. A first wall loading (plasma on) of 1.0 MW(th)/m 2 has been assumed. The first option is composed of SAP (sintered aluminum product) modules. The oval shaped SAP shell, in which approximately 45 percent of the fusion energy is removed, is maintained at a temperature of approximately 400 0 C by a He coolant stream. The remaining 55 percent of the fusion energy is deposited in a thermally insulated hot interior (SiC and B 4 C) and removed by a separate He coolant, with exit temperature of 800 0 C. In the second option, the blanket is a thick graphite block structure (approximately 50 cm thickness) with SAP coolant tubes carrying He (50 atm) embedded deep within the graphite to minimize radiation damage. The neutron and gamma energy deposited in the graphite is radiated along internal slots and conducted through the graphite to the coolant tubes. To reduce surface evaporation above 2000 0 C, the blanket surface is radiatively cooled to a low temperature radiation sink, a bank of He cooled SAP tubes. Approximately 20 percent of the fusion energy is removed in this region, the remaining 80 percent in the primary graphite-aluminum blanket. Both blanket options are mounted on heavy Al backing plates, cooled by He, which are in turn supported from the fixed shield

  10. Solid tritium breeder materials-Li2O and LiAlO2: a data base review

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Billone, M.C.; Clemmer, R.G.; Fischer, A.K.; Hollenberg, G.W.; Tam, S.W.

    1985-01-01

    The fabrication, properties, and irradiation behavior of Li 2 O and γ-LiAlO 2 are reviewed and assessed to determine the potential of these materials to satisfy the basic solid breeder blanket performance requirements. Based on the data analysis and theoretical modeling, a set of major technical uncertainties is identified. These uncertainties include: fabricability of sphere-pac solid breeders; high fluence and burnup effects on thermal conductivity and microstructural stability; high fluence and burnup effects on tritium diffusion coefficients at low temperature; relationship among purge flow chemistry, surface adsorption, and species of released tritium; and mechanical properties and the loads imposed on the structural materials by the breeder during blanket operation. Resolution of these issues is important in assuring that solid breeder blankets can be designed with confidence

  11. Thermal conductivity of fusion solid breeder materials

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tam, S.W.

    1986-06-01

    Several simple and useful formulae for estimating the thermal conductivity of lithium-containing ceramic tritium breeder materials for fusion reactor blankets are given. These formulae account for the effects of irradiation, as well as solid breeder configuration, i.e., monolith or a packed bed. In the latter case, a coated-sphere concept is found more attractive in incorporating beryllia (a neutron multiplier) into the blanket than a random mixture of solid breeder and beryllia spheres

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

    International Nuclear Information System (INIS)

    Raffray, A.R.; Federici, G.; Billone, M.C.; Tanaka, S.

    1994-01-01

    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

  13. Fusion Breeder Program interim report

    International Nuclear Information System (INIS)

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

    1982-01-01

    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

  14. Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor

    Directory of Open Access Journals (Sweden)

    Bong Guen Hong

    2018-02-01

    Full Text Available A configuration of a fusion-driven transmutation reactor with a low aspect ratio tokamak-type neutron source was determined in a self-consistent manner by using coupled analysis of tokamak systems and neutron transport. We investigated the impact of blanket configuration on the characteristics of a fusion-driven transmutation reactor. It was shown that by merging the TRU burning blanket and tritium breeding blanket, which uses PbLi as the tritium breeding material and as coolant, effective transmutation is possible. The TRU transmutation capability can be improved with a reduced blanket thickness, and fast fluence at the first wall can be reduced.  Article History: Received: July 10th 2017; Received: Dec 17th 2017; Accepted: February 2nd 2018; Available online How to Cite This Article: Hong, B.G. (2018 Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor. International Journal of Renewable Energy Development, 7(1, 65-70. https://doi.org/10.14710/ijred.7.1.65-70

  15. Simulation and design of an electron beam ion source charge breeder for the californium rare isotope breeder upgrade

    Directory of Open Access Journals (Sweden)

    Clayton Dickerson

    2013-02-01

    Full Text Available An electron beam ion source (EBIS will be constructed and used to charge breed ions from the californium rare isotope breeder upgrade (CARIBU for postacceleration into the Argonne tandem linear accelerator system (ATLAS. Simulations of the EBIS charge breeder performance and the related ion transport systems are reported. Propagation of the electron beam through the EBIS was verified, and the anticipated incident power density within the electron collector was identified. The full normalized acceptance of the charge breeder with a 2 A electron beam, 0.024π  mm mrad for nominal operating parameters, was determined by simulating ion injection into the EBIS. The optics of the ion transport lines were carefully optimized to achieve well-matched ion injection, to minimize emittance growth of the injected and extracted ion beams, and to enable adequate testing of the charge bred ions prior to installation in ATLAS.

  16. Study on compact design of remote handling equipment for ITER blanket maintenance

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi

    2006-03-01

    In the ITER, the neutrons created by D-T reactions activate structural materials, and thereby, the circumstance in the vacuum vessel is under intense gamma radiation field. Thus, the in-vessel components such as blanket are handled and replaced by remote handling equipment. The objective of this report is to study the compactness of the remote handling equipment (a vehicle/manipulator) for the ITER blanket maintenance. In order to avoid the interferences between the blanket and the equipment during blanket replacement in the restricted vacuum vessel, a compact design of the equipment is required. Therefore, the compact design is performed, including kinematic analyses aiming at the reduction of the sizes of the vehicle equipped with a manipulator handling the blanket and the rail for the vehicle traveling in the vacuum vessel. Major results are as follows: 1. The compact vehicle/manipulator is designed concentration on the reduction of the rail size and simplification of the guide roller mechanism as well as the reduction of the gear diameter for vehicle rotation around the rail. Height of the rail is reduced from 500 mm to 400 mm by a parameter survey for weight, stiffness and stress of the rail. The roller mechanism is divided into two simple functional mechanisms composed of rollers and a pad, that is, the rollers support relatively light loads during rail deployment and vehicle traveling while a pad supports heavy loads during blanket replacement. Regarding the rotation mechanism, the double helical gear is adopted, because it has higher contact ratio than the normal spur gear and consequently can transfer higher force. The smaller double helical gear, 996 mm in diameter, can achieve 26% higher output torque, 123.5 kN·m, than that of the original spur gear of 1,460 mm in diameter, 98 kN·m. As a result, the manipulator becomes about 30% lighter, 8 tons, than the original weight, 11.2 tons. 2. Based on the compact design of the vehicle/manipulator, the

  17. Evaluation of steam as a potential coolant for nonbreeding blanket designs

    International Nuclear Information System (INIS)

    Stevens, H.C.; Misra, B.; Youngdahl, C.K.

    1978-01-01

    A steam-cooled nonbreeding blanket design has been developed as an evolution of the Argonne Experimental Power Reactor (EPR) studies. This blanket concept complete with maintenance considerations is to function at temperatures up to 650 0 C utilizing nickel-based alloys such as Inconel 625. Thermo-mechanical analyses were carried out in conjunction with thermal hydraulic analysis to determine coolant chennel arrangements that permit delivery of superheated steam at 500 0 C directly to a modern fossil plant-type turbine. A dual-cycle system combining a pressurized water circuit coupled with a superheated steam circuit can produce turbine plant conversion efficiencies approaching 41.5%

  18. Review of tokamak power reactor and blanket designs in the United States

    International Nuclear Information System (INIS)

    Baker, C.; Brooks, J.; Ehst, D.; Gohar, Y.; Smith, D.; Sze, D.

    1986-01-01

    The last major conceptual design study of a tokamak power reactor in the United States was STARFIRE which was carried out in 1979-1980. Since that time US studies have concentrated on engineering test reactors, demonstration reactors, parametric systems studies, scoping studies, and studies of selected critical issues such as pulsed vs. steady-state operation and blanket requirements. During this period, there have been many advancements in tokamak physics and reactor technology, and there has also been a recognition that it is desirable to improve the tokamak concept as a commercial power reactor candidate. During 1984-1985 several organizations participated in the Tokamak Power Systems Study (TPSS) with the objective of developing ideas for improving the tokamak as a power reactor. Also, the US completed a comprehensive Blanket Comparison and Selection Study which formed the basis for further studies on improved blankets for fusion reactors

  19. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  20. Conceptual design of the blanket mechanical attachment for the helium-cooled lithium-lead reactor

    International Nuclear Information System (INIS)

    Barrera, G.; Branas, B.; Lucas, J.; Doncel, J.; Medrano, M.; Garcia, A.; Giancarli, L.; Ibarra, A.; Li Puma, A.; Maisonnier, D.; Sardain, P.

    2008-01-01

    The conceptual design of a new type of fusion reactor based on the helium-cooled lithium-lead (HCLL) blanket has been performed within the European Power Plant Conceptual Studies. As part of this activity, a new attachment system suitable for the HCLL blanket modules had to be developed. This attachment is composed of two parts. The first one is the connection between module and the first part of a shield, called high temperature shield, which operates at a temperature around 500 deg. C, close to that of the blanket module. This connection must be made at the lateral walls, in order to avoid openings through the first wall and breeding zone thus avoiding complex design and fabrication issues of the module. The second connection is the one between the high temperature shield and a second shield called low temperature shield, which has a temperature during reactor operation around 150 deg. C. The design of this connection is complex because it must allow the large differential thermal expansion (up to 30 mm) between the two components. Design proposals for both connections are presented, together with the results of finite element mechanical analyses which demonstrate the feasibility to support the blanket and shield modules during normal and accidental operation conditions

  1. Preconceptual design of a packed fluidized bed blanket for a fission suppressed thorium-fueled CTHR

    International Nuclear Information System (INIS)

    Chi, J.W.H.; Karbowski, J.S.; Chapin, D.L.

    1981-01-01

    This paper describes a thorium-fueled PFB blanket concept for a Commercial Tokamak Hybrid Reactor. A preliminary mechanical concept is presented and the results of neutronics, thermal-hydraulics and economics analyses are discussed. Futher work needed to design and advance the concept is recommended

  2. Present day design challenges exemplified by the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Dickson, P.W. Jr.; Anderson, C.A. Jr.

    1976-01-01

    The present day design challenges faced by the Clinch River Breeder Reactor Plant engineer result from two causes. The first cause is aspiration to achieve a design that will operate at conditions which are desirable for future LMFBRs in order for them to achieve low power costs and good breeding. The second cause is the licensing impact. Although licensing the CRBRP won't eliminate future licensing effort, many licensing questions will have been resolved and precedents set for the future LMFBR industry

  3. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    International Nuclear Information System (INIS)

    Pope, R.B.; Diggs, J.M.

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented

  4. Mechanical design and analysis for a EPR first wall/blanket/shield system

    International Nuclear Information System (INIS)

    Stevens, H.C.; Misra, B.; Youngdahl, C.K.

    1978-01-01

    Continuing studies are in progress at ANL to expand upon the design of a first wall/blanket/shield FW/B/S system and power conversion for a tokamak type Experimental Power Reactor (EPR). The FW/B/S system has evolved from an earlier design for a low beta, circular cross section plasma (major radius = 6 m) to one for a higher beta elongated plasma with a 4.7 m major radius. Basic mechanical design and layout features of the old and new EPR designs showing some of the more important design developments are given. These developments are aimed at simplifying the design, reducing the costs and in addition, improving the plant thermal efficiency and overall maintainability. In the area of the reactor blanket, significant thermal hydraulic and stress analysis have been performed to substantiate the integrity of the chosen concept. This paper deals with the discussion of these improved features

  5. CM-244 as multiplier and breeder in a ThO/sub 2/ hybrid blanket driven by a (D,T) source

    International Nuclear Information System (INIS)

    Sahin, S.; Al-Kusayer, T.A.

    1986-01-01

    The safeguard aspects of Cm-244 - a nuclear waste product in LWRs - in a cylindrical hybrid blanket, driven by a (D,T) fusion neutron source have been analyzed. Cm-244 is investigated for two different applications: 1) as a neutron multiplier between the first wall and the fuel zone in a blanket with ThO/sub 2/; and 2) as a component of the mixed fuel, ThO/sub 2/-Cm/sup 244/O/sub 2/, used for power flattening in a hybrid blanket. The calculations show that a relatively small driven with 100 MW/sub th/ fusion power could produce about 5 kg/year Cm-245, enough to provide nuclear fuel for up to 50 explosives. The study suggests an extension of the safe-guarding regulations prior to the commercial introduction of fusion reactors in the energy market

  6. Bouyancy effects on sodium coolant temperature profiles measured in an electrically heated mock-up of a 61-rod breeder reactor blanket assembly

    International Nuclear Information System (INIS)

    Engel, F.C.; Markley, R.A.; Minushkin, B.

    1978-01-01

    The paper describes test results selected to demonstrate the effect of buoyancy on the temperature profiles in a 61-rod electrically heated mock-up of an LMFBR radial blanket assembly. In these assemblies, heat transfer occurs over a wide range of complex operating conditions. The range and complexity of conditions are the result of the steep flux and power gradients which are an inherent feature of the blanket region and the power generation level in an assembly which can vary from 20 to 1100 kW

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

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Hiroshi; Ishitsuka, Etsuo (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Sakamoto, Naoki; Kato, Masakazu; Takatsu, Hideyuki.

    1992-11-01

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

  8. Status of ITER blanket attachment design and related R and D

    International Nuclear Information System (INIS)

    Sadakov, S.; Khomiakov, S.; Calcagno, B.; Chappuis, Ph.; Dellopoulos, G.; Kolganov, V.; Merola, M.; Poddubnyi, I.; Raffray, R.; Raharijaona, J.J.; Ulrickson, M.; Zhmakin, A.

    2013-01-01

    Highlights: • ITER blanket attachment system went through a significant design upgrade and become basically compliant with specified design loads and required cyclic lifetime. • Upgrade of flexible supports allowed the doubling of cross sections of central bolts. Ceramic coatings were relocated to much larger areas on conical pairs screwed into shield blocks. • Key pads were relocated from keys of vacuum vessel into keyways of shield blocks and reshaped to enlarge areas of lateral interfaces with ceramic electro-insulating coatings. • Ceramic coatings are hidden between pads and enclosures in keyways with a purpose to minimize their wear rate, which depends on peak friction stress and cyclic sliding path. • Ceramic coatings to be verified by experiment, with several R and D aimed to collect statistically sufficient data on their reliability and durability in ITER relevant cyclic loading conditions. -- Abstract: Main function of the ITER blanket system [1–3] is to shield the vacuum vessel (VV) from nuclear radiation and thermal energy coming from the plasma. Blanket system consists of discrete blanket modules (BM). Each BM is composed of a first wall panel and a shield block (SB). The shield block is attached to the VV by means of four flexible supports and three or four shear keys, through key pads. All listed supports do have parts with ceramic electro-insulating coatings necessary to exclude the largest loops of eddy currents and restrict EM loads. Electrical connection of each SB to the VV is through two elastic electrical straps. Cooling water is supplied to each BM by one coaxial water connector. This paper summarizes the recent evolution of the blanket attachment system toward design solutions compatible with design loads and numbers of load cycles, and providing sufficient reliability and durability. This evolution was done in a frame of pre-defined external interfaces. The ongoing supporting R and D is also briefly described

  9. Nuclear design of the blanket/shield system for a Tokamak Experimental Power Reactor

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1976-01-01

    The various options and trade-offs in the nuclear design of the blanket/shield for a Tokamak Experimental Power Reactor (TEPR) are investigated. The TEPR size and cost are particularly sensitive to the blanket/shield thickness, Δ/sub BS/, on the inner side of the torus. Radition damage to the components of the superconducting magnet and refrigeration power requirements set lower limits on Δ/sub BS/. These limits are developed in terms of TEPR design parameters such as the wall loading, duty cycle, and frequency of magnet anneals. The study of the nuclear performance of various material compositions shows that mixtures of tungsten, or tantalum, or stainless-steel alloys and boron carbide require the smallest Δ/sub BS/ for a given attenuation. This Δ/sub BS/ has to be doubled if the low induced activation materials graphite and aluminum are used. The space problems are greatly eased in the Argonne National Laboratory ANL-TEPR reference design by using two separate segments of the blanket/shield. The inner segment occupies the region of the high magnetic field, uses very efficient attenuators (tungsten- or tantalum- or stainless-steel-boron carbide mixtures), and is only 1 m thick. The outer blanket/shield is 131 cm and consists of an optimized composition of stainless steel and boron carbide. For the design parameters of 0.2 MW/m 2 neutron wall loading and 50 percent duty cycle, the reactor components can operate satisfactorily up to (a) 10 yr for the stainless-steel first wall, (b) 10 yr for the superconductor composite after which magnet warmup becomes necessary, and (c) 30 yr for the Mylar insulation. Nuclear heat generation rates in the blanket/shield and magnet are well within the practical limits for heat removal

  10. Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M.

    2000-01-01

    Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve ∼50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed

  11. Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K. E-mail: iokik@itereu.deiokik@ipp.mpg.de; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M

    2000-12-01

    Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve {approx}50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed.

  12. Design and fabrication methods of FW/blanket, divertor and vacuum vessel for ITER

    Science.gov (United States)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Ibbott, C.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Tivey, R.; Utin, Y.; Yamada, M.

    2000-12-01

    Design has progressed on the vacuum vessel, FW/blanket and Divertor for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design [K. Ioki et al., J. Nucl. Mater. 258-263 (1998) 74]. Design and fabrication methods of the components have been improved to achieve ˜50% reduction of the construction cost. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R&D performed so far during the Engineering Design Activities (EDAs) are still applicable. Further cost reduction methods are also being investigated and additional R&D is being performed.

  13. Tritium containment and blanket design challenges for a 1 GWe mirror fusion central power station

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1976-06-01

    Tritium containment and removal problems associated with the blanket and power-systems for a mirror fusion reactor are identified and conceptual process designs are devised to reduce emissions to the environment below 1 Ci/day. The blanket concept development proceeds by starting with this emission goal of 1 Ci/day and working inward to the blanket. At each decision point, worker safety, operational labor costs, and capital cost tradeoffs are contrasted. The conceptual design uses air for the reactor hall with a continuous catalytic oxidizer-molecular sieve adsorber cleanup system to maintain a 40 μCi/m 3 tritium level (5 μCi/m 3 HTO) against 180 Ci/day leakage from reactor components, energy recovery systems, and process piping. This blanket contains submodules with Li 2 Be 2 O 3 --Be for tritium breeding and submodules with Be for mostly energy production. Tritium production in both is handled by separately containing this breeding material and scavenging this container with lithium vapor-doped helium gas stream

  14. Design Feasible Area on Water Cooled Thorium Breeder Reactor in Equilibrium States

    International Nuclear Information System (INIS)

    Sidik Permana; Naoyuki Takaki; Hiroshi Sekimoto

    2006-01-01

    Thorium as supplied fuel has good candidate for fuel material if it is converted into fissile material 233 U which shows superior characteristics in the thermal region. The Shippingport reactor used 233 U-Th fuel system, and the molten salt breeder reactor (MSBR) project showed that breeding is possible in a thermal spectrum. In the present study, feasibility of water cooled thorium breeder reactor is investigated. The key properties such as flux, η value, criticality and breeding performances are evaluated for different moderator to fuel ratios (MFR) and burn-ups. The results show the feasibility of breeding for different MFR and burn-ups. The required 233 U enrichment is about 2% - 9% as charge fuel. The lower MFR and the higher enrichment of 233 U are preferable to improve the average burn-up; however the design feasible window is shrunk. This core shows the design feasible window especially in relation to MFR with negative void reactivity coefficient. (authors)

  15. Breeding blanket for Demo

    International Nuclear Information System (INIS)

    Proust, E.; Giancarli, L.

    1992-01-01

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

  16. Design safety limits in prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Puthiyavinayagam, P.; Roychowdhury, D.G.; Govindarajan, S.; Chellapandi, P.; Singh, Om Pal; Chetal, S.C.

    2002-01-01

    Full text: PFBR is designed to operate at 450 W/cm peak linear heat rating to a peak burn up of 100,000 MWd/t which corresponds to a damage dose of 85 dpa. The targetted reliability is to restrict pin failure to 1 in 10,000. All the design basis events are classified into four categories. Design safety limits imposed for DBE are in terms of temperatures, radiation doses and structural design parameters. Radiation limits are imposed in relation to RCB from the plant personnel and public point of view. Fuel pin integrity is assured with a detailed damage analysis by adopting cumulative damage concept for fixing clad temperature limits. Fuel temperatures are limited to melting point to preclude fuel slumping for events up to category 3. Partial melting is allowed for events in category 4 and the results obtained from transients experiments show that partial melting up to 50% of pellet area does not result in clad failure. Coolant temperatures are limited to boiling point to avoid burnout and reactivity effects

  17. Thermal insulation system design and fabrication specification (nuclear) for the Clinch River Breeder Reactor plant

    International Nuclear Information System (INIS)

    1978-01-01

    This specification defines the design, analysis, fabrication, testing, shipping, and quality requirements of the Insulation System for the Clinch River Breeder Reactor Plant (CRBRP), near Oak Ridge, Tennessee. The Insulation System includes all supports, convection barriers, jacketing, insulation, penetrations, fasteners, or other insulation support material or devices required to insulate the piping and equipment cryogenic and other special applications excluded. Site storage, handling and installation of the Insulation System are under the cognizance of the Purchaser

  18. Current status on detail design and fabrication techniques development of ITER blanket shield block in Korea

    International Nuclear Information System (INIS)

    Kim, Duck Hoi; Cho, Seungyon; Ahn, Mu-Young; Lee, Eun-Seok; Jung, Ki Jung

    2007-01-01

    The allocation of components and systems to be delivered to ITER on an in-kind basis, was agreed between the ITER Parties. Among parties, Korea agreed to procure inboard blanket modules 1, 2 and 6, which consists of FW and shield block. Regarding shield block the detail design and Fabrication techniques development have been undertaken in Korea. Especially manufacturing feasibility study on shield block had been performed and some technical issues for the fabrication were selected. Based on these results, fabrication techniques using EB welding are being developed. Meanwhile, the detail design of inboard standard module has been carried out. The optimization of flow driver design to improve the cooling performance was executed. And, thermo-hydraulic analysis on half block of inboard standard module was performed. In this study, current status and some results from Fabrication techniques development on ITER blanket shield block are described. The detail design activity and results on shield block are also introduced herein. (orig.)

  19. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.

    1995-01-01

    Proposed International Thermonuclear Experimental Reactor (ITER) convertible blankets were reviewed. Key design difficulties were identified. A new particle filter concept is introduced and key performance parameters estimated. Results show that this particle filter concept can satisfy all of the convertible blanket design requirements except the generic issue of Be blanket lifetime. If the convertible blanket is an acceptable approach for ITER operation, this particle filter option should be a strong candidate

  20. Evaluation of US demo helium-cooled blanket options

    International Nuclear Information System (INIS)

    Wong, C.P.C.; McQuillan, B.W.; Schleicher, R.W.

    1995-10-01

    A He-V-Li blanket design was developed as a candidate for the U.S. fusion demonstration power plant. This paper presents an 18 MPa helium-cooled, lithium breeder, V-alloy design that can be coupled to the Brayton cycle with a gross efficiency of 46%. The critical issue of designing to high gas pressure and the compatibility between helium impurities and V-alloy are addressed

  1. Blanket and vacuum vessel design of the next tokamak. (Swimming pool type)

    International Nuclear Information System (INIS)

    Iida, H.; Minato, A.; Kitamura, K.

    1983-01-01

    The structural design study of a reactor module for a swimming pool type reactor (SPTR) was conducted. Since pool water plays the role of radiation shielding in the SPTR, the module does not have a solid shield. It consists of tritium breeding blankets, divertor collector plates and a vacuum vessel. The object of this study is to show the reactor module design which has a simple structure and a sufficient tritium breeding ratio. A large coverage of the plasma chamber surface with tritium breeding blanket is essential in order to obtain a high tritium breeding ratio. A breeding blanket is also placed behind the divertor collector plate, i.e. in the upper and lower region, as well as in the outboard and inboard regions of the module. A concept in which the first wall is an integral part of the blanket is employed to minimize the thickness of structural and cooling material brazed in front of the breeding material (Li 2 O) and to enhance the tritium breeding capability. In order to simplify the module structure the vacuum vessel and breeding blanket is also integrated in the inboard region. One of the features inherent in the swimming pool type reactor is an additional external force on the vacuum vessel, namely hydraulic pressure. A detailed structural analysis of the vacuum vessel is performed. Divertor collector plates are assemblies of co-axial tubes. They minimize the electromagnetic force on the plate induced by the plasma disruption. A thermal and structural analysis and life time estimation of the first wall and divertor collector plates are performed. (author)

  2. Target/blanket conceptual design for the Los Alamos ATW concept

    International Nuclear Information System (INIS)

    Ames, K.; Cappiello, M.; Ireland, J.; Sapir, J.; Farnum, G.

    1992-01-01

    The Los Alamos Accelerator Transmutation of Waste (ATW) concept has many potential applications that include defense waste transmutation, defense material production (i.e., tritium and 238 Pu), and the transmutation of hazardous nuclear wastes from commercial nuclear reactors (fission products and actinides). A more advanced long-term Los Alamos effort is investigating the potential of an accelerator- driven system to produce fission energy with a minimal nuclear waste stream. All applications employ a high-energy (800- to 1600-MeV), high-current (25--250 mA) proton linear accelerator as the driver. In this report, we discuss only the target/blanket conceptual design for the commercial nuclear waste application. A conceptual design for the target/blanket of the Los Alamos ATW concept has been presented. The neutronics, mechanical design, and heat transfer have been investigated in some detail for the base-case design. Much more work needs to be done, but at this point it appears that the design is feasible and will approach the design goal of supporting two commercial power reactors with each target/blanket module

  3. Preconceptual engineering design for the APT 3He Target/Blanket concept

    International Nuclear Information System (INIS)

    Mensink, D.L.

    1994-01-01

    A preconceptual engineering design has been developed for the 3 He Target/Blanket (T/B) System for the Accelerator Production of Tritium Project. This concept uses an array of pressure tubes containing tungsten rods for the neutron spallation source and 3 He gas contained in a metal tank and blanket tubes as the tritium production material. The engineering design is based on a physics model optimized for efficient tritium production. Principle engineering consideration were: provisions for cooling all materials including the 3 He gas; containment of the gas and radionuclides; remote handling; material compatibility; minimization of 3 He, D 2 O, and activated waste; modularity; and manufacturability. The design provides a basis for estimating the cost to implement the system

  4. Design of the low energy beam transport line between CARIBU and the EBIS charge breeder

    Energy Technology Data Exchange (ETDEWEB)

    Perry, A., E-mail: aperry4@hawk.iit.edu [Argonne National Laboratory, Argonne, IL 60439, USA and Illinois Institute of Technology, Chicago, IL 60616 (United States); Ostroumov, P. N.; Barcikowski, A.; Dickerson, C.; Kondrashev, S. A.; Mustapha, B.; Savard, G. [Argonne National Laboratory, Argonne, IL 60439 (United States)

    2015-01-09

    An Electron Beam Ion Source Charge Breeder (EBIS-CB) has been developed to breed radioactive beams from the CAlifornium Rare Isotope Breeder Upgrade (CARIBU) facility at ATLAS. The EBIS-CB will replace the existing ECR charge breeder to increase the intensity and improve the purity of reaccelerated radioactive ion beams. The EBIS-CB is in the final stage of off-line commissioning. Currently, we are developing a low energy beam transport (LEBT) system to transfer CARIBU beams to the EBIS-CB. As was originally planned, an RFQ cooler-buncher will precede the EBIS-CB. Recently, it was decided to include a multi-reflection time-of-flight (MR-TOF) mass-spectrometer following the RFQ. MR-TOF is a relatively new technology used to purify beams with a mass-resolving power up to 3×10{sup 5} as was demonstrated in experiments at CERN/ISOLDE. Very high purity singly-charged radioactive ion beams will be injected into the EBIS for charge breeding and due to its inherent properties, the EBIS-CB will maintain the purity of the charge bred beams. Possible contamination of residual gas ions will be greatly suppressed by achieving ultra-high vacuum in the EBIS trap. This paper will present and discuss the design of the LEBT and the overall integration of the EBIS-CB into ATLAS.

  5. Progress on solid breeder TBM at SWIP

    International Nuclear Information System (INIS)

    Feng, K.M.; Pan, C.H.; Zhang, G.S.; Luo, T.Y.; Zhao, Z.; Chen, Y.J.; Ye, X.F.; Hu, G.; Wang, P.H.; Yuan, T.; Feng, Y.J.; Xiang, B.; Zhang, L.; Wang, Q.J.; Cao, Q.X.; Li, Z.X.; Wang, F.

    2010-01-01

    Current progress on the design and R and D of Chinese helium-cooled solid breeder test blanket module, CN HCSB TBM is presented. The updated design on structural, neutronics, thermal-hydraulics and safety analysis has been completed. In order to accommodate the HCSB TBM ancillary system, the design and necessary R and Ds corresponding sub-systems have being developed. Current status on the development of function materials, structure material and the helium test loop are also presented. The Chinese low-activation ferritic/martensitic steels CLF-1, which is the structural material for the HCSB TBM is being manufactured by industry. The neutron multiplier Be and tritium breeder Li 4 SiO 4 pebbles are being prepared in laboratory scale.

  6. Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

    International Nuclear Information System (INIS)

    Palermo, I.; Gómez-Ros, J.M.; Veredas, G.; Sanz, J.; Sedano, L.

    2012-01-01

    Highlights: ► Dual-Coolant He/Pb15.7Li breeding blanket for a DEMO fusion reactor is studied. ► An iterative process optimizes neutronic responses minimizing reactor dimension. ► A 3D toroidally symmetric geometry has been generated from the CAD model. ► Overall TBR values support the feasibility of the conceptual model considered. ► Power density in TF coils is below load limit for quenching. - Abstract: The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.

  7. Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine

    International Nuclear Information System (INIS)

    Powers, J.

    2008-01-01

    The Laser Inertial Confinement Fusion Fission Energy (LIFE) Program being developed at Lawrence Livermore National Laboratory (LLNL) aims to design a hybrid fission-fusion subcritical nuclear engine that uses a laser-driven Inertial Confinement Fusion (ICF) system to drive a subcritical fission blanket. This combined fusion-fission hybrid system could be used for generating electricity, material transmutation or incineration, or other applications. LIFE does not require enriched fuel since it is a sub-critical system and LIFE can sustain power operation beyond the burnup levels at which typical fission reactors need to be refueled. In light of these factors, numerous options have been suggested and are being investigated. Options being investigated include fueling LIFE engines with spent nuclear fuel to aid in disposal/incineration of commercial spent nuclear fuel or using depleted uranium or thorium fueled options to enhance proliferation resistance and utilize non-fissile materials (1]. LIFE engine blanket designs using a molten salt fuel system represent one area of investigation. Possible applications of a LIFE engine with a molten salt blanket include uses as a spent nuclear fuel burner, fissile fuel breeding platform, and providing a backup alternative to other LIFE engine blanket designs using TRISO fuel particles in case the TRISO particles are found to be unable to withstand the irradiation they will be subjected to. These molten salts consist of a mixture of LiF with UF 4 or ThF 4 or some combination thereof. Future systems could look at using PuF 3 or PuF 4 as well, though no work on such system with initial plutonium loadings has been performed for studies documented in this report. The purpose of this report is to document preliminary neutronics design studies performed to support the development of a molten salt blanket LIFE engine option, as part of the LIFE Program being performed at Lawrence Livermore National laboratory. Preliminary design studies

  8. Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J

    2008-10-23

    The Laser Inertial Confinement Fusion Fission Energy (LIFE) Program being developed at Lawrence Livermore National Laboratory (LLNL) aims to design a hybrid fission-fusion subcritical nuclear engine that uses a laser-driven Inertial Confinement Fusion (ICF) system to drive a subcritical fission blanket. This combined fusion-fission hybrid system could be used for generating electricity, material transmutation or incineration, or other applications. LIFE does not require enriched fuel since it is a sub-critical system and LIFE can sustain power operation beyond the burnup levels at which typical fission reactors need to be refueled. In light of these factors, numerous options have been suggested and are being investigated. Options being investigated include fueling LIFE engines with spent nuclear fuel to aid in disposal/incineration of commercial spent nuclear fuel or using depleted uranium or thorium fueled options to enhance proliferation resistance and utilize non-fissile materials [1]. LIFE engine blanket designs using a molten salt fuel system represent one area of investigation. Possible applications of a LIFE engine with a molten salt blanket include uses as a spent nuclear fuel burner, fissile fuel breeding platform, and providing a backup alternative to other LIFE engine blanket designs using TRISO fuel particles in case the TRISO particles are found to be unable to withstand the irradiation they will be subjected to. These molten salts consist of a mixture of LiF with UF{sub 4} or ThF{sub 4} or some combination thereof. Future systems could look at using PuF{sub 3} or PuF{sub 4} as well, though no work on such system with initial plutonium loadings has been performed for studies documented in this report. The purpose of this report is to document preliminary neutronics design studies performed to support the development of a molten salt blanket LIFE engine option, as part of the LIFE Program being performed at Lawrence Livermore National laboratory

  9. Materials for breeding blankets

    International Nuclear Information System (INIS)

    Mattas, R.F.; Billone, M.C.

    1995-09-01

    There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as Primary Blanket Materials, which have the greatest influence in determining the overall design and performance, and Secondary Blanket Materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified

  10. Materials for breeding blankets

    International Nuclear Information System (INIS)

    Mattas, R.F.; Billone, M.C.

    1996-01-01

    There are several candidate concepts for tritium breeding blankets that make use of a number of special materials. These materials can be classified as primary blanket materials, which have the greatest influence in determining the overall design and performance, and secondary blanket materials, which have key functions in the operation of the blanket but are less important in establishing the overall design and performance. The issues associated with the blanket materials are specified and several examples of materials performance are given. Critical data needs are identified. (orig.)

  11. System engineering approach in the EU Test Blanket Systems Design Integration

    International Nuclear Information System (INIS)

    Panayotov, D.; Sardain, P.; Boccaccini, L.V.; Salavy, J.-F.; Cismondi, F.; Jourd'Heuil, L.

    2011-01-01

    The complexity of the Test Blanket Systems demands diverse and comprehensive integration activities. Test Blanket Modules - Consortia of Associates (TBM-CA) applies the system engineering methods in all stages of the Test Blanket System (TBS) design integration. Completed so far integration engineering tasks cover among others status and initial set of TBS operating parameters; list of codes, standards and regulations related to TBS; planning of the TBS interfaces and baseline documentation. Most of the attention is devoted to the establishment the Helium-Cooled Lithium Lead (HCLL) and Helium-Cooled Pebble Bed Lead (HCPB) TBS configuration baseline, TBS break down into sub-systems, identification, definition and management of the internal and external interfaces, development of the TBS plant break down structure (PBS), establishment and management of the required TBS baseline documentation infrastructure. Break down of the TBS into sub-systems that is crucial for the further design and interfaces' management has been selected considering several options and using specific evaluation criteria. Process of the TBS interfaces management covers the planning, definition and description, verification and review, non-conformances and deviations, and modification and improvement processes. Process of interfaces review is developed, identifying the actors, input, activities and output of the review. Finally the relations and interactions of system engineering processes with TBM configuration management and TBM-CA Quality Management System are discussed.

  12. Design considerations for large heterogeneous liquid-metal fast breeder reactors

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Barthold, W.P.

    1977-01-01

    A systematic method for designing heterogeneous configurations having a near-zero value of sodium void reactivity is presented. It is based on the following principles: (a) the thickness of the internal blanket zones should be such that the reactivity change resulting from voiding any core zone is practically independent of any further increase in the thickness of these zones, and (b) the sodium void reactivity of each core zone must have a near-zero value. Neutronic coupling among the core zones of heterogeneous configurations decreases as the thickness of the internal blanket zones increases. To quantify coupling, Avery's coupling coefficients are used. Reduced coupling among the core zones of a heterogeneous design, compared to a homogeneous design, results in (a) increased sensitivity of the power distribution to enrichment distribution perturbations, (b) reduced reactivity worth of local perturbations, and (c) higher cladding temperatures during operational transients initiated by local perturbations. Heterogeneous designs compared to equivalent homogeneous designs have (a) lower core Doppler coefficient values, (b) larger fuel compaction reactivities, and (c) higher maximum cladding temperatures

  13. Tritium Cycle Design for He-cooled Blankets for Demo

    International Nuclear Information System (INIS)

    Sedano, L. A.

    2007-01-01

    Final goal of COMPU task is to develop a reliable tritium Process Flow Diagram (PFD) modelling tool for DEMO tritium cycle. With this aim, the COMPU task is devoted to: (1) Review of existing available documentation related on configuration layouts, and systems and tritium control process key technologies. (2) To select those validated and considered relevant as basis for code development. (3) Implement results from (1), and (2) in the PFD TRICICLO. This fi rst deliverable focuses on item (1) and is conceived as a managerial tool to: (1) establish and discuss the correct inputs, (2) to identify existing lack of basic information and (3) to establish the general demands and characteristics for the development of an advanced PFD model. Thus, in order to discuss and determine the basic information required for future new developments of the task, this report presents a review of the documentation of: (1) The outline of total cycle and system configuration with the main tritium system design specifications. (2) The ultimate processing technologies with the associated design of their implementing units. (3) Key parameters needed to describe processes and modes of operation of the system units. (4) An overview of the existing models for cycle and units with a general analysis of their performances and limitations. Thus, this report is a direct review of the base information generated previously in the context of tasks of the EU FT Programmers (reported in EFDA Green Books) and available results in open fields literature provided by parallel Programmes abroad (JP, US, RF). (Author) 102 refs

  14. Tritium Cycle Design for He-cooled Blankets for Demo

    Energy Technology Data Exchange (ETDEWEB)

    Sedano, L. A.

    2007-09-27

    Final goal of COMPU task is to develop a reliable tritium Process Flow Diagram (PFD) modelling tool for DEMO tritium cycle. With this aim, the COMPU task is devoted to: (1) Review of existing available documentation related on configuration layouts, and systems and tritium control process key technologies. (2) To select those validated and considered relevant as basis for code development. (3) Implement results from (1), and (2) in the PFD TRICICLO. This fi rst deliverable focuses on item (1) and is conceived as a managerial tool to: (1) establish and discuss the correct inputs, (2) to identify existing lack of basic information and (3) to establish the general demands and characteristics for the development of an advanced PFD model. Thus, in order to discuss and determine the basic information required for future new developments of the task, this report presents a review of the documentation of: (1) The outline of total cycle and system configuration with the main tritium system design specifications. (2) The ultimate processing technologies with the associated design of their implementing units. (3) Key parameters needed to describe processes and modes of operation of the system units. (4) An overview of the existing models for cycle and units with a general analysis of their performances and limitations. Thus, this report is a direct review of the base information generated previously in the context of tasks of the EU FT Programmers (reported in EFDA Green Books) and available results in open fields literature provided by parallel Programmes abroad (JP, US, RF). (Author) 102 refs.

  15. Assessment of the accident response of a light-water-moderated breeder-reactor system: AWBA development program

    International Nuclear Information System (INIS)

    High, H.M.

    1983-05-01

    The predicted accident response for a light water moderated, thorium/U-233 fueled, seed-blanket reactor concept was assessed. The first part of the assessment compared breeder accident response with that of a current commercial pressurized water reactor design for several different types of transients. Based on these comparisons and a review of the various parameter differences between the breeder and a U-235 fueled plant, the second part of the assessment studied the breeder accident behavior in more detail, particularly in areas of potential concern. Based on the two parts of the assessment, it was concluded that the breeder accident response would be very similar to that of present commercial pressurized water reactor plants. The large Doppler and moderator reactivity coefficients of the breeder would significantly reduce the severity of many of the accidents that must be considered. It is expected that the accident response of the breeder can be shown to meet regulatory criteria

  16. Current design of the European TBM systems and implications on DEMO breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Ricapito; Calderoni, P. [Fusion for Energy, 08019 Barcelona (Spain); Aiello, A. [ENEA, Bacino del Brasimone, I-40032 Camugnano, Bo (Italy); Ghidersa, B. [Karlsruher Institut für Technologie, D-76021 Karlsruhe (Germany); Poitevin, Y.; Pacheco, J. [Fusion for Energy, 08019 Barcelona (Spain)

    2016-11-01

    Highlights: • Description of the Helium Cooling Systems of HCLL and HCPB-TBS after the Conceptual Design Review. • Description of the PbLi loop of HCLL-TBS after the Conceptual Design Review. • Description of the possible ROX (Return of Experience) from design and operation of the Test Blanket Systems. • Discussion on the DEBO relevancy of the main technologies adopted in the Helium Cooling Systems and PbLi loop. - Abstract: Europe is committed in developing the design of the two Test Blanket Systems (TBS) based on HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed) breeding blanket (BB) concepts. The complexity of the TBS design comes not only from the innovative fabrication technologies and materials adopted for Test Blanket Modules (TBM) but also from the requirements and functions that the TBM ancillary systems have to satisfy and implement. Indeed, the main TBM ancillary systems, namely the Helium Cooling System, the Coolant Purification System and Tritium Extraction System, all belonging to the Safety Important Class (SIC), have to implement fundamental functions, like the transport of the surface and volumetric heat from the TBM to the heat sink, the extraction and processing of the tritium generated in the TBM, the confinement of radioactive inventory, the support to the investment protection and safety functions. On top of the full compliance with the ITER safety principles, the design of the TBM systems is focused on providing high operational reliability and availability not to jeopardize ITER program and, at the same time, also a good operational flexibility to make possible the achievement of the main TBM scientific objectives. This paper gives an overview of the design status of the HCLL and HCPB-TBM (ancillary) systems, updated to the conclusion of the conceptual design phase (CDR). The most relevant technologies, the still open points, the main issues related to the integration in ITER and last relevant results from the on

  17. Comparison of different fusion nuclear data libraries using the European INTOR blanket design

    International Nuclear Information System (INIS)

    Pelloni, S.; Stepanek, J.; Dudziak, D.

    1982-12-01

    The European Community International Tokamak Reactor (INTOR-EC) was used to investigate the influence of different cross-section libraries on the tritium breeding ratio. Nucleonic analyses were performed using the discrete-ordinates transport codes ANISN and ONEDANT, and the recently developed Swiss surface-flux code SURCU, for the Li 17 Pb 83 and Li 2 SiO 3 blanket designs. Nuclear data considered were from the DLC-37, VITAMIN-C (DLC-41) and Los Alamos-NJOY fusion libraries. In addition the reaction rates were estimated using the MACKLIB-IV response library. It is shown that very good agreement (within 0.5%) between the breeding ratios obtained using the VITAMIN-C and Los Alamos libraries could be obtained, whereas the corresponding values calculated using VITAMIN-C and MACKLIB-IV data sets collapsed into 25 neutron and 21 gamma groups differ up to 23%. It is found that this large discrepancy is due to the 6 Li(n, α) reaction cross sections in the low energy range between 4 and 0.03 eV. Furthermore, the collapsed DLC-37 library is not adequate for fusion blankets with a soft spectrum. It is important that greater care be given to preparation of broad group cross section sets, especially in the thermal energy region for blankets containing highly moderating materials. (Auth.)

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

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

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

    International Nuclear Information System (INIS)

    Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2015-01-01

    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

  20. Tritium dynamics in fusion reactor solid breeder

    International Nuclear Information System (INIS)

    Violante, V.

    1986-01-01

    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

  1. ITER blanket module connectors. Design, analysis and testing for procurement arrangement

    International Nuclear Information System (INIS)

    Khomiakov, S.; Poddubnyi, I.; Kolganov, V.; Zhmakin, A.; Parshutin, E.; Danilov, I.; Strebkov, Yu.; Skladnov, K.; Vlasov, D.; Cheburova, A.; Romannikov, A.; Raffray, R.; Egorov, K.; Chappuis, Ph.; Sadakov, S.; Calcagno, B.; Roccella, R.

    2016-01-01

    Highlights: • Procurement Arrangement on Blanket Module Connections (BMC) was signed by ITER Organization and Russian Federation Domestic Agency in late 2014. • “N.A. Dollezhal Research and Development Institute of Power Engineering” (NIKIET) was selected as a general supplier of BMC. • NIKIET plays a key role in design development, analytical and experimental justification and manufacturing of BMC. • NIKIET shall fabricate, test and deliver to ITER 2109 flexible supports, 2561 pads, 1053 electrical straps and 1053 pedestals. - Abstract: A standard ITER Blanket module (BM) is attached to the Vacuum Vessel (VV) with a special system of Blanket Module Connections (BMCs) comprising flexible supports, insulating key pads and electrical straps. BMCs fix the modules relative to the VV and manage the current flow. They accommodate transient, cyclic, thermal and electro-magnetic (EM) loads in a vacuum environment and under neutron radiation. Dynamic, thermal-structural and strength analyses have been performed in support of the BMC design and the results have been experimentally confirmed. The components with uncertain behavior including partially and non-preloaded threads, insulation coating, and electrical contacts were designed by experiments. The effort to develop a reliable and robust design of the BMCs in time for the signature of the Procurement Arrangement on BMCs between ITER Organization and Russian Federation in late 2014 spanned several years. It includes design and analysis as well as experimental activities by the ITER Organization and by JSC “NIKIET” (Russia), which, as an affirmed subcontractor will manufacture and supply BMCs to the ITER site. This paper summarizes the overall effort focusing in particular on the more recent PA supporting activities.

  2. ITER blanket module connectors. Design, analysis and testing for procurement arrangement

    Energy Technology Data Exchange (ETDEWEB)

    Khomiakov, S., E-mail: khomias58@mail.ru [Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Str. 2/8, Moscow (Russian Federation); Poddubnyi, I.; Kolganov, V.; Zhmakin, A.; Parshutin, E.; Danilov, I.; Strebkov, Yu.; Skladnov, K.; Vlasov, D.; Cheburova, A. [Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Str. 2/8, Moscow (Russian Federation); Romannikov, A. [Institution “Project Center ITER”, 123098, Academic Kurchatov' s Sq.,1, Moscow (Russian Federation); Raffray, R.; Egorov, K.; Chappuis, Ph.; Sadakov, S.; Calcagno, B.; Roccella, R. [ITER Organization, Route de Vinon sur Verdon, 13067 St. Paul-Lez-Durance (France)

    2016-11-01

    Highlights: • Procurement Arrangement on Blanket Module Connections (BMC) was signed by ITER Organization and Russian Federation Domestic Agency in late 2014. • “N.A. Dollezhal Research and Development Institute of Power Engineering” (NIKIET) was selected as a general supplier of BMC. • NIKIET plays a key role in design development, analytical and experimental justification and manufacturing of BMC. • NIKIET shall fabricate, test and deliver to ITER 2109 flexible supports, 2561 pads, 1053 electrical straps and 1053 pedestals. - Abstract: A standard ITER Blanket module (BM) is attached to the Vacuum Vessel (VV) with a special system of Blanket Module Connections (BMCs) comprising flexible supports, insulating key pads and electrical straps. BMCs fix the modules relative to the VV and manage the current flow. They accommodate transient, cyclic, thermal and electro-magnetic (EM) loads in a vacuum environment and under neutron radiation. Dynamic, thermal-structural and strength analyses have been performed in support of the BMC design and the results have been experimentally confirmed. The components with uncertain behavior including partially and non-preloaded threads, insulation coating, and electrical contacts were designed by experiments. The effort to develop a reliable and robust design of the BMCs in time for the signature of the Procurement Arrangement on BMCs between ITER Organization and Russian Federation in late 2014 spanned several years. It includes design and analysis as well as experimental activities by the ITER Organization and by JSC “NIKIET” (Russia), which, as an affirmed subcontractor will manufacture and supply BMCs to the ITER site. This paper summarizes the overall effort focusing in particular on the more recent PA supporting activities.

  3. Prospects of ceramic tritium breeder materials

    International Nuclear Information System (INIS)

    Roth, E.; Roux, N.; Conservatoire National des Arts et Metiers; CEA Centre d'Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette

    1989-01-01

    In this paper the authors examine the prospects of the main ceramics proposed as breeder materials for fusion reactors, i.e. Li-2O, Li-2ZrO-3, LiAlO-2, Li-4SiO-4. To do so they review terms of reference of contemplated blankets for NET, ITER and DEMO, and the proposed blanket concepts and materials. Issues respective to the use of each breeder material are examined, and from this review it is concluded that ceramics are the most favorable breeder materials whose use can be contemplated as well for a driver blanket for NET or ITER and for a DEMO blanket. Ceramics are then compared between themselves and it is seen that, subject to the confirmation of recent experimental results, lithium zirconate could be used with advantage in any of the present blanket concepts, except in those employing lithium at its natural isotopic abundance, in which case only Li-2O can be used. However in specific cases, or in parts of a blanket, other ceramics may be profitably employed. As a general conclusion suggestions are made to further improve ceramic breeder performances, and it is recommended to intensify also work on problems that have to be solved in order to operate ceramic breeder blankets e.g. tritium extraction and recovery systems and conditions of beryllium use. (author). 37 refs.; 12 tabs

  4. Adaptation of the HCPB DEMO TBM as breeding blanket for ITER : Neutronic and thermal analyses

    International Nuclear Information System (INIS)

    Aquaro, D.; Morellini, D.; Cerullo, N.

    2006-01-01

    Two breeding blanket are presently developed in Europe for the DEMO reactor: the first one, the Helium Cooled Lithium Lead (HCLL) uses a liquid breeder while the other , the Helium Cooled Pebble Bed (HCPB), uses a solid breeder in form of pebble bed. The modules of these blankets, called Test Blanket Modules (TBM) will be located in correspondence of the equatorial ports of ITER in order to be tested. ITER FEAT was designed with shielding blankets, therefore in the final stage of the experiment, in the foreseen tritium -deuterium operation phase, the tritium will be supplied to the reactor and not produced inside it. Since the production of tritium is of main importance for the feasibility of a nuclear fusion reactor, perhaps in the ITER final stage, the shielding blanket could be substituted by means of a breeding blanket. The geometry and composition of this breeding blanket would be, of course, similar to that of TBM which demonstrated to have the best performances. This paper illustrates a neutronic and thermal analysis of an hypothetical triziogen blanket for ITER FEAT made similar to a HCPB test module. The main aims of the performed analyses are to determine the Tritium Breeding Ratio (TBR) considering different solid breeders (Li 4 SiO 4 and Li 2 TiO 3 ) with different enrichment in 6 Li and different structural materials (a 9%CRWVTa reduced activation ferritic martensitic steel (EUROFER) or ceramic matrix composites like SiCf/SiC). The breeding blanket design is compared considering the highest value of TBR and the verification of the temperature constraints ( 550 o C for the steel, 950 o C for the breeder and 650 o C for the Beryllium). The neutronic analyses have been performed by means of MCNP-4C code and the thermal analyses using the MSC-MARC code. A TBR about equal 1 was obtained with a SiCf/SiC structural material and a Li 4 SiO 4 breeder. The performed analyses have to be considered preliminary and an academic exercise, nevertheless they could give

  5. Thermodynamic considerations for the use of vanadium alloys with ceramic breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, C.E.; Johnson, I.; Kopasz, J.P.

    1995-12-31

    Fusion energy is considered to be an attractive energy form because of its minimal environmental impact. In order to maintain this favorable status, every effort needs to be made to use low activation materials wherever possible. The tritium breeder blanket is a focal point of system design engineers who must design environmentally attractive blankets through the use of low activation materials. Of the several candidate lithium-containing ceramics being considered for use in the breeder blanket, Li{sub 2}O, Li{sub 2}TiO{sub 3}, are attractive choices because of their low activation. Also, low activation materials like the vanadium alloys are being considered for use as structural materials in the blanket. The suitability of vanadium alloys for containment of lithium ceramics is the subject of this study. Thermodynamic evaluations are being used to estimate the compatibility and stability of candidate ceramic breeder materials (Li{sub 2}O, Li{sub 2}TiO{sub 3}, and Li{sub 2}ZrO{sub 3}) with vanadium and vanadium alloys. This thermodynamic evaluation will focus first on solid-solid interactions. As a tritium breeding blanket will use a purge gas for tritium recovery, gas-solid systems will also receive attention.

  6. Thermodynamic considerations for the use of vanadium alloys with ceramic breeder materials

    International Nuclear Information System (INIS)

    Johnson, C.E.; Johnson, I.; Kopasz, J.P.

    1995-01-01

    Fusion energy is considered to be an attractive energy form because of its minimal environmental impact. In order to maintain this favorable status, every effort needs to be made to use low activation materials wherever possible. The tritium breeder blanket is a focal point of system design engineers who must design environmentally attractive blankets through the use of low activation materials. Of the several candidate lithium-containing ceramics being considered for use in the breeder blanket, Li 2 O, Li 2 TiO 3 , are attractive choices because of their low activation. Also, low activation materials like the vanadium alloys are being considered for use as structural materials in the blanket. The suitability of vanadium alloys for containment of lithium ceramics is the subject of this study. Thermodynamic evaluations are being used to estimate the compatibility and stability of candidate ceramic breeder materials (Li 2 O, Li 2 TiO 3 , and Li 2 ZrO 3 ) with vanadium and vanadium alloys. This thermodynamic evaluation will focus first on solid-solid interactions. As a tritium breeding blanket will use a purge gas for tritium recovery, gas-solid systems will also receive attention

  7. Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

    Energy Technology Data Exchange (ETDEWEB)

    Scaller, K; Vrillon, B

    1980-02-01

    Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component.

  8. Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

    International Nuclear Information System (INIS)

    Scaller, K.; Vrillon, B.

    1980-01-01

    Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component

  9. IAMBUS, a computer code for the design and performance prediction of fast breeder fuel rods

    International Nuclear Information System (INIS)

    Toebbe, H.

    1990-05-01

    IAMBUS is a computer code for the thermal and mechanical design, in-pile performance prediction and post-irradiation analysis of fast breeder fuel rods. The code deals with steady, non-steady and transient operating conditions and enables to predict in-pile behavior of fuel rods in power reactors as well as in experimental rigs. Great effort went into the development of a realistic account of non-steady fuel rod operating conditions. The main emphasis is placed on characterizing the mechanical interaction taking place between the cladding tube and the fuel as a result of contact pressure and friction forces, with due consideration of axial and radial crack configuration within the fuel as well as the gradual transition at the elastic/plastic interface in respect to fuel behavior. IAMBUS can be readily adapted to various fuel and cladding materials. The specific models and material correlations of the reference version deal with the actual in-pile behavior and physical properties of the KNK II and SNR 300 related fuel rod design, confirmed by comparison of the fuel performance model with post-irradiation data. The comparison comprises steady, non-steady and transient irradiation experiments within the German/Belgian fuel rod irradiation program. The code is further validated by comparison of model predictions with post-irradiation data of standard fuel and breeder rods of Phenix and PFR as well as selected LWR fuel rods in non-steady operating conditions

  10. Engineering design of a direct-cycle steam-generating blanket for a long-pulse fusion reactor

    International Nuclear Information System (INIS)

    Cort, G.E.; Hagenson, R.L.; Teasdale, R.W.; Fox, W.E.; Soran, P.D.; Cullingford, H.S.; Bathke, C.G.; Krakowski, R.A.

    1979-01-01

    A comprehensive neutronics, thermohydraulic, and mechanical design of a tritium-breeding blanket for use by a conceptual long-pulse Reversed-Field Pinch Reactor (RFPR) is described. On the basis of constraints imposed by cost and the desire to use existing technology, a direct-cycle steam system and stainless-steel construction were used. For reasons of plasma stability, the RFPR blanket supports a 20-mm-thick copper first wall. Located behind the 1.5-m-radius first wall is a 0.50-m-thick stainless-steel blanket containing a granular bed of Li 2 O through which flows low-pressure helium (0.1 MPa) for tritium extraction. Water/steam tubes radially penetrate this packed bed. The large thermal capacity and low thermal diffusivity of the Li 2 O blanket are sufficient to maintain a nearly constant temperature during the approx. 25-s burn period

  11. Thermal-hydraulic and neutronic considerations for designing a lithium-cooled tokamak blanket

    International Nuclear Information System (INIS)

    Chao, J.; Mikic, B.; Todreas, N.

    1978-12-01

    A methodology for the design of lithium cooled blankets is developed. The thermal-hydraulics, neutronics and interactions between them are extensively investigated. In thermal hydraulics, two models illustrate the methodology used to obtain the acceptable ranges for a set of design parameters. The methodology can be used to identify the limiting constraints for a particular design. A complete neutronic scheme is set up for the calculations of the volumetric heating rate as a function of the distance from the first wall, the breeding ratio as a function of the amount of structural material in the blanket, and the radiation damage in terms of atom displacements and gas production rate. Different values of the volume percent of Type-316 stainless steel are assigned in four breeding zones to represent a nonuniformly distributed structural material which satisfies various thermal-hydraulic requirements. The role that the radiation damage plays in the overall design methodology is described. The product of the first wall lifetime and neutron loading is limited by the radiation damage which degrades the mechanical properties of the material

  12. Establishment of design and fabrication technology and domestic qualification for ITER blanket system

    International Nuclear Information System (INIS)

    Hong, Bong Guen; In, S. R.; Bae, Y. D.

    2006-02-01

    To obtain and analyze the detailed design and manufacturing technology of the blanket system for each components, the related data are collected through the various sources. And also, design processes and results of the FWs, shield blocks, and TBMs are investigated. From these analysis of the blanket R and D status of each party, we develop the KO R and D plan and it is used in the selection of manufacturing method and the materials. For the ITA16-10 subtask1, we had the official agreement with ITER IT in December 2004 for the qualification of the FW panel fabrication methods and to establish the NDT methods for the FW panel. From the technical reports we published, we compare the manufacturing methods and the proposed material for each component according to the parties. Be is proposed as a plasma facing material and most parties have interest in S-65C. Cu alloy is proposed as a heat sink material and DSCu or CuCrZr are investigated now. For the structural material, stainless steel such as SS316L(N) is investigated internationally. HIP and brazing are proposed as the manufacturing methods. In order to establish the blanket system technology, design contents of shield block by ITER IT and other parties were investigated through participating the international workshop and meeting, dispatching the researcher to the ITER IT or other parties to collect the drafting and 3D modeling files. The modification items of blanket design were investigated and a researcher was dispatched in the ITER IT and participated in the analysis on cooling problem in shield block such as front header and drilled manifold. To investigate the development status of TBM, we participated the 14th TBWG meeting and proposed the KO HCSB and HCML as candidates. And also, we obtain the R and D results of other parties and make document about the R and D status of other parties for the TBM. Finally, we establish the KO TBM R and D plan and proposed it to ITER IT and other parties. In which, the

  13. Establishment of design and fabrication technology and domestic qualification for ITER blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; In, S. R.; Bae, Y. D. (and others)

    2006-02-15

    To obtain and analyze the detailed design and manufacturing technology of the blanket system for each components, the related data are collected through the various sources. And also, design processes and results of the FWs, shield blocks, and TBMs are investigated. From these analysis of the blanket R and D status of each party, we develop the KO R and D plan and it is used in the selection of manufacturing method and the materials. For the ITA16-10 subtask1, we had the official agreement with ITER IT in December 2004 for the qualification of the FW panel fabrication methods and to establish the NDT methods for the FW panel. From the technical reports we published, we compare the manufacturing methods and the proposed material for each component according to the parties. Be is proposed as a plasma facing material and most parties have interest in S-65C. Cu alloy is proposed as a heat sink material and DSCu or CuCrZr are investigated now. For the structural material, stainless steel such as SS316L(N) is investigated internationally. HIP and brazing are proposed as the manufacturing methods. In order to establish the blanket system technology, design contents of shield block by ITER IT and other parties were investigated through participating the international workshop and meeting, dispatching the researcher to the ITER IT or other parties to collect the drafting and 3D modeling files. The modification items of blanket design were investigated and a researcher was dispatched in the ITER IT and participated in the analysis on cooling problem in shield block such as front header and drilled manifold. To investigate the development status of TBM, we participated the 14th TBWG meeting and proposed the KO HCSB and HCML as candidates. And also, we obtain the R and D results of other parties and make document about the R and D status of other parties for the TBM. Finally, we establish the KO TBM R and D plan and proposed it to ITER IT and other parties. In which, the

  14. Limitations on blanket performance

    International Nuclear Information System (INIS)

    Malang, S.

    1999-01-01

    The limitations on the performance of breeding blankets in a fusion power plant are evaluated. The breeding blankets will be key components of a plant and their limitations with regard to power density, thermal efficiency and lifetime could determine to a large degree the attractiveness of a power plant. The performance of two rather well known blanket concepts under development in the frame of the European Blanket Programme is assessed and their limitations are compared with more advanced (and more speculative) concepts. An important issue is the question of which material (structure, breeder, multiplier, coatings) will limit the performance and what improvement would be possible with a 'better' structural material. This evaluation is based on the premise that the performance of the power plant will be limited by the blankets (including first wall) and not by other components, e.g. divertors, or the plasma itself. However, the justness of this premise remains to be seen. It is shown that the different blanket concepts cover a large range of allowable power densities and achievable thermal efficiencies, and it is concluded that there is a high incentive to go for better performance in spite of possibly higher blanket cost. However, such high performance blankets are usually based on materials and technologies not yet developed and there is a rather high risk that the development could fail. Therefore, it is explained that a part of the development effort should be devoted to concepts where the materials and technologies are more or less in hand in order to ensure that blankets for a DEMO reactor can be developed and tested in a given time frame. (orig.)

  15. Current Design of the Flange Type Hydrogen Permeation Sensor in Liquid Breeder

    International Nuclear Information System (INIS)

    Lee, E. H.; Jin, H. G.; Yoon, J. S.; Kim, S. K.; Lee, D. W.; Lee, H. G.

    2015-01-01

    In 2004, A. Ciampichetti et al. proposed a hollow capsule shape permeation sensor and they theoretically and experimentally evaluated the performance of the sensor made of Nb membrane at test condition of 500 .deg. C. However, the evaluation result showed the measured hydrogen permeation flux in the sensor much lower than the predicted one and they concluded that, the result is due to the formation of an oxide layer on the sensor membrane surface. Three years later, A. Ciampichetti et al. observed that a hollow capsule shape permeation sensor has too long response time to measure hydrogen concentration in liquid breeder. However, they suggested optimizing the sensor geometry with the reduction of the ratio 'total sensor volume/permeation surface' to overcome the low hydrogen permeating flux. For development of the liquid breeding technologies in nuclear fusion, the permeation sensor to measure tritium concentration in liquid metal breeder has been developed. Lee et al. proposed a flange type permeation sensor to dramatically reduce the ratio sensor 'inside volume/permeation surface' and to remove membrane welding during sensor manufacture process. However, the flange type sensor has problem with sealing. In present study, the modified flange sensor design with a metallic C-ring spring gasket is introduced. The modified sensor will be verified and evaluated under high temperature conditions by end of 2015

  16. Current Design of the Flange Type Hydrogen Permeation Sensor in Liquid Breeder

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    In 2004, A. Ciampichetti et al. proposed a hollow capsule shape permeation sensor and they theoretically and experimentally evaluated the performance of the sensor made of Nb membrane at test condition of 500 .deg. C. However, the evaluation result showed the measured hydrogen permeation flux in the sensor much lower than the predicted one and they concluded that, the result is due to the formation of an oxide layer on the sensor membrane surface. Three years later, A. Ciampichetti et al. observed that a hollow capsule shape permeation sensor has too long response time to measure hydrogen concentration in liquid breeder. However, they suggested optimizing the sensor geometry with the reduction of the ratio 'total sensor volume/permeation surface' to overcome the low hydrogen permeating flux. For development of the liquid breeding technologies in nuclear fusion, the permeation sensor to measure tritium concentration in liquid metal breeder has been developed. Lee et al. proposed a flange type permeation sensor to dramatically reduce the ratio sensor 'inside volume/permeation surface' and to remove membrane welding during sensor manufacture process. However, the flange type sensor has problem with sealing. In present study, the modified flange sensor design with a metallic C-ring spring gasket is introduced. The modified sensor will be verified and evaluated under high temperature conditions by end of 2015.

  17. Combined system of accelerator molten-salt breeder (AMSB) apd molten-salt converter reactor (MSCR)

    International Nuclear Information System (INIS)

    Furukawa, K.; Kato, Y.; Ohmichi, T.; Ohno, H.

    1983-01-01

    A design and research program is discUssed of the development of accelerator molten-salt breeder (AMSB) consisting of a proton accelerator and a molten fluoride target. The target simultaneously serves as a blanket for fissionable material prodUction. An addition of some amoUnt of fissile nuclides to a melt expands the AMSB potentialities as the fissionable material production increases and the energy generation also grows up to the level of self-provision. Besides the blanket salts may be used as nuclear fuel for molten-salt converter reactor (MSCR). The combined AM SB+MSCR system has better parameters as compared to other breeder reactors, molten-salt breeder reactors (MSBR) included

  18. ITER structural design criteria and their extension to advanced reactor blankets

    International Nuclear Information System (INIS)

    Majumdar, S.; Kalinin, G.

    2000-01-01

    Applications of the recent ITER structural design criteria (ISDC) are illustrated by two components. First, the low-temperature-design rules are applied to copper alloys that are particularly prone to irradiation embrittlement at relatively low fluences at certain temperatures. Allowable stresses are derived and the impact of the embrittlement on allowable surface heat flux of a simple first-wall/limiter design is demonstrated. Next, the high-temperature-design rules of ISDC are applied to evaporation of lithium and vapor extraction (EVOLVE), a blanket design concept currently being investigated under the US Advanced Power Extraction (APEX) program. A single tungsten first-wall tube is considered for thermal and stress analyses by finite-element method

  19. Simplified inelastic analysis methods applied to fast breeder reactor core design

    International Nuclear Information System (INIS)

    Abo-El-Ata, M.M.

    1978-01-01

    The paper starts with a review of some currently available simplified inelastic analysis methods used in elevated temperature design for evaluating plastic and thermal creep strains. The primary purpose of the paper is to investigate how these simplified methods may be applied to fast breeder reactor core design where neutron irradiation effects are significant. One of the problems discussed is irradiation-induced creep and its effect on shakedown, ratcheting, and plastic cycling. Another problem is the development of swelling-induced stress which is an additional loading mechanism and must be taken into account. In this respect an expression for swelling-induced stress in the presence of irradiation creep is derived and a model for simplifying the stress analysis under these conditions is proposed. As an example, the effects of irradiation creep and swelling induced stress on the analysis of a thin walled tube under constant internal pressure and intermittent heat fluxes, simulating a fuel pin, is presented

  20. Design and development of a zoom lens objective for the fast breeder test reactor periscope

    International Nuclear Information System (INIS)

    Das, N.C.; Udupa, D.V.; Shukla, R.P.

    2003-10-01

    A three lens optically compensated zoom lens useful for the 5 meter long periscope in the Fast Breeder Test Reactor (FBTR) has been designed, fabricated and tested. The zoom lens fabricated using radiation resistant glasses has a zoom ratio of 2.5 with a focal length range of l00 mm to 250 mm. The zoom lens objective has been designed for viewing the objects kept at a distance in the range of 1.5 m to 3 m from the objective lens. It is found that the zoom lens objective can be used for resolving objects with a linear resolution of 0.2 mm inside the reactor when viewed with an eye piece of focal length 50 mm. (author)

  1. Influence of thermal performance on design parameters of a He/LiPb dual coolant DEMO concept blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Mas de les Valls, E., E-mail: elisabet.masdelesvalls@gits.ws [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Department of Heat Engines, Barcelona (Spain); Batet, L. [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Department of Physics and Nuclear Engineering, Barcelona (Spain); Medina, V. de [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Sediment Transport Research Group, Department of Engineering Hydraulic, Marine and Environmental Engineering, Barcelona (Spain); Fradera, J. [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Department of Physics and Nuclear Engineering, Barcelona (Spain); Sanmarti, M. [bFUS-IREC, Jardins de les Dones de Negre 1, 08930 Sant Adria del Besos (Spain); Sedano, L.A. [EURATOM-CIEMAT Association, 28040 Madrid (Spain)

    2012-08-15

    Spanish Breeding Blanket Technology Programme TECNO{sub F}US is exploring the technological capabilities of a Dual-Coolant He/Pb15.7Li breeding blanket for DEMO and studying new breeding blanket design specifications. The progress of the channel conceptual design is being conducted in parallel with the extension of MHD computational capabilities of CFD tools and the underlying physics of MHD models. A qualification of MHD effects under present blanket design specifications and some approaches to their modelling were proposed by the authors in . The analysis was accomplished with the 2D transient algorithm from Sommeria and Moreau and implemented in the OpenFOAM CFD toolbox . The thermal coupling was implemented by means of the Boussinesq hypothesis. Previous analyses showed the need of improvement of FCI thickness and thermal properties in order to obtain a desirable liquid metal temperature gain of 300 Degree-Sign C. In the present study, an assessment through sensitivity and parametric analyses of the required FCI thickness is performed. Numerical simulations have been carried out considering a Robin-type thermal boundary condition which assumes 1D steady state thermal balance across the solid FCI and Eurofer layers. Such boundary condition has been validated with a fluid-solid coupled domain analysis. Results for the studied flow conditions and channel dimensions show that, in order to obtain a liquid metal temperature gain of about 300 Degree-Sign C, the required FCI material should have a very small effective heat transfer coefficient ((k/{delta}) {<=} 1 W/m{sup 2}K) and fluid velocities should be about 0.2 m/s or less. Moreover, special attention has to be placed on the temperature difference across the FCI layer. However, for a maximised liquid metal thermal gain, higher velocities would be preferable, what would also imply a reduced temperature difference across the FCI layer.

  2. Fusion-driven sub-critical dual-cooled waste transmutation blanket: design and analysis

    International Nuclear Information System (INIS)

    Wang Weihua; Wu Yican; Ke Yan; Kang Zhicheng; Wang Hongyan; Huang Qunying

    2003-01-01

    The Fusion-Driven Sub-critical System (FDS) is one of the Chinese programs to be further developed for fusion application. Its Dual-cooled Waste Transmutation Blanket (DWTB), as one the most important part of the FDS is cooled by helium and liquid metal, and have the features of safety, tritium self-sustaining, high efficiency and feasibility. Its conceptual design has been finished. This paper is mainly involved with the basic structure design and thermal-hydraulics analysis of DWTB. On the basis of a three-dimensional (3-D) model of radial-toroidal sections of the segment box, thermal temperature gradients and structure analysis made with a comprehensive finite element method (FEM) have been performed with the computer code ANSYS5.7 and computational fluid dynamic finite element codes. The analysis refers to the steady-state operating condition of an outboard blanket segment. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions have been also taken into account. All the above loads have been combined as an input for a FEM stress analysis and the resulting stress distribution has been evaluated. Finally, the structure design and Pb-17Li flow velocity has been optimized according to the calculations and analysis

  3. Detail Design of the hydrogen system and the gas blanketing system for the HANARO-CNS

    International Nuclear Information System (INIS)

    Choi, Jung Woon; Kim, Hark Rho; Kim, Young Ki; Wu, Sang Ik; Kim, Bong Su; Lee, Yong Seop

    2007-04-01

    The cold neutron source (CNS), which will be installed in the vertical CN hole of the reflector tank at HANARO, makes thermal neutrons to moderate into the cold neutrons with the ranges of 0.1 ∼ 10 meV passing through a moderator at about 22K. A moderator to produce cold neutrons is liquid hydrogen, which liquefies by the heat transfer with cryogenic helium flowing from the helium refrigeration system (HRS). Because of its installed location, the hydrogen system is designed to be surrounded by the gas blanketing system to notify the leakage on the system and to prevent hydrogen leakage out of the CNS. The hydrogen system, consisted of hydrogen charging unit, hydrogen storage unit, hydrogen buffer tank, and hydrogen piping, is designed to smoothly and safely supply hydrogen to and to draw back hydrogen from the IPA of the CNS under the HRS operation mode. Described is that calculation for total required hydrogen amount in the CNS as well as operation schemes of the hydrogen system. The gas blanketing system (GBS) is designed for the supply of the compressed nitrogen gas into the air pressurized valves for the CNS, to isolate the hydrogen system from the air and the water, and to prevent air or water intrusion into the vacuum system as well as the hydrogen system. All detail descriptions are shown inhere as well as the operation scheme for the GBS

  4. A conceptual composite blanket design for the Tokamak type of thermonuclear reactor incorporating thermoelectric pumping of liquid lithium

    International Nuclear Information System (INIS)

    Dutta Gupta, P.B.

    1981-01-01

    The conceptual liquid lithium blanket design for the tokamak type of thermonuclear reactor put forward is a modification of the initial simple but novel design concept enunciated earlier that exploits the availability of suitably oriented magnetic fields and temperature gradients within the blanket to pump the liquid as has been shown feasible by laboratory model experiments. The modular construction of the blanket cells is retained but the earlier simple back to back double spiralling channel module is replaced by a composite unit of three radially nested layer-structures to optimise heat and tritium extraction from the blanket. The layer-structure at the first wall generates liquid lithium circulation by thermoelectric magnetohydrodynamic forces and the segregated double spiralling channels serve as inlet-outlet driving devices. The outermost layer-structure is cooled by helium. Liquid lithium in the intermediate layer-structure is pumped at a very slow rate. The choice of the relative dimensional proportions of the three layer-structure and the channel cross-section, material property and the spiralling contour is of critical importance for the design. This paper presents the design data for a conceptual design of such a blanket with a 5000 MW (th) rating. (author)

  5. Methods to enhance blanket power density

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Miller, L.G.; Bohn, T.S.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Wessol, D.E.; Abdou, M.A.

    1982-06-01

    The overall objective of this task is to investigate the extent to which the power density in the FED/INTOR breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow advanced reactor blanket modules to be tested on FED/INTOR under representative conditions

  6. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

    Ozawa, Yoshihiro; Uda, Tatsuhiko; Maki, Koichi.

    1993-01-01

    The present invention provides a blanket of a thermonuclear device which produces tritium fuels consumed in plasmas while converting neutrons generated in the plasmas into heat energy. That is, zirconium is coated to at least one of neutron breeder pebbles and breeder pebbles, to suppress reaction between them by being in direct contact with each other at a high temperature. Further, fins are attached to a cooling pipe at a pitch smaller than the diameter of both of the pebbles, to prevent direct contact at whole surface of the pebbles and the cooling pipe, which would lower a temperature excessively. The length of the fin is controlled to control the thickness of a helium gas gap. With such constitution, direct contact of neutron breeder pebbles and the breeder pebble which are to be filled and mixed, and tend to react at a high temperature, can be prevented. The temperature of the breeding blanket is reliably prevented from lowering below a tritium emitting temperature. The structure is simplified and the production is facilitated. (I.S.)

  7. Thermal response of a pin-type fusion reactor blanket during steady and transient reactor operation

    International Nuclear Information System (INIS)

    Grotz, S.; Ghoniem, N.M.

    1986-02-01

    The thermal analysis of the blanket examines both the steady-state and transient reactor operations. The steady-state analysis covers full power and fractional power operation whereas the transient analysis examines the effects of power ramps and blanket preheat. The blanket configuration chosen for this study is a helium cooled solid breeder design. We first discuss the full power, steady-state temperature fields in the first wall, beryllium rods, and breeder rods. Next we examine the effects of fractional power on coolant flow and temperature field distributions. This includes power plateaus of 10%, 20%, 50%, 80%, and 100% of full power. Also examined are the restrictions on the rates of power ramping between plateaus. Finally we discuss the power and time requirements for pre-heating the primary from cold iron conditions up to startup temperature (250 0 C)

  8. Recommended practices in elevated temperature design: A compendium of breeder reactor experiences (1970-1986): An overview

    International Nuclear Information System (INIS)

    Wei, B.C.; Cooper, W.L. Jr.; Dhalla, A.K.

    1987-09-01

    Significant experiences have been accumulated in the establishment of design methods and criteria applicable to the design of Liquid Metal Fast Breeder Reactor (LMFBR) components. The Subcommittee of the Elevated Temperature Design under the Pressure Vessel Research Council (PVRC) has undertaken to collect, on an international basis, design experience gained, and the lessons learned, to provide guidelines for next generation advanced reactor designs. This paper shall present an overview and describe the highlights of the work

  9. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    International Nuclear Information System (INIS)

    Adams, S.R.

    1985-10-01

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences

  10. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    Energy Technology Data Exchange (ETDEWEB)

    Adams, S.R.

    1985-10-01

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

  11. Japanese contribution to the design of primary module of shielding blanket in ITER-FEAT

    International Nuclear Information System (INIS)

    Kuroda, Toshimasa; Hatano, Toshihisa; Miki, Nobuharu; Hiroki, Seiji; Enoeda, Mikio; Ohmori, Junji; Akiba, Masato

    2003-02-01

    Japanese contributions to the design activity on the shielding blanket module consisting of the separable first wall and the shield block for ITER-FEAT are compiled. Temperature and stress distributions in the first wall and the shield block are analyzed and evaluated with 2-D and 3-D models for steady state and also for transient condition according to plasma ramp-up and ramp-down. While temperatures and stresses in the first wall satisfy their allowable values, those in a front part of the shield block exceed the allowable guideline. Based on this result, design improvements are suggested. Coolant flow and pressure distributions along the complicated coolant channel in the shield block are preliminary analyzed. Though heat removal is satisfactory in all coolant channels, back flows due to choking in coolant collectors are found. Design improvements to avoid the choking are suggested. Electromagnetic forces acting on blanket modules are analyzed with detailed 3-D models of solid elements for different disruption scenarios. The maximum moment around radial axis is 1.36 MNm on module no.5 under fast upward VDE, and the maximum moment around vertical axis is 1.47 MNm on module no.1 under fast downward VDE. The supporting beam of the first wall with welded attachment to the shield block is designed. Required welding thickness and support conditions to withstand electromagnetic forces are estimated. Strength of the shield block at the region mating the flexible cartridge is also estimated. Though the shield block surface attached by the flexible cartridge shows sufficient strength, the internal thread mating the Inconel bolt would need more length. In addition, water-to-water leak detection system in case main supply/return manifolds are located within the vacuum vessel is designed. By using Kr as the tracer material, the possibility of water-to-water leak detection and the concept of the detection system are shown. The design of the shielding blanket of ITER-FEAT has

  12. Fusion energy for alternate applications: the design of a high temperature falling bed as a long-lived blanket

    International Nuclear Information System (INIS)

    Harkness, S.D.; Stevens, H.C.; Hall, M.M.; Gohar, M.Y.A.; de Paz, J.F.

    1979-01-01

    The high temperature falling bed conceptual design work has consisted of a coordinated effort in neutronics, materials science, thermal hydraulics and mechanical design. The neutronics work has been based on a one-dimensional transport analysis and has been used to scope the implication of blanket dimensions, breeding materials, ceramic pebble material and coolant choice on both tritium breeding capabilities and energy deposition into the high temperature region of the blanket. The materials science effort has concentrated on defining the selection of a particular ceramic material. The thermal hydraulic analysis has been concerned with sizing the heat transfer system and defining the temperature gradients in the high temperature blanket. The mechanical design work has evaluated how such a system might be constructed from the point of view of maintainability and structural support

  13. A new physics design of control safety rods for prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Devan, K.; Riyas, A.; Alagan, M.; Mohanakrishnan, P.

    2008-01-01

    The absorber rods of 500 MWe prototype fast breeder reactor (PFBR), which is under construction at Kalpakkam, have been designed to provide sufficient shutdown margin during normal and accidental conditions for ensuring the safe shut down. There are nine control and safety rods (CSR) and 3 diverse safety rods (DSR). Absorber material used is initially 65% enriched B 4 C. Based on the reported experiments in PHENIX reactor and design of absorber rods in SUPERPHENIX, the design of CSR is modified by introducing 20 cm length natural B 4 C at the top and bottom of absorber column and maintaining the remaining portion with 65% enriched B 4 C. This design ensures sufficient shutdown margin (SDM) during normal operation and also during the one stuck rod condition. For comparison of the above two designs, a CSR of 57% of enrichment was considered which gives the same worth as the revised CSR design with natural B 4 C sections in top and bottom. There is significant savings in the initial inventory of enriched B 4 C for CSR. The annual requirement of enriched boron also reduces. This new CSR can last for about 5 cycles, based on its clad life. But, it is planned to be replaced after every 3 cycles (1 cycle equals 180 efpd) of operation due to radiation damage effects in hexcan D9 steel. Use of ferritic steel for hexcan can extend the life of CSR to 5 cycles

  14. Thermal hydraulic and power cycle analysis of liquid lithium blanket designs

    International Nuclear Information System (INIS)

    Misra, B.; Stevens, H.C.; Maroni, V.A.

    1977-01-01

    Thermal hydraulic and power cycle analyses were performed for the first-wall and blanket systems of tokamak-type fusion reactors under a typical set of design and operating conditions. The analytical results for lithium-cooled blanket cells show that with stainless steel as construction material and with no divertor present, the maximum allowable neutron wall loading is approximately 2 MW/m 2 and is limited by thermal stress criteria. With vanadium alloy as construction material and no divertor present, the maximum allowable neutron wall loading is approximately 8 MW/m 2 and is limited by an interplay of constraints imposed on the maximum allowable structural temperature and the minimum allowable coolant inlet temperature. With a divertor these wall loadings can be increased by from 40 to 90 percent. The cost of the vanadium system is found to be competitive with the stainless steel system because of the higher allowable structural temperatures and concomitant higher thermal efficiencies afforded by the vanadium alloys

  15. Fluidized-bed design for ICF reactor blankets using solid-lithium compounds

    International Nuclear Information System (INIS)

    Sucov, E.W.; Malick, F.S.; Green, L.; Hall, B.O.

    1983-01-01

    A fluidized-bed concept for blankets of dry or wetted first-wall ICF reactors using solid-lithium compounds is described. The reaction chamber is a right cylinder, 32 m high and 20 m in diameter; the blanket is composed of 36 steel tanks, 32 m high, which carry the sintered Li 2 O particles in the fluidizing helium gas. Each tank has a radial thickness of 2 m which generates a tritium breeding ration (TBR) of 1.27 and absorbs over 98% of the neutron energy; reducing the thickness to 1.2 m produces a TBR of 1.2 and energy absorption of 97% which satisfy the design goals. Calculations of tritium diffusion through the grains and heat removal from the grains showed that neither could be removed by the carrier gas; tritium and heat are therefore removed by removing the grains themselves by varying the helium flow rate. The particles are continuously fed into the bottom of the tanks at 300 0 C and removed at the top at 475 0 C. Tritium and heat extraction are easily and conveniently done outside the reactor

  16. A design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

    International Nuclear Information System (INIS)

    Kobayashi, Noboru; Ogawa, Takashi; Ohki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari

    2009-01-01

    The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650degC, and the maximum fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40. (author)

  17. Design of the segment structure and coolant ducts for a fusion reactor blanket and shield

    International Nuclear Information System (INIS)

    Briaris, D.A.; Stanbridge, J.R.

    1978-05-01

    An outline design and analysis of a support structure for the replaceable first wall of a helium cooled fusion reactor blanket has been undertaken. The proposed structure supports all the segment gravitational loads with maximum deflections limited to < 10 mm, and is itself supported off the outer shield by a simple vee-in-groove arrangement. It is a feature of the design that the coaxial coolant pipes and the segment structure operate at the same temperature, making it possible for them to be integrated, thereby avoiding the necessity for pipe bellows. The requirements of cooling the inner arm of the structure and increasing the major radius of the torus by approximately = 0.5 m, have been identified as problems associated with the 'horseshoe' shaped structure applicable to the reactor with divertor. For a ring structure, i.e. reactor without divertor, these problems do not arise. (author)

  18. Thermal-hydraulic investigations on the CEA-ENEA DEMO relevant helium cooled poloidal blanket

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Polazzi, G.; Vallette, F.; Proust, E.; Eid, M.

    1994-01-01

    The CEA-ENEA design of an Helium Cooled Solid Breeder Blanket (HCSBB) for the DEMO reactor, with a breeder in tube (BIT) poloidal arrangement, is based on the use of lithium ceramic pellets, the ENEA γ-LiAlO 2 or the CEA Li 2 ZrO 3 . Due to the geometry of the DEMO reactor plasma chamber, these breeder bundles are adapted to the Vacuum Vessel with a strong poloidal curvature. This curvature influences the thermal-hydraulic behaviour of the coolant flowing inside the bundle. The paper presents the CEA-ENEA first results of the experimental and theoretical programme, aiming at optimizing the breeder module thermal hydraulic design. (author) 6 refs.; 7 figs.; 1 tab

  19. Fusion blanket testing in MFTF-α + T

    International Nuclear Information System (INIS)

    Kleefeldt, K.

    1985-01-01

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

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

    Science.gov (United States)

    Yu, Guanying; Liu, Xufeng; Liu, Songlin

    2016-10-01

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

  1. Convertible shielding to ceramic breeding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Kurasawa, Toshimasa; Sato, Satoshi; Nakahira, Masataka; Togami, Ikuhide; Hashimoto, Toshiyuki; Takatsu, Hideyuki; Kuroda, Toshimasa.

    1995-05-01

    Four concepts have been studied for the ITER convertible blanket: 1)Layered concept 2)BIT(Breeder-Inside-Tube)concept 3)BOT(Breeder-Out of-Tube)concept 4)BOT/mixed concept. All concepts use ceramic breeder and beryllium neutron multiplier, both in the shape of small spherical pebbles, 316SS structure, and H 2 O coolant (inlet/outlet temperatures : 100/150degC, pressure : 2 MPa). During the BPP, only beryllium pebbles (the primary pebble in case of BOT/mixed concept) are filled in the blanket for shielding purpose. Then, before the EPP operation, breeder pebbles will be additionally inserted into the blanket. Among possible conversion methods, wet method by liquid flow seems expecting for high and homogeneous pebble packing. Preliminary 1-D neutronics calculation shows that the BOT/mixed concept has the highest breeding and shielding performance. However, final selection should be done by R and D's and more detail investigation on blanket characteristics and fabricability. Required R and D's are also listed. With these efforts, the convertible blanket can be developed. However, the following should be noted. Though many of above R and D's are also necessary even for non-convertible blanket, R and D's on convertibility will be one of the most difficult parts and need significant efforts. Besides the installation of convertible blanket with required structures and lines for conversion will make the ITER basic machine more complicated. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-22

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

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

    International Nuclear Information System (INIS)

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

    1984-12-01

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

  4. Modelling and design of high compression electron guns for EBIS/T charge breeders

    CERN Document Server

    AUTHOR|(CDS)2087190; Zschornack, G.; Lettry, J.; Wenander, F.

    In this thesis the optimization of the REXEBIS charge breeder at the ISOLDE facility is presented. REXEBIS in its current state provides a current density of 200A/cm² inside the trapping region at 2 T and will be optimized to the physical limit of its design. To overcome this limit a new electron gun, the HEC² gun, was designed in collaboration with the BNL and is in commission at TestEBIS. This electron gun promises a current density of >10 kA/cm², which decreases the charge breeding time significantly. This thesis presents novel simulation techniques supporting the commissioning phase by explaining the sources of occurring loss current and, in addition, evaluate the currently installed collector for compatibility with the HEC2 gun operating at its design limit. The experience gained from the commission of the HEC² gun and the established numerical techniques lead to the development of a smaller high-compression electron gun for medical purposes, the MEDeGUN. This electron gun should provide a high-quali...

  5. Design of a 'two-ion-source' charge breeder with a dual frequency ECR ion source

    International Nuclear Information System (INIS)

    Naik, D.; Naik, V.; Chakrabarti, A.; Dechoudhury, S.; Nayak, S.K.; Pandey, H.K.; Nakagawa, T.

    2005-01-01

    A charge breeder, 'two-ion-source' has been designed which consists of a surface ionisation source followed by an ECR ion source working in two-frequency mode. In this system low charge state ion beam (1+)of radioactive atoms are obtained from the first ion source close to the target chamber and landed into the ECR where those are captured and become high charged state after undergoing a multi ionisation process. This beam dynamics design has been done to optimise the maximum possible transfer of 1 + beam from the first ion source into the ECR, its full capture within the ECR zone and design of an efficient dual frequency ECR. The results shows that 1 + beam of 100 nA and 1μA (A=100) are successfully transmitted and it's beam size at the centre of ECR zone are 12 mm and 21 mm respectively, which are very less than 65 mm width ECR zone of dual frequency ECR heating at 14 GHz and 10 GHz. (author)

  6. A robust helium-cooled shield/blanket design for ITER

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Bourque, R.F.; Baxi, C.B.

    1993-11-01

    General Atomics Fusion and Reactor Groups have completed a helium-cooled, conceptual shield/blanket design for ITER. The configuration selected is a pressurized tubes design embedded in radially oriented plates. This plate can be made from ferritic steel or from V-alloy. Helium leakage to the plasma chamber is eliminated by conservative, redundant design and proper quality control and inspection programs. High helium pressure at 18 MPa is used to reduce pressure drop and enhance heat transfer. This high gas pressure is believed practical when confined in small diameter tubes. Ample industrial experience exists for safe high gas pressure operations. Inboard shield design is highlighted in this study since the allowable void fraction is more limited. Lithium is used as the thermal contacting medium and for tritium breeding, its safety concerns are minimized by a modular, low inventory design that requires no circulation of the liquid metal for the purpose of heat removal. This design is robust, conservative, reliable, and meets all design goals and requirements. It can also be built with present-day technology

  7. Nuclear Analyses of Indian LLCB Test Blanket System in ITER

    Science.gov (United States)

    Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

    2017-04-01

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

  8. Mechanical design and development of analyzing magnet for the RIB charge breeder beam line

    International Nuclear Information System (INIS)

    Bhattacharyya, Pranab; Ghosh, Sundeep; Bhattacharya, Mahuwa; Gupta, Anjan Dutta; Pal, Gautam; Naik, Vaishali

    2015-01-01

    An iron dominated analyzing magnet of uniform peak field 0.6 T has been developed for Radioactive Ion Beam (RIB) charge breeder beam line at VECC, Kolkata. It has two room temperature coils made of copper having channels for passage of cooling water. The other important parts of the magnet includes two yokes, two poles, vacuum chamber, view port, iron plates and the support structures. The most important properties of magnets designed and fabricated for this application is the need for high field quality. The magnet assembly has got overall dimensions of 1.1 metres x 0.91 metres x 0.63 metres and the required field uniformity is 6 x 10 -4 over pole width of ± 2.5 cm. The most critical parameter that has to be maintained to achieve the desired magnetic field is the pole gap of 80 ± 0.25 rom.This paper describes in detail about the mechanical design, coil cooling analysis, development and assembly of this magnet. (author)

  9. Present status of irradiation tests on tritium breeding blanket for fusion reactor

    International Nuclear Information System (INIS)

    Futamura, Yoshiaki; Sagawa, Hisashi; Shimakawa, Satoshi; Tsuchiya, Kunihiko; Kuroda, Toshimasa; Kawamura, Hiroshi.

    1994-01-01

    To develop a tritium breeding blanket for a fusion reactor, irradiation tests in fission reactors are indispensable for obtaining data on irradiation effects on materials, and neutronics/thermal characteristics and tritium production/recovery performance of the blanket. Various irradiation tests have been conducted in the world, especially to investigate tritium release characteristics from tritium breeding and neutron multiplier materials, and materials integrity under irradiation. In Japan, VOM experiments at JRR-2 for ceramic breeders and experiments at JMTR for ceramic breeders and beryllium as a neutron multiplier have been performed. Several universities have also investigated ceramic breeders. In the EC, the EXOTIC experiments at HFR in the Netherlands and the SIBELIUS, the LILA, the LISA and the MOZART experiments for ceramic breeders have carried out. In Canada, NRU has been used for the CRITIC experiments. The TRIO experiments at ORR(ORNL), experiments at RTNS-II, FUBR and ATR have been conducted in the USA. The last two are experiments with high neutron fluence aiming at investigating materials integrity under irradiation. The BEATRIX-I and -II experiments have proceeded under international collaboration of Japan, Canada, the EC and the USA. This report shows the present status of these irradiation tests following a review of the blanket design in the ITER CDA(Conceptual Design Activity). (author)

  10. Shielding design study of the demonstration fast breeder reactor. 2. Shielding design on the basis of the JASPER analysis

    International Nuclear Information System (INIS)

    Suzuoki, Zenro; Tabayashi, Masao; Handa, Hiroyuki; Iida, Masaaki; Takemura, Morio

    2000-01-01

    Conceptual shielding design has been performed for the Demonstration Fast Breeder Reactor (DFBR) to achieve further optimization and reduction of the plant construction cost. The design took into account its implications in overall plant configuration such as reduction of shields in the core, adoption of fission gas plenum in the lower portion of fuel assemblies, and adoption of gas expansion modules. Shielding criteria applied for the design are to secure fast neutron fluence on in-vessel structures as well as responses of the nuclear instrumentation system and to restrict secondary sodium activation. The design utilized the cross sections and the one- and two-dimensional discrete ordinates transport codes, whose verification had been performed by the JASPER experiment analysis. Correction factors yielded by the JASPER analysis were applied to the design calculations to obtain design values with improved accuracy. Design margins, which are defined by the ratios of the design criteria to the design values, were more than two for all shielding issues of interest, showing the adequacy of the shielding design of the DFBR. (author)

  11. RF DEMO ceramic helium cooled blanket, coolant and energy transformation systems

    International Nuclear Information System (INIS)

    Kovalenko, V.; Leshukov, A.; Poliksha, V.; Popov, A.; Strebkov, Yu.; Borisov, A.; Shatalov, G.; Demidov, V.; Kapyshev, V.

    2004-01-01

    RF DEMO-S reactor is a prototype of commercial fusion reactors for further generation. A blanket is the main element unit of the reactor design. The segment structure is the basis of the ceramic blanket. The segments mounting/dismounting operations are carried out through the vacuum vessel vertical port. The inboard/outboard blanket segment is the modules welded design, which are welded by back plate. The module contains the back plate, the first wall, lateral walls and breeding zone. The 9CrMoVNb steel is used as structural material. The module internal space formed by the first wall, lateral walls and back plate is used for breeding zone arrangement. The breeding zone design based upon the poloidal BIT (Breeder Inside Tube) concept. The beryllium is used as multiplier material and the lithium orthosilicate is used as breeder material. The helium at 0.1 MPa is used as purge gas. The cooling is provided by helium at 10 MPa. The coolant supply/return to the blanket modules are carrying out on the two independent circuits. The performed investigations of possible transformation schemes of DEMO-S blanket heat power into the electricity allowed to make a conclusion about the preferable using of traditional steam-turbine facility in the secondary circuit. (author)

  12. European Helium Cooled Pebble Bed (HCPB) test blanket. ITER design description document. Status 1.12.1996

    International Nuclear Information System (INIS)

    Albrecht, H.; Boccaccini, L.V.; Dalle Donne, M.; Fischer, U.; Gordeev, S.; Hutter, E.; Kleefeldt, K.; Norajitra, P.; Reimann, G.; Ruatto, P.; Schleisiek, K.; Schnauder, H.

    1997-04-01

    The Helium Cooled Pebble Bed (HCPB) blanket is based on the use of separate small lithium orthosilicate and beryllium pebble beds placed between radial toroidal cooling plates. The cooling is provided by helium at 8 MPa. The tritium produced in the pebble beds is purged by the flow of helium at 0.1 MPa. The structural material is martensitic steel. It is foreseen, after an extended R and D work, to test in ITER a blanket module based on the HCPB design, which is one of the two European proposals for the ITER Test Blanket Programme. To facilitate the handling operation the Blanket Test Module (BTM) is bolted to a surrounding water cooled frame fixed to the ITER shield blanket back plate. For the design of the test module, three-dimensional Monte Carlo neutronic calculations and thermohydraulic and stress analyses for the operation during the Basic Performance Phase (BPP) and during the Extended Performance Phase (EPP) of ITER have been performed. The behaviour of the test module during LOCA and LOFA has been investigated. Conceptual designs of the required ancillary loops have been performed. The present report is the updated version of the Design Description Document (DDD) for the HCPB Test Module. It has been written in accordance with a scheme given by the ITER Joint Central Team (JCT) and accounts for the comments made by the JCT to the previous version of this report. This work has been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhne and it is supported by the European Union within the European Fusion Technology Program. (orig.) [de

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

    International Nuclear Information System (INIS)

    Daigo Tsuru; Mikio Enoeda; Masato Akiba

    2006-01-01

    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

  14. Improved thermal/MHD design of self-cooled blankets for high-power-density fusion reactors

    International Nuclear Information System (INIS)

    Sedehi, S.; Lund, K.O.

    1986-01-01

    In this work, an improved self-cooled blanket design is conceived that seeks to minimize the induced current and pressure loss, while maintaining effective cooling and power output. Standard solutions for fully developed MHD flows in rectangular ducts are utilized to describe the magnetic pressure drop in rectangular ducts in terms of the duct aspects ratio. A newly available analytical result for developing and fully developed temperatures is utilized in determining the maximum wall temperature and outlet temperature. Based on results from rectangular ducts, improved annular-type duct designs are proposed and evaluated. The results from the rectangular duct analysis indicate reduced pressure drop and increased thermal performance for large aspect ratio (ratio of duct width in the toroidal B-field direction to width normal to B-field). An infinite aspect ratio occurs for the annular duct design and it is shown that this configuration has superior characteristics as a self-cooled blanket design concept

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

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-01-01

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

  16. Fusion fuel blanket technology

    International Nuclear Information System (INIS)

    Hastings, I.J.; Gierszewski, P.

    1987-05-01

    The fusion blanket surrounds the burning hydrogen core of a fusion reactor. It is in this blanket that most of the energy released by the nuclear fusion of deuterium-tritium is converted into useful product, and where tritium fuel is produced to enable further operation of the reactor. As fusion research turns from present short-pulse physics experiments to long-burn engineering tests in the 1990's, energy removal and tritium production capabilities become important. This technology will involve new materials, conditions and processes with applications both to fusion and beyond. In this paper, we introduce features of proposed blanket designs and update and status of international research. In focusing on the Canadian blanket technology program, we discuss the aqueous lithium salt blanket concept, and the in-reactor tritium recovery test program

  17. Nuclear, thermo-mechanical and tritium release analysis of ITER breeding blanket

    International Nuclear Information System (INIS)

    Kosaku, Yasuo; Kuroda, Toshimasa; Enoeda, Mikio; Hatano, Toshihisa; Sato, Satoshi; Miki, Nobuharu; Akiba, Masato

    2003-06-01

    The design of the breeding blanket in ITER applies pebble bed breeder in tube (BIT) surrounded by multiplier pebble bed. It is assumed to use the same module support mechanism and coolant manifolds and coolant system as the shielding blankets. This work focuses on the verification of the design of the breeding blanket, from the viewpoints which is especially unique to the pebble bed type breeding blanket, such as, tritium breeding performance, tritium inventory and release behavior and thermo-mechanical performance of the ITER breeding blanket. With respect to the neutronics analysis, the detailed analyses of the distribution of the nuclear heating rate and TBR have been performed in 2D model using MCNP to clarify the input data for the tritium inventory and release rate analyses and thermo-mechanical analyses. With respect to the tritium inventory and release behavior analysis, the parametric analyses for selection of purge gas flow rate were carried out from the view point of pressure drop and the tritium inventory/release performance for Li 2 TiO 3 breeder. The analysis result concluded that purge gas flow rate can be set to conventional flow rate setting (88 l/min per module) to 1/10 of that to save the purge gas flow and minimize the size of purge gas pipe. However, it is necessary to note that more tritium is transformed to HTO (chemical form of water) in case of Li 2 TiO 3 compared to other breeder materials. With respect to the thermo-mechanical analyses of the pebble bed blanket structure, the analyses have been performed by ABAQUS with 2D model derived from one of eight facets of a blanket module, based on the reference design. Analyses were performed to identify the temperature distribution incorporating the pebble bed mechanical simulation and influence of mechanical behavior to the thermal behavior. The result showed that the maximum temperature in the breeding material was 617degC in the first row of breeding rods and the minimum temperature was 328

  18. Size limitations for microwave cavity to simulate heating of blanket material in fusion reactor

    International Nuclear Information System (INIS)

    Wolf, D.

    1987-01-01

    The power profile in the blanket material of a nuclear fusion reactor can be simulated by using microwaves at 200 MHz. Using these microwaves, ceramic breeder materials can be thermally tested to determine their acceptability as blanket materials without entering a nuclear fusion environment. A resonating cavity design is employed which can achieve uniform cross sectional heating in the plane transverse to the neutron flux. As the sample size increases in height and width, higher order modes, above the dominant mode, are propagated and destroy the approximation to the heating produced in a fusion reactor. The limits at which these modes develop are determined in the paper

  19. Cost study of the ESPRESSO blanket for a Tandem Mirror Reactor

    International Nuclear Information System (INIS)

    Raffray, A.R.; Hoffman, M.A.; Gaskins, T.

    1986-02-01

    A detailed cost study of the ESPRESSO blanket concept for the Tandem Mirror Fusion Reactor (TMR) has been performed to complement the thermal-hydraulic parametric study and to help narrow down the choice of parameters for the final design. The ESPRESSO blanket consists of a number of structurally independent ring modules. Each ring module is made up of a number of mutually pressure-supporting canisters containing arrays of breeder tubes. Two separate helium coolant flows are used: a main flow to cool the tube bank and a cooler first wall flow

  20. Computer aided design of operational units for tritium recovery from Li17Pb83 blanket of a DEMO fusion reactor

    International Nuclear Information System (INIS)

    Malara, C.; Viola, A.

    1995-01-01

    The problem of tritium recovery from Li 17 Pb 83 blanket of a DEMO fusion reactor is analyzed with the objective of limiting tritium permeation into the cooling water to acceptable levels. To this aim, a mathematical model describing the tritium behavior in blanket/recovery unit circuit has been formulated. By solving the model equations, tritium permeation rate into the cooling water and tritium inventory in the blanket are evaluated as a function of dimensionless parameters describing the combined effects of overall resistance for tritium transfer from Li 17 Pb 83 alloy to cooling water, circulating rate of the molten alloy in blanket/recovery unit circuit and extraction efficiency of tritium recovery unit. The extraction efficiency is, in turn, evaluated as a function of the operating conditions of recovery unit. The design of tritium recovery unit is then optimized on the basis of the above parametric analysis and the results are herein reported and discussed for a tritium permeation limit of 10 g/day into the cooling water. 14 refs., 9 figs., 2 tabs

  1. The blanket interface to TSTA

    International Nuclear Information System (INIS)

    Clemmer, R.G.; Finn, P.A.; Grimm, T.L.; Sze, D.K.; Anderson, J.L.; Bartlit, J.R.; Naruse, Y.; Yoshida, H.

    1988-01-01

    The requirements of tritium technology are centered in three main areas, (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The Tritium Systems Test Assembly (TSTA) now in operation at Los Alamos National Laboratory (LANL) is dedicated to developing and demonstrating the tritium technology for fuel processing and containment. TSTA is the only fusion fuel processing facility that can operate in a continuous closed-loop mode. The tritium throughput of TSTA is 1000 g/d. However, TSTA does not have a blanket interface system. The authors have initiated a study to define a Breeder Blanket Interface (BBIO) for TSTA. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. Various methods of tritium recovery from liquid lithium were assessed: yttrium gettering, permeation windows, and molten salt extraction. The authors' evaluation concluded that the best method was molten salt extraction

  2. Target/Blanket Design for the Accelerator Production of Tritium Plant

    International Nuclear Information System (INIS)

    Cappiello, M. W.

    1997-01-01

    The Accelerator Production of Tritium Target/Blanket (T/B) system is comprised of the T/B assembly and the attendant heat removal systems. The T/B assembly produces tritium using a high energy proton beam, and a spallation neutron source. The supporting heat removal systems safely remove the heat deposited by the proton beam during both normal and off-normal conditions. All systems reside within the T/B building, which is located at the end of a linear accelerator. Protons are accelerated to an energy of 1700 MeV at a current of 100 mA and are directed onto the T/B assembly. The protons interact with tungsten and lead nuclei to produce neutrons through the process of nuclear spallation. Neutron capture in 3 He gas produces tritium which is removed on a continual basis in an adjacent Tritium Separation Facility (TSF). The T/B assembly is modular to allow for replacement of spent components and minimization of waste. Systems and components are designed with safety as a primary consideration to minimize risk to the workers and the public

  3. Physical Model Development and Benchmarking for MHD Flows in Blanket Design

    Energy Technology Data Exchange (ETDEWEB)

    Ramakanth Munipalli; P.-Y.Huang; C.Chandler; C.Rowell; M.-J.Ni; N.Morley; S.Smolentsev; M.Abdou

    2008-06-05

    An advanced simulation environment to model incompressible MHD flows relevant to blanket conditions in fusion reactors has been developed at HyPerComp in research collaboration with TEXCEL. The goals of this phase-II project are two-fold: The first is the incorporation of crucial physical phenomena such as induced magnetic field modeling, and extending the capabilities beyond fluid flow prediction to model heat transfer with natural convection and mass transfer including tritium transport and permeation. The second is the design of a sequence of benchmark tests to establish code competence for several classes of physical phenomena in isolation as well as in select (termed here as “canonical”,) combinations. No previous attempts to develop such a comprehensive MHD modeling capability exist in the literature, and this study represents essentially uncharted territory. During the course of this Phase-II project, a significant breakthrough was achieved in modeling liquid metal flows at high Hartmann numbers. We developed a unique mathematical technique to accurately compute the fluid flow in complex geometries at extremely high Hartmann numbers (10,000 and greater), thus extending the state of the art of liquid metal MHD modeling relevant to fusion reactors at the present time. These developments have been published in noted international journals. A sequence of theoretical and experimental results was used to verify and validate the results obtained. The code was applied to a complete DCLL module simulation study with promising results.

  4. Physical Model Development and Benchmarking for MHD Flows in Blanket Design

    International Nuclear Information System (INIS)

    Munipalli, Ramakanth; Huang, P.-Y.; Chandler, C.; Rowell, C.; Ni, M.-J.; Morley, N.; Smolentsev, S.; Abdou, M.

    2008-01-01

    An advanced simulation environment to model incompressible MHD flows relevant to blanket conditions in fusion reactors has been developed at HyPerComp in research collaboration with TEXCEL. The goals of this phase-II project are two-fold: The first is the incorporation of crucial physical phenomena such as induced magnetic field modeling, and extending the capabilities beyond fluid flow prediction to model heat transfer with natural convection and mass transfer including tritium transport and permeation. The second is the design of a sequence of benchmark tests to establish code competence for several classes of physical phenomena in isolation as well as in select (termed here as 'canonical',) combinations. No previous attempts to develop such a comprehensive MHD modeling capability exist in the literature, and this study represents essentially uncharted territory. During the course of this Phase-II project, a significant breakthrough was achieved in modeling liquid metal flows at high Hartmann numbers. We developed a unique mathematical technique to accurately compute the fluid flow in complex geometries at extremely high Hartmann numbers (10,000 and greater), thus extending the state of the art of liquid metal MHD modeling relevant to fusion reactors at the present time. These developments have been published in noted international journals. A sequence of theoretical and experimental results was used to verify and validate the results obtained. The code was applied to a complete DCLL module simulation study with promising results.

  5. Liquid metal magnetohydrodynamic flows in manifolds of dual coolant lead lithium blankets

    Energy Technology Data Exchange (ETDEWEB)

    Mistrangelo, C., E-mail: chiara.mistrangelo@kit.edu; Bühler, L.

    2014-10-15

    Highlights: • MHD flows in model geometries of DCLL blanket manifolds. • Study of velocity, pressure distributions and flow partitioning in parallel ducts. • Flow partitioning affected by 3D MHD pressure drop and velocity distribution in the expanding zone. • Reduced pressure drop in a continuous expansion compared to a sudden expansion. - Abstract: An attractive blanket concept for a fusion reactor is the dual coolant lead lithium (DCLL) blanket where reduced activation steel is used as structural material and a lead lithium alloy serves both to produce tritium and to remove the heat in the breeder zone. Helium is employed to cool the first wall and the blanket structure. Some critical issues for the feasibility of this blanket concept are related to complex induced electric currents and 3D magnetohydrodynamic (MHD) phenomena that occur in distributing and collecting liquid metal manifolds. They can result in large pressure drop and undesirable flow imbalance in parallel poloidal ducts forming blanket modules. In the present paper liquid metal MHD flows are studied for different design options of a DCLL blanket manifold with the aim of identifying possible sources of flow imbalance and to predict velocity and pressure distributions.

  6. Neutronic calculations for the conceptual design of an in-reactor solid breeder experiment, TRIO-01

    International Nuclear Information System (INIS)

    Childs, R.L.; Gabriel, T.A.; Lillie, R.A.

    1981-03-01

    Neutronics calculations have been performed to obtain tritium production and heat generation rates for the irradiation of solid tritium breeding materials in the Oak Ridge Research Reactor (ORR). Two breeder materials, Li 2 O and LiAlO 2 , were considered. Burnup calculations were performed to estimate the amount of 6 Li present as a function of time

  7. Design of a high-temperature first wall/blanket for a d-d compact Reversed-Field-Pinch reactor (CRFPR)

    International Nuclear Information System (INIS)

    Dabiri, A.E.; Glancy, J.E.

    1983-05-01

    A high-temperature first wall/blanket which would take full advantage of the absence of tritium breeding in a d-d reactor was designed. This design which produces steam at p = 7 MPa and T = 538 0 C at the blanket exit eliminates the requirement for a separate steam generator. A steam cycle with steam-to-steam reheat yielding about 37.5 percent efficiency is compatible with this design

  8. The thermo-mechanical design of the water cooled PB-17Li test blanket module for ITER

    International Nuclear Information System (INIS)

    Nardi, C.; Palmieri, A.; Pinna, T.; Porfini, M.T.; Rapisarda, M.; Roccella, M.; Futterer, M.; Lucca, F.

    1998-01-01

    The Water Cooled Lithium Lead (WCLL) blanket is one of the two European concepts to be further developed. A Test Blanket Module (TBM) representative of the DEMO blanket shall be tested in ITER. This paper reports on the activities related to the thermo-mechanical design analysis, taking into account the electromagnetic and neutronic loads in normal and off normal conditions. These loads were applied to a finite elements model of the structure, and the structural response was compared to the allowable value, dependent on the operating conditions. Besides the loads assumed by the design specifications (pressure, temperature, etc), electro-mechanical and thermal loads have been evaluated. A model of the TBM has been performed to compute the loads related to the electromagnetic effects of a centered plasma disruption. The thermal loads have been evaluated considering the heat deposition from the plasma and from the neutrons. The neutronic analysis has been carried out also in order to evaluate the shielding characteristics of the TBM. Taking into account the thermal and mechanical loads a fracture mechanics analysis has been carried out. From this analysis the J Ic parameter was evaluated at the crack tip and compared with the allowable value. The work carried out showed that the TBM present design fulfills ITER normal operation requirements. (authors)

  9. Study on tritium recovery from breeder materials

    International Nuclear Information System (INIS)

    Moriyama, H.; Moritani, K.

    1997-01-01

    For the development of fusion reactor blanket systems, some of the key issues on the tritium recovery performance of solid and liquid breeder materials were studied. In the case of solid breeder materials, a special attention was focussed on the effects of irradiation on the tritium recovery performance, and tritium release experiments, luminescence measurements of irradiation defects and modeling studies were systematically performed. For liquid breeder materials, tritium recovery experiments from molten salt and liquid lithium were performed, and the technical feasibility of tritium recovery methods was discussed. (author)

  10. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

  11. Thermosyphoning analysis with the CATHENA model of the blanket and first wall cooling loop for the SEAFP reactor design

    International Nuclear Information System (INIS)

    Ross, W.E.

    1994-02-01

    This report documents the thermosyphoning analysis which was performed with the CATHENA network model of one of the blanket and first wall cooling loops of the SEAFP reactor design. This thermosyphoning analysis includes four simulations, each with a slightly different model feature or assumption. These simulations are performed to assess the primary heat transport system behaviour for a complete loss of electrical power event (total loss of flow) and to estimate the rate and extent of heat-up of the incore components. For each event, a description of some of the important aspects of the transient thermalhydraulic behaviour including coolant temperatures, circuit and sector flows, circuit pressure, pressurizer level and outflow, and first wall and blanket temperatures is provided. (author). 4 refs., 2 tabs., 32 figs

  12. Overview of the Last Progresses for the European Test Blanket Modules Projects

    International Nuclear Information System (INIS)

    Salavy, J.-F.; Rampal, G.; Boccaccini, L.V.; Meyder, R.; Neuberger, H.; Laesser, R.; Poitevin, Y.; Zmitko, M.; Rigal, E.

    2006-01-01

    The long-term objective of the EU Breeding Blankets programme is the development of DEMO breeding blankets, which shall assure tritium self-sufficiency, an economically attractive use of the heat produced inside the blankets for electricity generation and a sufficiently high shielding of the superconducting magnets for long time operation. In the short-term so-called DEMO relevant Test Blanket Modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. The Helium Cooled Lithium-Lead (HCLL) breeder blanket and the Helium Cooled Pebble Bed (HCPB) concepts are the two breeder blanket lines presently developed by the EU. The main objective of the EU test strategy related to TBMs in ITER is to provide the necessary information for the design and fabrication of breeding blankets for a future DEMO reactor. EU TBMs shall therefore use the same structural and functional materials, apply similar fabrication technologies, and test adequate processes and components. This paper gives an overview of the last progresses in terms of system design, calculations, test program, safety and R-and-D done these last two years in order to cope with the ambitious objective to introduce two EU TBM systems for day-1 of ITER operation. The engineering design of the two systems is mostly concluded and the priority is now on the development and qualification of the fabrication technologies. From calculations point of view, the last modelling efforts related to the thermal-hydraulic of the first wall, the tritium behaviour, and the box thermal and mechanical resistance in accidental conditions are presented. Last features of the TBM and cooling system designs and integration in ITER reactor are highlighted. In particular, this paper also describes the safety and licensing analyses performed for each concept to be able to include the TBM systems in the ITER preliminary safety report

  13. ITER-FEAT vacuum vessel and blanket design features and implications for the R and D programme

    International Nuclear Information System (INIS)

    Ioki, K.; Cardella, A.; Elio, F.; Onozuka, M.; Daenner, W.; Koizumi, K.; Krylov, V.

    2001-01-01

    A tight fitting configuration of the VV to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the TF ripple. The blanket modules are supported directly by the VV. A full-scale VV sector model has provided critical information related to fabrication technology, and the magnitude of welding distortions and achievable tolerances. This R and D validated the fundamental feasibility of the double-wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and the robustness of solid HIP joining was demonstrated in R and D, by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal. (author)

  14. ITER-FEAT vacuum vessel and blanket design features and implications for the R and D programme

    International Nuclear Information System (INIS)

    Ioki, K.; Cardella, A.; Elio, F.; Onozuka, M.; Daenner, W.; Koizumi, K.; Krylov, V.A.

    2001-01-01

    A configuration in which the vacuum vessel (VV) fits tightly to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the toroidal field ripple. The blanket modules are supported directly by the VV. A full scale VV sector model has provided critical information related to fabrication technology and for testing the magnitude of welding distortions and achievable tolerances. This R and D validated the fundamental feasibility of the double wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and robustness of solid hot isostatic pressing joining were demonstrated in the R and D by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal. (author)

  15. ITER-FEAT vacuum vessel and blanket design features and implications for the R&D programme

    Science.gov (United States)

    Ioki, K.; Dänner, W.; Koizumi, K.; Krylov, V. A.; Cardella, A.; Elio, F.; Onozuka, M.; ITER Joint Central Team; ITER Home Teams

    2001-03-01

    A configuration in which the vacuum vessel (VV) fits tightly to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the toroidal field ripple. The blanket modules are supported directly by the VV. A full scale VV sector model has provided critical information related to fabrication technology and for testing the magnitude of welding distortions and achievable tolerances. This R&D validated the fundamental feasibility of the double wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and robustness of solid hot isostatic pressing joining were demonstrated in the R&D by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal.

  16. Summary of several hydraulic tests in support of the light water breeder reactor design (LWBR development program)

    International Nuclear Information System (INIS)

    McWilliams, K.D.; Turner, J.R.

    1979-05-01

    As part of the Light Water Breeder Reactor development program, hydraulic tests of reactor components were performed. This report presents the results of several of those tests performed for components which are somewhat unique in their application to a pressurized water reactor design. The components tested include: triplate orifices used for flow distribution purposes, multiventuri type flowmeters, tight lattice triangular pitch rod support grids, fuel rod end support plates, and the balance piston which is a major component of the movable fuel balancing system. Test results include component pressure loss coefficients, flowmeter coefficients and fuel rod region pressure drop characteristics

  17. Technical issues for beryllium use in fusion blanket applications

    International Nuclear Information System (INIS)

    McCarville, T.J.; Berwald, D.H.; Wolfer, W.; Fulton, F.J.; Lee, J.D.; Maninger, R.C.; Moir, R.W.; Beeston, J.M.; Miller, L.G.

    1985-01-01

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

  18. KNK II, third core: Design Report of the breeder assemblies NR-230R, NR-231R, NR-233D and NR-234D

    International Nuclear Information System (INIS)

    Ricken, E.

    1988-07-01

    The report describes the blanket assemblies of the third core of KNK II and their operational behavior. The assemblies and their individual components are described and the design criteria, design methods and design results are presented. With the help of enveloping proofs the respectation of the design targets up to the foreseen residence time of 720 equivalent full-power days is evidenced [de

  19. An assessment of the base blanket for ITER

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  20. Symbiosis of near breeder HTR's with hybrid fusion reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1978-07-01

    In this contribution to INFCE a symbiotic fusion/fission reactor system, consisting of a hybrid beam-driven micro-explosion fusion reactor (HMER) and associated high-temperature gas-cooled reactors (HTR) with a coupled fuel cycle, is proposed. This system is similar to the well known Fast Breeder/Near Breeder HTR symbiosis except that the fast fission breeder - running on the U/Pu-cycle in the core and the axial blankets and breeding the surplus fissile material as U-233 in its radial thorium metal or thorium oxide blankets - is replaced by a hybrid micro-explosion DT fusion reactor

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

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    Malara, C.; Casini, G.; Viola, A.

    1995-01-01

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

  6. Design of a boiling water reactor core based on an integrated blanket-seed thorium-uranium concept

    International Nuclear Information System (INIS)

    Nunez-Carrera, Alejandro; Francois, Juan Luis; Martin-del-Campo, Cecilia; Espinosa-Paredes, Gilberto

    2005-01-01

    This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket-seed (BS) assembly. The integrated BS concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned out in a once-through cycle. The idea behind the lattice design is to use the thorium conversion capability in a BWR spectrum, taking advantage of the 233 U build-up. A core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235 U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the fuel assembly

  7. Tritium recovery from fusion blankets using solid lithium compounds. I. Design and minimization of tritium inventory

    International Nuclear Information System (INIS)

    Powell, J.R.

    1975-01-01

    Tritium blanket inventories of 100 curies/MW(e) are readily achievable, and inventories as low as 10 curies/MW(e) are possible for blankets with small lithium compound particulates (less than or equal to 50μ) at T greater than or equal to 800 0 C. Of the three release modes [A - to the main coolant (e.g., He) stream; B - to a separate processing circuit; and C - to the plasma region], mode A appears optimum for blankets using gas-cooled metallic structures (e.g., Al, stainless), while mode C appears optimum for high temperature refractory (e.g., C, SiC) structures. The greater structural complexity of mode B makes it less attractive than modes A and C. No recovery method is required for mode C release. With mode A release, tritium inventory in the coolant circuits ranges from 1 to 10 curies/MW(e), depending on processing parameters. Tritium leak rates to the environment during normal operation can be kept to less than or equal to 10 -3 curies/MW(e) per day with low permeability barriers. In general, a mixture of T 2 and T 2 O is present in the coolant stream. Three methods of tritium recovery are examined: (1) Conversion to T 2 followed by absorption in a metal hydride bed. (2) Conversion to T 2 followed by condensation at approximately 6 0 K. (3) Conversion to T 2 O followed by condensation at approximately 100 0 K

  8. Present status of the Liquid Breeder Validation Module for IFMIF

    International Nuclear Information System (INIS)

    Casal, Natalia; Mas, Avelino; Mota, Fernando; García, Ángela; Rapisarda, David; Nomen, Oriol; Arroyo, Jose Manuel; Abal, Javier; Mollá, Joaquín; Ibarra, Ángel

    2013-01-01

    Highlights: • The LBVM will be used to perform irradiation experiments on functional materials for fusion reactors. • It houses 16 experimental rigs, each one containing a EUROFER capsule partially filled with lithium lead, at 300–550 °C. • A helium purge gas will sweep the tritium permeated through the capsule walls to a tritium measuring station. • A helium cooling system will keep tritium diffusion within safe margins and guarantee its mechanical integrity. • Thermal hydraulic and mechanical calculations, the module instrumentation and aspects as safety or RAMI are presented. -- Abstract: One of the objectives of IFMIF (International Fusion Materials Irradiation Facility), as stated in its specifications, is the validation of breeder blanket concepts for DEMO design. The so-called Liquid Breeder Validation Module (LBVM) will be used in IFMIF to perform experiments under irradiation on functional materials related to liquid breeder concepts for future fusion reactors. This module, not considered in previous IFMIF design phases, is currently under design by CIEMAT in the framework of the IFMIF/EVEDA project. In this paper, the present status of the design of the LBVM is presented

  9. Present status of the Liquid Breeder Validation Module for IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Casal, Natalia, E-mail: natalia.casal@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Mas, Avelino; Mota, Fernando; García, Ángela; Rapisarda, David [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Nomen, Oriol [Institut de Recerca en Energia de Catalunya (IREC), Barcelona (Spain); Centre de Disseny d’Equips Industrials (CDEI), Technical University of Catalonia (UPC), Barcelona (Spain); Arroyo, Jose Manuel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Abal, Javier [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Mollá, Joaquín; Ibarra, Ángel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain)

    2013-10-15

    Highlights: • The LBVM will be used to perform irradiation experiments on functional materials for fusion reactors. • It houses 16 experimental rigs, each one containing a EUROFER capsule partially filled with lithium lead, at 300–550 °C. • A helium purge gas will sweep the tritium permeated through the capsule walls to a tritium measuring station. • A helium cooling system will keep tritium diffusion within safe margins and guarantee its mechanical integrity. • Thermal hydraulic and mechanical calculations, the module instrumentation and aspects as safety or RAMI are presented. -- Abstract: One of the objectives of IFMIF (International Fusion Materials Irradiation Facility), as stated in its specifications, is the validation of breeder blanket concepts for DEMO design. The so-called Liquid Breeder Validation Module (LBVM) will be used in IFMIF to perform experiments under irradiation on functional materials related to liquid breeder concepts for future fusion reactors. This module, not considered in previous IFMIF design phases, is currently under design by CIEMAT in the framework of the IFMIF/EVEDA project. In this paper, the present status of the design of the LBVM is presented.

  10. The ITER EC H&CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

    NARCIS (Netherlands)

    Gessner, R.; Aiello, G.; Grossetti, G.; Meier, A.; Ronden, D.; Spaeh, P.; Scherer, T.; Schreck, S.; Strauss, D.; Vaccaro, A.

    2013-01-01

    The final design of the structural system for the ITER EC H&CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield

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

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, B.J., E-mail: Brad.Merrill@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Wong, C.P.C. [General Atomics, San Diego, CA 92186-5608 (United States); Cadwallader, L.C. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Abdou, M.; Morley, N.B. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States)

    2014-10-15

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

  12. Neutronic design and analysis on dual-cooled waste transmutation blanket for the fusion driven sub-critical system

    International Nuclear Information System (INIS)

    Zheng Shanliang; Wu Yican; Gao Chunjing; Xu Dezheng; Li Jingjing; Zhu Xiaoxiang

    2004-01-01

    Neutronics design and analysis of dual-cooled multi-functional waste transmutation blanket (DWTB) for the fusion driven sub-critical system (FDS) are performed to ensure the system be able to meet the requirements of fuel-sufficiency and more waste transmutation ratio with low initial loading fuel inventory, which is based on 1-D burn-up calculations with home-developed code Visual BUS and the multi-group (175 neutron groups-42 Gamma groups coupled) data library HENDL1.0/MG (Hybrid Evaluated Nuclear Data Library). (authors)

  13. Competitive breeder power plants

    International Nuclear Information System (INIS)

    Winkleblack, R.K.

    1984-01-01

    To utilize the fissile material that is accumulating in the utilities' spent fuel pools, breeder plants must be less expensive than current LWR costs (or utilities will not buy nuclear plants in the near future) and also be highly reliable. The fundamental differences between LWRs and LMFBRs are discussed and recommendations are made for making the most of these differences to design a superior breeder plant that can sell in the future, opening the way to U.S. utilities becoming self-sufficient for fuel supply for centuries

  14. Conceptual design of the blanket and power conversion system for a mirror hybrid fusion-fission reactor. 12-month progress report, July 1, 1975--June 30, 1976

    International Nuclear Information System (INIS)

    Schultz, K.R.; Baxi, C.B.; Rao, R.

    1976-01-01

    This report presents the conceptual design and preliminary feasibility assessment for the hybrid blanket and power conversion system of the Mirror Hybrid Fusion-Fission Reactor. Existing gas-cooled fission reactor technology is directly applicable to the Mirror Hybrid Reactor. There are a number of aspects of the present conceptual design that require further design and analysis effort. The blanket and power conversion system operating parameters have not been optimized. The method of supporting the blanket modules and the interface between these modules and the primary loop helium ducting will require further design work. The means of support and containment of the primary loop components must be studied. Nevertheless, in general, the conceptual design appears quite feasible

  15. STRUCTURAL DESIGN CRITERIA FOR TARGET/BLANKET SYSTEM COMPONENT MATERIALS FOR THE ACCELERATOR PRODUCTION OF TRITIUM PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    W. JOHNSON; R. RYDER; P. RITTENHOUSE

    2001-01-01

    The design of target/blanket system components for the Accelerator Production of Tritium (APT) plant is dependent on the development of materials properties data specified by the designer. These data are needed to verify that component designs are adequate. The adequacy of the data will be related to safety, performance, and economic considerations, and to other requirements that may be deemed necessary by customers and regulatory bodies. The data required may already be in existence, as in the open technical literature, or may need to be generated, as is often the case for the design of new systems operating under relatively unique conditions. The designers' starting point for design data needs is generally some form of design criteria used in conjunction with a specified set of loading conditions and associated performance requirements. Most criteria are aimed at verifying the structural adequacy of the component, and often take the form of national or international standards such as the ASME Boiler and Pressure Vessel Code (ASME B and PV Code) or the French Nuclear Structural Requirements (RCC-MR). Whether or not there are specific design data needs associated with the use of these design criteria will largely depend on the uniqueness of the conditions of operation of the component. A component designed in accordance with the ASME B and PV Code, where no unusual environmental conditions exist, will utilize well-documented, statistically-evaluated developed in conjunction with the Code, and will not be likely to have any design data needs. On the other hand, a component to be designed to operate under unique APT conditions, is likely to have significant design data needs. Such a component is also likely to require special design criteria for verification of its structural adequacy, specifically accounting for changes in materials properties which may occur during exposure in the service environment. In such a situation it is common for the design criteria

  16. Molecule-surface interaction processes of relevance to gas blanket type fusion device divertor design

    Energy Technology Data Exchange (ETDEWEB)

    Snowdon, K.J. [Newcastle Univ. (United Kingdom). Dept. of Physics; Tawara, H.

    1997-01-01

    The mechanisms which may lead to the departure of molecular species from surfaces exposed to low energy (0.1-100 eV) particle or photon and electron irradiation are reviewed. Where possible, the charge and electronic state, angular, translational and internal energy distributions of the departing molecules are described and the physical origin of the nature of those distributions identified. The consequences, for the departing molecules, of certain material choices become apparent from such an analysis. Such information may help guide the choice of appropriate materials for plasma facing components of gas-blanket type divertors such as that recently proposed for the International Thermonuclear Experimental Reactor (ITER). (author). 71 refs.

  17. Alternative breeder reactor technologies

    International Nuclear Information System (INIS)

    Spinrad, B.I.

    1978-01-01

    The significance of employing breeder reactors to stretch the world resources of nuclear fuels is briefly discussed, and the various types of breeder concepts are described. General descriptions, advantages, and disadvantages of the liquid metal cooled fast breeder, gas cooled fast breeder, molten salt breeder, thermal breeders, and spectral-shift control reactors are presented. Aspects of safeguarding fissile material connected with breeder operation are examined. 31 references

  18. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    International Nuclear Information System (INIS)

    1980-01-01

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases

  19. Beryllium and lithium resource requirements for solid blanket designs for fusion reactors

    International Nuclear Information System (INIS)

    Powell, J.R.

    1975-01-01

    The lithium and beryllium requirements are analyzed for an economy of 10 6 MW(e) CTR 3 capacity using solid blanket fusion reactors. The total lithium inventory in fusion reactors is only approximately 0.2 percent of projected U. S. resources. The lithium inventory in the fusion reactors is almost entirely 6 Li, which must be extracted from natural lithium. Approximately 5 percent of natural lithium can be extracted as 6 Li. Thus the total feed of natural lithium required is approximately 20 times that actually used in fusion reactors, or approximately 4 percent of U. S. resources. Almost all of this feed is returned to the U. S. resource base after 6 Li is extracted, however. The beryllium requirements are on the order of 10 percent of projected U. S. resources. Further, the present cost of lithium and the cost of beryllium extraction could both be increased tenfold with only minor effects on CTR capital cost. Such an increase should substantially multiply the economically recoverable resources of lithium and beryllium. It is concluded that there are no lithium or beryllium resource limitations preventing large-scale implementation of solid blanket fusion reactors. (U.S.)

  20. Beryllium research on FFHR molten salt blanket

    International Nuclear Information System (INIS)

    Terai, T.; Tanaka, S.; Sze, D.-K.

    2000-01-01

    Force-free helical reactor, FFHR, is a demo-relevant heliotron-type D-T fusion reactor based on the great amount of R and D results obtained in the LHD project. Since 1993, collaboration works have made great progress in design studies of FFHR with standing on the major advantage of current-less steady operation with no dangerous plasma disruptions. There are two types of reference designs, FFHR-1 and FFHR-2, where molten Flibe (LiF-BeF2) is utilized as tritium breeder and coolant. In this paper, we present the outline of FFHR blanket design and some related R and D topics focusing on Be utilization. Beryllium is used as a neutron multiplier in the design and Be pebbles are placed in the front part of the tritium breeding zone. In a Flibe blanket, HF (TF) generated due to nuclear transmutation will be a problem because of its corrosive property. Though nickel-based alloys are thought to be intact in such a corrosive environment, FFHR blanket design does not adopt the alloys because of their induced radioactivity. The present candidate materials for the structure are low-activated ferritic steel (JLF-1), V-4Cr-4Ti, etc. They are capable to be corroded by HF in the operation condition, and Be is expected to work as a reducing agent in the system as well. Whether Be pebbles placed in a Flibe flow can work well or not is a very important matter. From this point, Be solubility in Flibe, reaction rate of the Redox reaction with TF in the liquid and on the surface of Be pebbles under irradiation, flowing behavior of Flibe through a Be pebble bed, etc. should be investigated. In 1997, in order to establish more practical and new data bases for advanced design works, we started a collaboration work of R and D on blanket engineering, where the Be research above mentioned is included. Preliminary dipping-test of Be sheets and in-situ tritium release experiment from Flibe with Be sheets have got started. (orig.)

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

    International Nuclear Information System (INIS)

    Hoshino, Tsuyoshi; Oikawa, Fumiaki; Nishitani, Takeo

    2010-01-01

    DEMO reactors require ' 6 Li-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 Li 2 TiO 3 pebbles and of Li 4 SiO 4 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 6 Li. The development of the production and 6 Li 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 HNO 3 and H 2 O 2 . The removal efficiency of 60 Co 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

  2. Thermo-mechanical characterization of ceramic pebbles for breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Lo Frano, Rosa, E-mail: rosa.lofrano@ing.unipi.it; Aquaro, Donato; Scaletti, Luca

    2016-11-01

    Highlights: • Experimental activities to characterize the Li{sub 4}SiO{sub 4}. • Compression tests of pebbles. • Experimental evaluation of thermal conductivity of pebbles bed at different temperatures. • Experimental test with/without compression load. - Abstract: An open issue for fusion power reactor is to design a suitable breeding blanket capable to produce the necessary quantity of the tritium and to transfer the energy of the nuclear fusion reaction to the coolant. The envisaged solution called Helium-Cooled Pebble Bed (HCPB) breeding blanket foresees the use of lithium orthosilicate (Li{sub 4}SiO{sub 4}) or lithium metatitanate (Li{sub 2}TiO{sub 3}) pebble beds. The thermal mechanical properties of the candidate pebble bed materials are presently extensively investigated because they are critical for the feasibility and performances of the numerous conceptual designs which use a solid breeder. This study is aimed at the investigation of mechanical properties of the lithium orthosilicate and at the characterization of the main chemical, physical and thermo-mechanical properties taking into account the production technology. In doing that at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa adequate experiments were carried out. The obtained results may contribute to characterize the material of the pebbles and to optimize the design of the envisaged fusion breeding blankets.

  3. An option for the Brazilian nuclear project: necessity of fast breeder reactors and core design for an experimental fast reactor

    International Nuclear Information System (INIS)

    Ishiguro, Y.

    1983-01-01

    In order to assure the continued utilization of fission energy, development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed as the best type for future Brazilian nuclear systems. The inherent safety characteristics are superior to current FBRs and an efficient utilization of the abundant thorium is possible. A first step and a basic tool for the development of FBR technologies is the construction and operation of an experimental fast reactor (EFR). A series of core designs for a 90 MW EFR is studied and possible options and the magnitudes of principal parameters are identified. Flexible modifications of the core and sufficiently high fast fluxes for fuel and materials irradiations appear possible. (Author) [pt

  4. Feasible region of design parameters for water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2007-01-01

    The performances of a light water cooled thorium breeder reactor have been investigated. A feasible region of fresh fuel enrichment and moderator to fuel ratio (MFR) is found to satisfy the constrains of criticality, breeding, and negative void coefficient for several burnups of discharged fuel. The equilibrium fuel cycle burnup calculation has been performed which is coupled with the cell calculation. The MFR is changed to investigate its effect to the breeding capability and void reactivity coefficient profile for different average discharged burnups. For moderated cases, the conversion ratio (CR) decreases with increasing burnup and MFR. The ratio of fissile inventory in equilibrium core to the initial fissile loading (FIR) has the maximum value at certain burnups depending on the MFR and its value increases with the decreasing MFR. Considering to the breeding capability of the reactor, for burnups of equal to 30 GWd/t or higher, the MFR≤0.3 is needed. For the larger MFR and lower burnups, the void reactivity coefficient becomes more negative with an increasing void fraction. The most negative value of the void reactivity coefficient is obtained at MFR=0.3. (author)

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

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Seki, Masahiro; Minato, Akio

    1987-03-01

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

  6. Improved structure and long-life blanket concepts for heliotron reactors

    International Nuclear Information System (INIS)

    Sagara, A.; Imagawa, S.; Mitarai, O.

    2005-01-01

    New design approaches are proposed for the LHD-type heliotron D-T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius of over 14m is selected to permit a blanket-shield thickness of about 1m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armour tiles that soften the neutron energy spectrum incident on the self-cooled flibe-reduced activation ferritic steel blanket. In this adaptation of the spectral-shifter and tritium breeder blanket (STB) concept a local tritium breeding ratio over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the superconducting magnet coils is also significantly improved. Using constant cross sections of a helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armour tiles. The key R and D issues for developing the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated. (author)

  7. Improved structure and long-life blanket concepts for heliotron reactors

    Science.gov (United States)

    Sagara, A.; Imagawa, S.; Mitarai, O.; Dolan, T.; Tanaka, T.; Kubota, Y.; Yamazaki, K.; Watanabe, K. Y.; Mizuguchi, N.; Muroga, T.; Noda, N.; Kaneko, O.; Yamada, H.; Ohyabu, N.; Uda, T.; Komori, A.; Sudo, S.; Motojima, O.

    2005-04-01

    New design approaches are proposed for the LHD-type heliotron D-T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius of over 14 m is selected to permit a blanket-shield thickness of about 1 m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armour tiles that soften the neutron energy spectrum incident on the self-cooled flibe-reduced activation ferritic steel blanket. In this adaptation of the spectral-shifter and tritium breeder blanket (STB) concept a local tritium breeding ratio over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the superconducting magnet coils is also significantly improved. Using constant cross sections of a helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armour tiles. The key R&D issues for developing the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated.

  8. Activation analysis of tritium breeder lithium lead irradiated by fusion neutrons in FDS-II

    International Nuclear Information System (INIS)

    Mingliang Chen

    2006-01-01

    R-and-D of fusion materials, especially their activation characteristics, is one of the key issues for fusion research in the world. Research on activation characteristics for low activation materials, such as reduced activation ferritic/martensitic steels, vanadium alloys and SiCf/SiC composites, is being done throughout the world to ensure the attractiveness of fusion power regarding safety and environmental aspects. However, there is less research on the activation characteristics of the other important fusion materials, such as tritium breeder etc.. Lithium lead (Li 17 Pb 83 ) is presently considered as a primary candidate tritium breeder for fusion power reactors because of its attractive characteristics. It can serve as a tritium breeder, neutron multiplier and coolant in the blanket at the same time. The radioactivity of Li 17 Pb 83 by D-T fusion neutrons in FDS-II has been calculated and analyzed. FDS-II is a concept design of fusion power reactor, which consists of fusion core with advanced plasma parameters extrapolated from the ITER (International Thermonuclear Experimental Reactor) and two candidates of liquid lithium breeder blankets (named SLL and DLL blankets). The neutron transport and activation calculation are carried out based on the one-dimensional model for FDS-II with the home-developed multi-functional code system VisualBUS and the multi-group data library HENDL1.0/MG and European Activation File EAF-99. The effects of irradiation time on the activation characteristics of Li 17 Pb 83 were analyzed and it concludes that the irradiation time has an important effect on the activation level of Li 17 Pb 83 . Furthermore, the results were compared with the activation levels of other tritium breeders, such as Li 4 SiO 4 , Li 2 TiO 3 , Li 2 O and Li etc., under the same irradiation conditions. The dominant nuclides to dose rate and activity of Li 17 Pb 83 were analyzed as well. Tritium generated by Li has a great contribution to the afterheat and

  9. Collection of Summaries of reports on result of research at basic experiment device for nuclear fusion reactor blanket design, 1994

    International Nuclear Information System (INIS)

    1995-07-01

    The development of nuclear fusion reactors reached such stage that the generation of fusion power output comparable with the input power into core plasma is possible. At present, the engineering design of the international thermonuclear fusion experimental reactor, ITER, is advanced by the cooperation of Japan, USA, Europe and Russia, aiming at the start of operation at the beginning of 21st century. This meeting for reporting the results has been held every year, and this time, it was held on May 19, 1995 at University of Tokyo with the theme ''The interface properties of fusion reactor materials and the control of particle transport''. About 50 participants from academic, governmental and industrial circles discussed actively on the theme. Three lectures on the topics of fusion reactor engineering and materials and seven lectures on the basic experiment of fusion reactor blanket design related to the next period project were given at the meeting. (K.I.)

  10. Status on DEMO Helium Cooled Lithium Lead breeding blanket thermo-mechanical analyses

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, G.; Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Kiss, B. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Morin, A. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France)

    2016-11-01

    Highlights: • CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. The DEMO HCLL breeding blanket design capitalizes on the experience acquired on the HCLL Test Blanket Module designed for ITER. Design improvements are being implemented to adapt the design to DEMO specifications and performance objectives. • Thermal and mechanical analyses have been carried out in order to justify the design of the HCLL breeding blanket showing promising results for tie rods modules’ attachments system and relatively good behavior of the box in case of LOCA when comparing to RCC-MRx criteria. • CFD thermal analyses on generic breeding unit have enabled the consolidation of the results obtained with previous FEM design analyses. - Abstract: The EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the breeding blanket surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which is investigated for DEMO. It is made of a Eurofer structure and uses the eutectic liquid lithium–lead as tritium breeder and neutron multiplier, and helium gas as coolant. Within the EUROfusion organization, CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. This paper presents the status of the thermal and mechanical analyses carried out on the HCLL breeding blanket in order to justify the design. CFD thermal analyses on generic breeding unit including stiffening plates and cooling plates have been performed with ANSYS in order to consolidate results obtained with previous FEM design analyses. Moreover in order to expand the justification of the HCLL Breeding blanket design, the most loaded area of

  11. The design and fabrication of an optical periscope for core viewing of fast breeder test reactor (FBTR)

    International Nuclear Information System (INIS)

    Das, N.C.; Sanjiva Kumar; Udupa, D.V.; Shukla, R.P.; Kadu, A.M.; Modi, R.K.

    2004-08-01

    A FBTR (Fast Breeder Test Reactor) periscope has been designed and fabricated indigenously for viewing and photography/ video recording the objects in the reactor core. The periscope consists of a scanning prism mechanism, zoom lens objective, a system of relay lenses and an eyepiece sub-assembly for viewing the objects. The objective of the periscope is a zoom lens system for obtaining a continuously varying magnification from 2X to 5X. Zoom lens objective system has a variable focal length from 100 mm to 250 mm with an aperture varying from 10 mm to 25 mm respectively. This covers a semi- field angle of 3 deg for the objective lens of focal length of 250 mrn and 4 deg for the objective of focal length of l00 mm. Two prisms of 45 deg -90 deg -45 deg types are used for scanning the object space in vertical direction. One prism is fixed, whereas the prism facing the object can be rotated about the horizontal axis through an angle of 110 deg. The rotation of the entire periscope assembly along the vertical axis scans the object space on the horizontal plane. The combination of these two rotations is used to scan the field of interest. It may be noted here that it is absolutely essential to introduce a Pechan prism before each eyepiece. Pechan prism is used for the rotation of the image, which is produced due to the rotation of the scanning prisms. The measured value of the linear resolution of the instrument is 0.7 mm at an object distance of 2.5 meter from the zoom lens objective system. The periscope has two arm labeled I and II. The arm I is used for visual inspection, while the arm II is used for video recording/photography. The periscope will be used as an in-service instrument for Fast Breeder Test Reactor, IGCAR, Kalpakkam. (author)

  12. Neutronic studies of fissile and fusile breeding blankets

    International Nuclear Information System (INIS)

    Taczanowski, S.

    1984-08-01

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

  13. Fast breeder reactors

    International Nuclear Information System (INIS)

    Waltar, A.E.; Reynolds, A.B.

    1981-01-01

    This book describes the major design features of fast breeder reactors and the methods used for their design and analysis. The foremost objective of this book is to fulfill the need for a textbook on Fast Breeder Reactor (FBR) technology at the graduate level or the advanced undergraduate level. It is assumed that the reader has an introductory understanding of reactor theory, heat transfer, and fluid mechanics. The book is expected to be used most widely for a one-semester general course on fast breeder reactors, with the extent of material covered to vary according to the interest of the instructor. The book could also be used effectively for a two-quarter or a two-semester course. In addition, the book could serve as a text for a course on fast reactor safety since many topics other than those appearing in the safety chapters relate to FBR safety. Methodology in fast reactor design and analysis, together with physical descriptions of systems, is emphasized in this text more than numerical results. Analytical and design results continue to change with the ongoing evolution of FBR design whereas many design methods have remained fundamentally unchanged for a considerable time

  14. Economic analysis of fusion breeders. Supplement

    International Nuclear Information System (INIS)

    Delene, J.G.

    1985-01-01

    Three fusion/fission hybrids and three converter reactors were considered in combination: (1) Li-Be (Opt-Li) blanket, (2) molten salt blanket (1.6 blanket energy multiplier), and (3) molten salt blanket (2.5 blanket energy multiplier). The following converter (fission) reactors were considered: (1) LWR, (2) HTGR, and (3) molten salt. In order to provide some perspective on the results of the hybrid analysis, LMFBRs were also examined: (1) methods applied consistently, and (2) range of LMFBR costs consistent with current thought on advanced designs

  15. High temperature blankets for non-electrical/electrical applications of fusion reactors: Progress report, July 15, 1983--November 30, 1984

    International Nuclear Information System (INIS)

    Ribe, F.L.; Woodruff, G.L.

    1988-01-01

    We report a continuation of work done in collaboration with the Lawrence Livermore National Laboratory (LLNL) on design studies of the tandem-mirror fusion reactor (TMR) coupled to the General Atomic (GA) sulfur-iodine thermochemical process for producing hydrogen. During this report period the emphasis was on a solid-breeder gas cooled ''cannister'' blanket for TMR-based hydrogen production. This work was integrated with the Department of Energy (DOE), Office of Fusion Energy (OFE) Blanket Comparison and Selection Study, coordinated by the Argonne National Laboratory (ANL). The areas investigated by the two principal investigators and their students were the following: Plasma engineering of the TMR, including the magnets. Neutronics transport support for the synfuel blanket and shield. Completion of studies of the GA sulfur-iodine process. Under subcontract D.S. Rowe of Rowe and Associates worked with both UW and LLNL personnel on Mechanical design and thermal hydraulics of a high temperature, solid breeder blanket. 2 refs., 3 figs

  16. BREED: a CDC-7600 computer program for the automation of breeder reactor design analysis (LWBR Development Program)

    International Nuclear Information System (INIS)

    Candelore, N.R.; Maher, C.M.

    1985-03-01

    BREED is an executive CDC-7600 program which was developed to facilitate the sequence of calculations and movement of data through a prescribed series of breeder reactor design computer programs in an uninterrupted single-job mode. It provides the capability to interface different application programs into a single computer run to provide a complete design function. The automation that can be achieved as a result of using BREED significantly reduces not only the time required for data preparation and hand transfer of data, but also the time required to complete an iteration of the total design effort. Data processing within a technical discipline and data transfer between technical disciplines can be accommodated. The input/output data processing is achieved with BREED by using a set of simple, easily understood user commands, usually short descriptive words, which the user inserts in his input deck. The input deck completely identifies and controls the calculational sequence needed to produce a desired end product. This report has been prepared to provide instructional material on the use of BREED and its user-oriented procedures to facilitate computer automation of design calculations

  17. Disruption problematics in segmented blanket concepts

    International Nuclear Information System (INIS)

    Crutzen, Y.; Fantechi, S.; Farfaletti-Casali, F.

    1994-01-01

    In Tokamaks, the hostile operating environment originated by plasma disruption events requires that the first wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence there is a need to improve the safety features of the blanket design concepts satisfying the disruption problematics and to formulate guidelines on the required internal reinforcements of the blanket components. The present paper describes the recent investigations on blanket reinforcement systems needed in order to optimize the first-wall/blanket/shield structural design for next step and commercial fusion reactors in the context of ITER, DEMO and SEAFP activities

  18. Packed-fluidized-bed blanket concept for a thorium-fueled commercial tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Chi, J.W.H.; Miller, J.W.; Karbowski, J.S.; Chapin, D.L.; Kelly, J.L.

    1980-09-01

    A preliminary design of a thorium blanket was carried out as a part of the Commercial Tokamak Hybrid Reactor (CTHR) study. A fixed fuel blanket concept was developed as the reference CTHR blanket with uranium carbide fuel and helium coolant. A fixed fuel blanket was initially evaluated for the thorium blanket study. Subsequently, a new type of hybrid blanket, a packed-fluidized bed (PFB), was conceived. The PFB blanket concept has a number of unique features that may solve some of the problems encountered in the design of tokamak hybrid reactor blankets. This report documents the thorium blanket study and describes the feasibility assessment of the PFB blanket concept

  19. Seismic design technology for Breeder Reactor structures. Volume 3: special topics in reactor structures

    International Nuclear Information System (INIS)

    Reddy, D.P.

    1983-04-01

    This volume is divided into six chapters: analysis techniques, equivalent damping values, probabilistic design factors, design verifications, equivalent response cycles for fatigue analysis, and seismic isolation

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

    Science.gov (United States)

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

    2011-10-15

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

  1. Tritium permeation and recovery for the Flibe/He blanket design

    International Nuclear Information System (INIS)

    Moir, R.

    1984-10-01

    This study assumes tritium to be a gas dissolved in molten salt, with TF formation suppressed. Tritium permeates readily through the hot steel tubes of the reactor and steam generator and will leak into the steam system at the rate of about one gram per day in the absence of special permeation barriers, assuming that 1% of the helium coolant flow rate is processed for tritium recovery at 90% efficiency per pass. Tritiated water in the steam system is a personnel hazard at concentration levels well below one part per million and this level would soon be reached without costly isotopic processing. Alternatively, including a combination of permeation barriers on reactor and steam generator tubes and molten salt processing is estimated to reduce the leak rate into the steam system by over two orders of magnitude. For the option with the lowest estimated leak rate, 55 Ci/d, it may be possible to purge the steam system continuously to prevent tritiated water buildup. At best, isotopic separation of dilute tritiated water may not be necessary and for higher leak-rate options the isotopic processing rate can be reduced. The proposed permeation barrier for the reactor tubes is a 10 μm layer of tungsten which, in principle, will reduce tritium blanket permeation by a factor of about 300 below the bare-steel rate

  2. Neutronic investigation and activation calculation for CFETR HCCB blankets

    Science.gov (United States)

    Shuling, XU; Mingzhun, LEI; Sumei, LIU; Kun, LU; Kun, XU; Kun, PEI

    2017-12-01

    The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder (HCCB) blanket were predicted for the Chinese Fusion Engineering Testing Reactor (CFETR) design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio (TBR). Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil. The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1 × 10-4 kW, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.

  3. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

    Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

    1999-07-01

    The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m 2 ) the average neutron power load on the first wall is below 1 MWm .2 , which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

  4. APT Blanket Detailed Bin Model Based on Initial Plate-Type Design -3D FLOWTRAN-TF Model

    International Nuclear Information System (INIS)

    Hamm, L.L.

    1998-01-01

    This report provides background information for a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report for the APT. This report gives a brief description of the FLOWTRAN-TF code which was used for detailed blanket bin modeling

  5. Neutronics investigation of advanced self-cooled liquid blanket systems in helical reactor

    International Nuclear Information System (INIS)

    Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M.Z.

    2006-10-01

    Neutronics performances of advanced self-cooled liquid blanket systems have been investigated in design activity of the helical-type reactor FFHR2. In the present study, a new three-dimensional (3-D) neutronics calculation system has been developed for the helical-type reactor to enhance quick feedback between neutronics evaluation and design modification. Using this new calculation system, advanced Flibe-cooled and Li-cooled liquid blanket systems proposed for FFHR2 have been evaluated to make clear design issues to enhance neutronics performance. Based on calculated results, modification of the blanket dimensions and configuration have been attempted to achieve the adequate tritium breeding ability and neutron shielding performance in the helical reactor. The total tritium breeding ratios (TBRs) obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. Issues in neutron shielding performance have been investigated quantitatively using 3-D geometry of the helical blanket system, support structures, poloidal coils etc. Shielding performance of the helical coils against direct neutrons from core plasma would achieve design target by further optimization of shielding materials. However, suppression of the neutron streaming and reflection through the divertor pumping areas in the original design is important issue to protect the poloidal coils and helical coils, respectively. Investigation of the neutron wall loading indicated that the peaking factor of the neutron wall load distribution would be moderated by the toroidal and helical effect of the plasma distribution in the helical reactor. (author)

  6. Preconceptual ABC design definition and system configuration layout

    International Nuclear Information System (INIS)

    Barthold, W.

    1995-03-01

    This document is the conceptual design document for the follow-on to the Molten Salt Breeder Reactor, known as the ABC type reactor. It addresses blanket design options, containment options, off-gas systems, drainage systems, and components/layouts of the primary, secondary, and tertiary systems, and it contains a number of diagrams for the configuration of the major systems

  7. Electrical Power System Design and Station Blackout (SBO) Management in Indian Fast Breeder Reactors

    International Nuclear Information System (INIS)

    Vijaya, N. M.; Theivarajan, N.; Madhusoodanan, K.

    2015-01-01

    In the nuclear new builds and projects in design stage SBO management measures have significant role. Depending on the onsite and offsite power supply configurations, deterministic SBO duration is established. Design of systems with adequately sized battery capacities for SBO duration, special SBO Diesel Generator Sets, structured load shedding strategy to conserve battery availability to cope with SBO and to monitor the plant safety beyond SBO duration are considered as part of electrical system design now. In the design of PFBR, SBO is given due importance right from conceptual design stage. Both deterministic SBO duration and probabilistic SBO duration versus frequency were established by detailed analysis. Dedicated DC power supply systems and additional SBO DG back-up systems are in place to cope with normal and extended SBO. After the Fukushima event, there is greater requirement to demonstrate plant safety during SBO for a long duration extended over several days. In light of this accident, thermal hydraulic synthesis of PFBR has been carried out to ascertain the capability of the plant to manage a prolonged station blackout event. This has brought out the robustness of the design. Safety design features of PFBR ensure comfortable management of extended SBO. In the design of future FBR projects, current trends in the new nuclear builds and recommendations of international bodies considering Fukushima are duly considered. SBO measures by means of alternate AC power sources, redundant emergency power supply sources with less dependence on other auxiliary systems and dedicated DC power systems are considered to cope with normal and extended SBO beyond design basis. Right from the conceptual design, the system robustness to manage normal and extended SBO will be taken care with the related thermal hydraulic and associated analysis. The paper highlights these SBO management strategies in PFBR and future FBRs. (author)

  8. Clinch River breeder reactor sodium fire protection system design and development

    International Nuclear Information System (INIS)

    Foster, K.W.; Boasso, C.J.; Kaushal, N.N.

    1984-01-01

    To assure the protection of the public and plant equipment, improbable accidents were hypothesized to form the basis for the design of safety systems. One such accident is the postulated failure of the Intermediate Heat Transfer System (IHTS) piping within the Steam Generator Building (SGB), resulting in a large-scale sodium fire. This paper discusses the design and development of plant features to reduce the consequences of the accident to acceptable levels. Additional design solutions were made to mitigate the sodium spray contribution to the accident scenario. Sodium spill tests demonstrated that large sodium leaks can be safely controlled in a sodium-cooled nuclear power plant

  9. Contribution of Clinch River Breeder Reactor plant design and development to the LMFBR fuel cycle

    International Nuclear Information System (INIS)

    Riley, D.R.; Dickson, P.W.

    1981-01-01

    This paper describes how the CRBRP development and CRBRP focus of the LMFBR base technology program have led to advances in the state of the art in physics, thermal-hydraulics, structural analysis, core restraint, seismic analysis, and analysis of hypothetical core-disruptive accident energetics, all of which have been incorporated through disciplined engineering into the final CRBRP design. The total development in the US of fuels and materials, the analytical advances made on CRBRP design, and the incorporation of the latest experimental results into that design have put the US technology in general and the CRBRP design in particular at the forefront of technology. This has placed the US in a position to develop the most favorable LMFBR fuel cycle

  10. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  11. The preliminary thermal–hydraulic analysis of a water cooled blanket concept design based on RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Guanghuai; Peng, Changhong; Guo, Yun, E-mail: guoyun79@ustc.edu.cn

    2016-11-01

    Highlights: • The superheated steam and PWR schemes are analyzed by RELAP5 code. • The influence of non-uniform heating sources is include. • A supposed slow flow decrease case is discussed and the PWR scheme is better. - Abstract: Water cooled blanket (WCB) is very important in the conceptual design and energy transfer in future fusion power plant. One conceptual design of WCB is under computational testing. RELAP5 code, which is mature and often used in transient analysis in Pressurizer water reactor (PWR), is selected as the simulation tool. The complex inner flow channels and heat sources are simplified according to its thermal–hydraulic characteristics. Then the nodal model for REALP5 is built for approximating the conceptual design. Two typical operating plans, superheated steam scheme and PWR scheme, are analyzed. After some adjustments of the inlet flow resistance coefficients of some flow channels, the reasonable stable conditions of both operation plans can be obtained. The stable fluid and wall temperature distributions and pressure drops are studied. At last, a supposed slow flow decreasing is discussed under two operating conditions separately. According to present results, the superheated steam scheme still needs to be further optimized. The PWR scheme shows a very good safety feature.

  12. The preliminary thermal–hydraulic analysis of a water cooled blanket concept design based on RELAP5 code

    International Nuclear Information System (INIS)

    Wang, Guanghuai; Peng, Changhong; Guo, Yun

    2016-01-01

    Highlights: • The superheated steam and PWR schemes are analyzed by RELAP5 code. • The influence of non-uniform heating sources is include. • A supposed slow flow decrease case is discussed and the PWR scheme is better. - Abstract: Water cooled blanket (WCB) is very important in the conceptual design and energy transfer in future fusion power plant. One conceptual design of WCB is under computational testing. RELAP5 code, which is mature and often used in transient analysis in Pressurizer water reactor (PWR), is selected as the simulation tool. The complex inner flow channels and heat sources are simplified according to its thermal–hydraulic characteristics. Then the nodal model for REALP5 is built for approximating the conceptual design. Two typical operating plans, superheated steam scheme and PWR scheme, are analyzed. After some adjustments of the inlet flow resistance coefficients of some flow channels, the reasonable stable conditions of both operation plans can be obtained. The stable fluid and wall temperature distributions and pressure drops are studied. At last, a supposed slow flow decreasing is discussed under two operating conditions separately. According to present results, the superheated steam scheme still needs to be further optimized. The PWR scheme shows a very good safety feature.

  13. Design of the upper internals structure for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Thompson, D.C.; Novendstern, E.H.

    1977-01-01

    The Upper Internals Structure (UIS) is located above the core and is supported from the head at four locations. It is designed to perform the following primary functions: provide secondary core holddown in the event of a malfunction of the core hydraulic holddown system; provide support for routing all in-vessel instrumentation to core assemblies; maintain alignment between the core assemblies, the UIS and the closure head; provide guidance and crossflow protection for the control rod drivelines; and mix/duct flow to the upper region of the vessel outlet plenum to minimize rapid temperature changes to components during a reactor trip transient. In accomplishing these functions, the UIS will experience a sodium environment with temperatures up to 1200 0 F (649 0 C), and as many as 7 x 10 8 cycles of fluid temperature fluctuations up to 250 0 F (121 0 C) at full power operation. It must be designed to survive these conditions in combination with seismic and flow-induced vibration loadings for its 30 year design life. The design program of designing to controlled functional requirements and design conditions is discussed. Included is a description of the significant parts of the design and the approach used to balance the requirement of tight joints. The thermal and hydraulic environment including the results of a comprehensive test program are discussed. The test program results establish the basis of the thermal boundary used in the structural evaluation, and the UIS vibration characteristics. A summary of the areas which have required design changes is included with a summary of the structural evaluation of these changes

  14. Analysis of a sustainable gas cooled fast breeder reactor concept

    International Nuclear Information System (INIS)

    Kumar, Akansha; Chirayath, Sunil S.; Tsvetkov, Pavel V.

    2014-01-01

    Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232 Th and 233 U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239 Pu and 233 U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232 Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235 U and nuclear waste management issues. A mixture of thorium and uranium ( 232 Th + 233 U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239 Pu. Initial feed of 233 U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233 U a theoretical evaluation has been used to derive the data for the source of 233 U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233 U as an alternate to 235 U and an alternate use for light water reactor used fuel as a

  15. Seismic design principles for the German fast breeder reactor SNR2

    International Nuclear Information System (INIS)

    Rangette, A.M.; Peters, K.A.

    1988-01-01

    The leading aim of a seismic design is, besides protection against seismic impacts, not to enhance the overall risk in the absence of seismic vibrations and, secondly, to avoid competition between operational needs and a seismic structural design. This approach is supported by avoiding overconservatism in the assumption of seismic loads and in the calculation of the structural response. Accordingly the seismic principles are stated as follows: restriction to German or equivalent low seismicity sites with intensities (SSE) lower VIII at frequency lower than 10 -4 /year; best estimate of seismic input-data without further conservatism; no consideration of OBE. The structural design principles are: 1. The secondary character of the seismic excitation is explicitly accounted for; 2. Energy absorption is allowed for by ductility of materials and construction. Accordingly strain criteria are used for failure predictions instead of stress criteria. (author). 1 fig

  16. Tritium transport and release from lithium ceramic breeder materials

    International Nuclear Information System (INIS)

    Johnson, C.E.; Kopasz, J.P.; Tam, S.W.

    1994-01-01

    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

  17. Influence of design features on decommissioning of a large fast breeder reactor

    International Nuclear Information System (INIS)

    Fournie, J.-L.; Alary, C.; Maire, D.; Seroux, N. de; Peyrard, G.

    1990-01-01

    The evolution of FBR design in Europe shows that pool-type design will become the reference design for future FBR and the projected European Fast Reactor (EFR) is based on this concept. The identification of design features shows that the main contributors of the sodium and structures activity are the Co60 for gamma radiation source and low decay, Ni63, Nb94 and Ni59 for long time decay. So, the technical benefits of a Co content reduction are interesting for the high activated structures and for diagrid thimbles coating and we made proposals to lower Co content in steels or alloys and to substitute coatings. We identify measures which must facilitate both the sodium draining and the reactor block and internal cleaning: all which improve the gravity draining and the downing of the sodium flow make easier the penetration of cleaning products. The features, connected with the dismantling of the very activated internal structures, of the roof and of the lay-out, are mentioned. (author)

  18. UF6 breeder reactor power plants for electric power generation

    International Nuclear Information System (INIS)

    Rust, J.H.; Clement, J.D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a 233 UF 6 core surrounded by a molten salt (Li 7 F, BeF 2 , ThF 4 ) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

  19. Seismic design principles for the German fast breeder reactor SNR 2

    International Nuclear Information System (INIS)

    Busch, K.A.; Peters, K.A.; Rosenhauer, W.

    1987-01-01

    The safety issue of an adequate and optimized external event protection is of course that unnecessary hardware precautions might promote internal disturbances or hamper their control. It has up to now not satisfactorily been realized that the only serious context for seismic impacts on a fast reactor is their attributed potential of overriding core disruptive accident prevention, see e.g. GRS 1982. General and exaggerated antiseismic design features not focussed upon this point may as well turn out to be non-negligible initators in the absence of seismic vibrations. Unexpected snubber difficulties requiring additional reactor scrams and decay heat removal phases may be named as a simple example. The presented seismic design principles reflect the progress made in the concerned fields of analysis and do serve on the other hand as guidelines for research and development efforts under work. (orig./GL)