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Sample records for future fusion blankets

  1. Blanket handling concepts for future fusion power plants

    International Nuclear Information System (INIS)

    Bogusch, E.; Gottfried, R.; Maisonnier, D.

    2003-01-01

    In the frame of the power plant conceptual studies (PPCS) launched by the European Commission, two main blanket handling concepts have been investigated with respect to engineering feasibility and the impact on the plant availability and on cost: the large module handling concept (LMHC) and the large sector handling concept (LSHC). The LMHC has been considered as the reference handling concept while the LSHC has been considered as an attractive alternative to the LMHC due to its potential of smaller replacement times and hence increasing the plant availability. Although no principle feasibility issue has been identified, a number of engineering issues have been highlighted for the LSHC that would require considerable efforts for their resolution. Since its availability of about 77% based on a replacement time for all the internals of about 4.2 months is slightly lower than for the LMHC, the LMHC remains the reference blanket replacement concept for a conceptual reactor

  2. 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

  3. 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)

  4. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Keough, R.F.

    1981-01-01

    Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

  5. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1986-06-01

    The fusion blanket surrounds the burning hydrogen core of the fusion reactor. It is in this blanket that most of the energy released by the DT fusion reaction is converted into useable product, and where tritium fuel is produced to enable further operation of the reactor. Blankets will involve new materials, conditions and processes. Several recent fusion blanket concepts are presented to illustrate the range of ideas

  6. Fusion blanket inherent safety assessment

    International Nuclear Information System (INIS)

    Sze, D.K.; Jung, J.; Cheng, E.T.

    1986-01-01

    Fusion has significant potential safety advantages. There is a strong incentive for designing fusion plants to ensure that inherent safety will be achieved. Accordingly, both the Tokamak Power Systems Studies and MINIMARS have identified inherent safety as a design goal. A necessary condition is for the blanket to maintain its configuration and integrity under all credible accident conditions. A main problem is caused by afterheat removal in an accident condition. In this regard, it is highly desirable to achieve the required level of protection of the plant capital investment and limitation of radioactivity release by systems that rely only on inherent properties of matter (e.g., thermal conductivity, specific heat, etc.) and without the use of active safety equipment. This paper assesses the conditions under which inherent safety is feasible. Three types of accident conditions are evaluated for two blankets. The blankets evaluated are a self cooled vanadium/lithium blanket and a self-cooled vanadium/Flibe blanket. The accident conditions evaluated are: (1) loss-of-flow accident; (2) loss-of-coolant accident (LOCA); and (3) partial loss-of-coolant accident

  7. 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

  8. 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

  9. The fusion blanket program at Chalk River

    International Nuclear Information System (INIS)

    Hastings, I.J.

    1986-03-01

    Work on the Fusion Blanket Program commenced at Chalk River in 1984 June. Co-funded by Canadian Fusion Fuels Technology Project and Atomic Energy of Canada Limited, the Program utilizes Chalk River expertise in instrumented irradiation testing, ceramics, tritium technology, materials testing and compound chemistry. This paper gives highlights of studies to date on lithium-based ceramics, leading contenders for the fusion blanket

  10. Neutronic investigations on the application of lithium aluminates in the tritium breeding blanket of future fusion reactors

    International Nuclear Information System (INIS)

    Mohsin, A.

    1981-02-01

    A survey is given about the state of development work at the blanket. It shows that present designs aim at a fusion reactor with low tritium inventory. This aim can be achieved with a solid blanket. In this paper this concept is described and the selection of appropriate materials for the solid blanket is discussed. The lithium aluminates turned out to be the most suitable materials. Comparing the different lithium aluminates the compounds Li 5 AlO 4 and LiAlO 2 proved to be the most favourable. The improvement of the breeding ratio when using lead as neutron multiplier was investigated. Employing, for example, a lead zone of 15 cm thickness in front of a 60 cm thick breeding zone, the tritium breeding ratio is raised to 1.65 for Li 5 Al 4 and to 1.48 for LiAlO 2 - The originally higher breeding ratio of the Li 5 AlO 4 in contrary to the LiAlO 2 is compensated hereby. By this LiAlO 2 becomes a very interesting material for a solid blanket since it furthermore exhibits a higher melting point and higher phase transition temperature. For experimental check of the nuclear data of this material and the computational techniques used, a test model was designed and built. This blanket model was used for measuring the space-dependent tritium production rate, which could be compared to corresponding computations. The assembly was made of a lead zone as neutron multiplier, LiAlO 2 as breeding material, and polyethylene as neutron reflector. (orig.) [de

  11. 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

  12. 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

  13. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Sawan, M.E.; Cheng, E.T.

    1985-01-01

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li/sub 17/Pb/sub 83/ and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the TBR to group structure and weighting spectrum increases and Li enrichment decrease with up to 20% discrepancies for thin natural Li/sub 17/Pb/sub 83/ blankets

  14. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Sawan, M.L.; Cheng, E.T.

    1986-01-01

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li 17 Pb 83 and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the tritium breeding ratio to group structure and weighting spectrum increases as the thickness and Li enrichment decrease with up to 20% discrepancies for thin natural Li 17 Pb 83 blankets. (author)

  15. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Young, J.R.

    1975-01-01

    Comparisons of alternative fusion reactor blanket/coolant systems suggest that environmental considerations will enter strongly into selection of design and materials. Liquid blankets and coolants tend to maximize transport of radioactive corrosion products. Liquid lithium interacts strongly with tritium, minimizing permeation and escape of gaseous tritium in accidents. However, liquid lithium coolants tend to create large tritium inventories and have a large fire potential compared to flibe and solid blankets. Helium coolants minimize radiation transport, but do not have ability to bind the tritium in case of accidental releases. (auth)

  16. 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

  17. Optimization of beryllium for fusion blanket applications

    International Nuclear Information System (INIS)

    Billone, M.C.

    1993-01-01

    The primary function of beryllium in a fusion reactor blanket is neutron multiplication to enhance tritium breeding. However, because heat, tritium and helium will be generated in and/or transported through beryllium and because the beryllium is in contact with other blanket materials, the thermal, mechanical, tritium/helium and compatibility properties of beryllium are important in blanket design. In particular, tritium retention during normal operation and release during overheating events are safety concerns. Accommodating beryllium thermal expansion and helium-induced swelling are important issues in ensuring adequate lifetime of the structural components adjacent to the beryllium. Likewise, chemical/metallurgical interactions between beryllium and structural components need to be considered in lifetime analysis. Under accident conditions the chemical interaction between beryllium and coolant and breeding materials may also become important. The performance of beryllium in fusion blanket applications depends on fabrication variables and operational parameters. First the properties database is reviewed to determine the state of knowledge of beryllium performance as a function of these variables. Several design calculations are then performed to indicate ranges of fabrication and operation variables that lead to optimum beryllium performance. Finally, areas for database expansion and improvement are highlighted based on the properties survey and the design sensitivity studies

  18. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    Carvalho, S.H. de.

    1980-03-01

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

  19. 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.)

  20. 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

  1. Utilization of fusion neutrons in the tokamak fusion test reactor for blanket performance testing and other nuclear engineering experiments

    International Nuclear Information System (INIS)

    Caldwell, C.S.; Pettus, W.G.; Schmotzer, J.K.; Welfare, F.; Womack, R.

    1979-01-01

    In addition to developing a set of reacting-plasma/blanket-neutronics benchmark data, the TFTR fusion application experiments would provide operational experience with fast-neutron dosimetry and the remote handling of blanket modules in a tokamak reactor environment; neutron streaming and hot-spot information invaluable for the optimal design of penetrations in future fusion reactors; and the identification of the most damage-resistant insulators for a variety of fusion-reactor components

  2. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

    Le Marois, G.; Federzoni, L.; Bucci, P.; Revirand, P.

    2000-01-01

    The benefit of HIP techniques applied to the fabrication of fusion internal components for higher performances, reliability and cost savings are emphasized. To demonstrate the potential of the techniques, design of new blankets concepts and mock-ups fabrication are currently performed by CEA. A coiled tube concept that allows cooling arrangement flexibility, strong reduction of the machining and number of welds is proposed for ITER IAM. Medium size mock-ups according to the WCLL breeding blanket concept have been manufactured. The fabrication of a large size mock-up is under progress. These activities are supported by numerical calculations to predict the deformations of the parts during HIP'ing. Finally, several HIP techniques issues have been identified and are discussed

  3. Structural materials for fusion reactor blanket systems

    International Nuclear Information System (INIS)

    Bloom, E.E.; Smith, D.L.

    1984-01-01

    Consideration of the required functions of the blanket and the general chemical, mechanical, and physical properties of candidate tritium breeding materials, coolants, structural materials, etc., leads to acceptable or compatible combinations of materials. The presently favored candidate structural materials are the austenitic stainless steels, martensitic steels, and vanadium alloys. The characteristics of these alloy systems which limit their application and potential performance as well as approaches to alloy development aimed at improving performance (temperature capability and lifetime) will be described. Progress towards understanding and improving the performance of structural materials has been substantial. It is possible to develop materials with acceptable properties for fusion applications

  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. 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

  6. Nuclear characteristics of D-D fusion reactor blankets

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao

    1978-01-01

    Fusion reactors operating on deuterium (D-D) cycle are considered to be of long range interest for their freedom from tritium breeding in the blanket. The present paper discusses the various possibilities of D-D fusion reactor blanket designs mainly from the standpoint of the nuclear characteristics. Neutronic and photonic calculations are based on presently available data to provide a basis of the optimal blanket design in D-D fusion reactors. It is found that it appears desirable to design a blanket with blanket/shield (BS) concept in D-D fusion reactors. The BS concept is designed to obtain reasonable shielding characteristics for superconducting magnet (SCM) by using shielding materials in the compact blanket. This concept will open the possibility of compact radiation shield design based on assured technology, and offer the advantage from the system economics point of view. (auth.)

  7. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1983-06-01

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matrices. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

  8. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1984-02-01

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matricies. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

  9. 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

  10. Li2O-pebble type tritium breeding blanket for fusion experimental reactor, 1

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Iida, Hiromasa; Tanaka, Yoshihisa

    1984-01-01

    The fusion experimental reactor is the next stage device in Japan, which is planned to be constructed following the critical plasma experimental device JT-60 being constructed at present. The breeding blanket installed in nuclear fusion reactors is one of most important structures, and it is required to satisfy the fundamental performance of producing and continuously recovering tritium as the nuclear fusion fuel, and other requirement in good coordination. The Li 2 O pebble type breeding blanket that Kawasaki Heavy Industries Ltd. has examined is the concept for resolving the problems of the mass transfer and thermal stress cracking of Li 2 O, which are important in blanket design. In this paper, the concept and characteristics of this breeding blanket are discussed from the viewpoint of the breeding and continuous recovery of tritium, the ease of manufacture and the maintenance of soundness. The breeding blanket is composed of breeding region, tritium purge region, cooling region, plasma stabilizing conductors and blanket container. Li 2 O is excellent in its tritium breeding performance and heat conductivity. The functions required for the breeding blanket, the fundamental structure, the examples of breeding blanket concept, the selection of breeding blanket concept, the characteristics of Li 2 O pebble type blanket and its future prospect are described. (Kako, I.)

  11. Experimental Investigation of Ternary Alloys for Fusion Breeding Blankets

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-26

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

  12. 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

  13. 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.)

  14. Minimum thickness blanket-shield for fusion reactors

    International Nuclear Information System (INIS)

    Karni, Y.; Greenspan, E.

    1989-01-01

    A lower bound on the minimum thickness fusion reactor blankets can be designed to have, if they are to breed 1.267 tritons per fusion neutron, is identified by performing a systematic nucleonic optimization of over a dozen different blanket concepts which use either Be, Li 17 Pb 83 , W or Zr for neutron multiplication. It is found that Be offers minimum thickness blankets; that the blanket and shield (B/S) thickness of Li 17 Pb 83 based blankets which are supplemented by Li 2 O and/or TiH 2 are comparable to the thickness of Be based B/S; that of the Be based blankets, the aqueous self-cooled one offers one of the most compact B/S; and that a number of blanket concepts might enable the design of B/S which is approximately 12 cm and 39 cm thinner than the B/S thickness of, respectively, conventional self-cooled Li 17 Pb 83 and Li blankets. Aqueous self-cooled tungsten blankets could be useful for experimental fusion devices provided they are designed to be heterogeneous. (orig.)

  15. 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

  16. Fusion blankets for high efficiency power cycles

    International Nuclear Information System (INIS)

    Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Usher, J.L.

    1980-04-01

    Definitions are given of 10 generic blanket types and the specific blanket chosen to be analyzed in detail from each of the 10 types. Dimensions, compositions, energy depositions and breeding ratios (where applicable) are presented for each of the 10 designs. Ultimately, based largely on neutronics and thermal hyraulics results, breeding an nonbreeding blanket options are selected for further design analysis and integration with a suitable power conversion subsystem

  17. Feasibility study of fusion breeding blanket concept employing graphite reflector

    International Nuclear Information System (INIS)

    Cho, Seungyon; Ahn, Mu-Young; Lee, Cheol Woo; Kim, Eung Seon; Park, Yi-Hyun; Lee, Youngmin; Lee, Dong Won

    2015-01-01

    Highlights: • A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. • Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. • A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. • Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. • In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. - Abstract: To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.

  18. Feasibility study of fusion breeding blanket concept employing graphite reflector

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Seungyon, E-mail: sycho@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Cheol Woo; Kim, Eung Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi-Hyun; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Highlights: • A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. • Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. • A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. • Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. • In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. - Abstract: To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.

  19. Blanket options for high-efficiency fusion power

    International Nuclear Information System (INIS)

    Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

  20. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

    Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

  1. Fusion blanket for high-efficiency power cycles

    International Nuclear Information System (INIS)

    Usher, J.L.; Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Taussig, R.

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperature (500 0 C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by Ar) utilizing Li 2 O for tritium breeding. In this design, approx. 60% of the fusion energy is deposited in the high-temperature interior. The maximum Ar temperature is 2230 0 C leading to an overall efficiency estimate of 55 to 60% for this reference case

  2. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

    Usher, J.L.; Lazareth, O.W.; Fillo, J.A.; Horn, F.L.; Powell, J.R.

    1981-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperatures (500 deg C) of conventional structural materials such as stainless steels. In this project 'two-zone' blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO 2 interior (cooled by argon) utilizing Li 2 O for tritium breeding. In this design, approximately 60% of the fusion energy is deposited in the high-temperature interior. The maximum argon temperature is 2230 deg C leading to an overall efficiency estimate of 55 to 60% for this reference case. (author)

  3. Nuclear characteristics of D-D fusion reactor blankets, (1)

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao; Seki, Yasushi.

    1977-01-01

    Fusion reactors operating on the deuterium (D-D) cycle are considered promising for their freedom from tritium breeding in the blanket. In this paper, neutronic and photonic calculations are undertaken covering several blanket models of the D-D fusion reactor, using presently available data, with a view to comparing the nuclear characteristics of these models, in particular, the nuclear heating rates and their spatial distributions. Nine models are taken up for the study, embodying various combinations of coolant, blanket, structural and reflector materials. About 10 MeV is found to be a typical value for the total nuclear energy deposition per source neutron in the models considered here. The realization of high energy gain is contingent upon finding a favorable combination of blanket composition and configuration. The resulting implications on the thermal design aspect are briefly discussed. (auth.)

  4. Application of vanadium alloys to a fusion reactor blanket

    Energy Technology Data Exchange (ETDEWEB)

    Bethin, J.; Tobin, A. (Grumman Aerospace Corp., Bethpage, NY (USA). Research and Development Center)

    1984-05-01

    Vanadium and vanadium alloys are of interest in fusion reactor blanket applications due to their low induced radioactivity and outstanding elevated temperature mechanical properties during neutron irradiation. The major limitation to the use of vanadium is its sensitivity to oxygen impurities in the blanket environment, leading to oxygen embrittlement. A quantitative analysis was performed of the interaction of gaseous impurities in a helium coolant with vanadium and the V-15Cr-5Ti alloy under conditions expected in a fusion reactor blanket. It was shown that the use of unalloyed V would impose severe restrictions on the helium gas cleanup system due to excessive oxygen buildup and embrittlement of the metal. However, internal oxidation effects and the possibly lower terminal oxygen solubility in the alloy would impose much less severe cleanup constraints. It is suggested that V-15Cr-5Ti is a promising candidate for certain blanket applications and deserves further consideration.

  5. 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

  6. Cassette blanket and vacuum building: key elements in fusion reactor maintenance

    International Nuclear Information System (INIS)

    Werner, R.W.

    1977-01-01

    The integration of two concepts important to fusion power reactors is discussed. The first concept is the vacuum building which improves upon the current fusion reactor designs. The second concept, the use of the cassette blanket within the vacuum building environment, introduces four major improvements in blanket design: cassette blanket module, zoning concept, rectangular blanket concept, and internal tritium recovery

  7. Fusion blanket high-temperature heat transfer

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1983-01-01

    Deep penetration of 14 MeV neutrons makes two-temperature region blankets feasible. A relatively low-temperature (approx. 300 0 C) metallic structure is the vacuum/coolant pressure boundary, while the interior of the blanket, which is a simple packed bed of nonstructural material, operates at very high temperatures (>1000 0 C). The water-cooled shell structure is thermally insulated from the steam-cooled interior. High-temperature steam can dramatically increase the efficiency of electric power generation, as well as produce hydrogen and oxygen-based synthetic fuels at high-efficiency

  8. 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

  9. Fusion-reactor blanket-material safety-compatibility studies

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Muhlestein, L.D.; Keough, R.F.; Cohen, S.

    1982-11-01

    Blanket material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Blanket material safety compatibility studies are being conducted to identify and characterize blanket-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate that : (1) ternary oxides (LiAlO 2 , Li 2 ZrO 3 , Li 2 SiO 3 , Li 4 SiO 4 and LiTiO 3 ) at postulated blanket operating temperatures are compatible with water coolant, while liquid lithium and Li 7 Pb 2 alloy reactions with water generate heat, aerosol and hydrogen; (2) lithium oxide and Li 17 Pb 83 alloy react mildly with water requiring special precautions to control hydrogen release; (3) liquid lithium reacts substantially, while Li 17 Pb 83 alloy reacts mildly with concrete to produce hydrogen; and (4) liquid lithium-air reactions present some major safety concerns

  10. Aqueous self-cooled blanket concepts for fusion reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

    A novel aqueous self-cooled blanket (ASCB) concept has been proposed. The water coolant also serves as the tritium breeding medium by dissolving small amounts of lithium compound in the water. The tritium recovery requirements of the ASCB concept may be facilitated by the novel in-situ radiolytic tritium separation technique in development at Chalk River Nuclear Laboratories. In this separation process deuterium gas is bubbled through the blanket coolant. Due to radiation induced processes, the equilibrium constant favors tritium migration to the deuterium gas stream. It is expected that the inherent simplicity of this design will result in a highly reliable, safe and economically attractive breeding blanket for fusion reactors. The available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate the proposed blanket concept. Tests for tritium separation and corrosion compatibility show encouraging results for the feasibility of this concept

  11. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2014-01-01

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM

  12. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tariq Siddique, M.; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of)

    2014-05-15

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM.

  13. 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

  14. Stress analysis of blanket vessel for JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Sako, K.; Minato, A.

    1979-01-01

    A blanket structure of JAERI Experimental Fusion Reactor (JXFR) consists of about 2,300 blanket cells with round cornered rectangular cross sections (twelve slightly different shapes) and is placed in a vacuum vessel. Each blanket vessel is a double-walled thin-shell structure made of Type 316 stainless steel with a spherical domed surface at the plasma side. Ribs for coolant channel are provided between inner and outer walls. The blanket cell contains Li 2 O pebbles and blocks for tritium breeding and stainless steel blocks for neutron reflection. A coolant is helium gas at 10 kgf/cm 2 (0.98 MPa) and its inlet and outlet temperatures are 300 0 C and 500 0 C. The maxima of heat flux and nuclear heating rate at the first wall are 12 W/cm 2 and 2 W/cc. A design philosophy of the blanket structure is based on high tritium breeding ratio and more effective shielding performance. The thin-shell vessel with a rectangular cross section satisfies the design philosophy. We have designed the blanket structure so that the adjacent vessels are mutually supporting in order to decrease the large deformation and stress due to internal pressure in case of the thin-shell vessel. (orig.)

  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. 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

  17. 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.).

  18. 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

  19. Direct LiT Electrolysis in a Metallic Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-30

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

  20. Direct LiT Electrolysis in a Metallic Fusion Blanket

    International Nuclear Information System (INIS)

    Olson, Luke

    2016-01-01

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

  1. Evaluation of the activity levels in fusion reactor blankets

    International Nuclear Information System (INIS)

    Gruber, J.

    1977-05-01

    The activation of a fusion reactor blanket (316 SS or V-10Cr-10Ti as structure) with a minimum lithium inventory has been calculated for 0.83 MW/m 2 wall load. The resulting radiation levels and waste problems are discussed. The dose rate near the steel structure will always be higher than 0.1 rem/h due to its niobium content. After 200 to 100,000 years of decay the potential biological hazard originating from this high level fusion reactor waste (with plutonium recyclation). (orig.) [de

  2. 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.)

  3. Structural performance of a graphite blanket in fusion reactors

    International Nuclear Information System (INIS)

    Wolfer, W.G.; Watson, R.D.

    1978-01-01

    Irradiation of graphite in a fusion reactor causes dimensional changes, enhanced creep, and changes in elastic properties and fracture strength. Temperature and flux gradients through the graphite blanket structure produce differential distortions and stress gradients. An inelastic stress analysis procedure is described which treats these variations of the graphite properties in a consistent manner as dictated by physical models for the radiation effects. Furthermore, the procedure follows the evolution of the stress and fracture strength distributions during the reactor operation as well as for possible shutdowns at any time. The lifetime of the graphite structure can be determined based on the failure criterion that the stress at any location exceeds one-half of the fracture strength. This procedure is applied to the most critical component of the blanket module in the SOLASE design

  4. 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)

  5. Modeling and experiments on tritium permeation in fusion reactor blankets

    Science.gov (United States)

    Holland, D. F.; Longhurst, G. R.

    The determination of tritium loss from helium-cooled fusion breeding blankets are discussed. The issues are: (1) applicability of present models to permeation at low tritium pressures; (2) effectiveness of oxide layers in reducing permeation; (3) effectiveness of hydrogen addition as a means to lower tritium permeation; and (4) effectiveness of conversion to tritiated water and subsequent trapping to reduce permeation. Theoretical models applicable to these issues are discussed, and results of experiments in two areas are presented; permeation of mixtures of hydrogen isotopes and conversion to tritiated water.

  6. Modeling and experiments on tritium permeation in fusion reactor blankets

    International Nuclear Information System (INIS)

    Holland, D.F.; Longhurst, G.R.

    1985-01-01

    Issues are discussed that are critical in determining tritium loss from helium-cooled fusion breeding blankets. These issues are: (a) applicability of present models to permeation at low tritium pressures, (b) effectiveness of oxide layers in reducing permeation, (c) effectiveness of hydrogen addition as a means to lower tritium permeation, and (d) effectiveness of conversion to tritiated water and subsequent trapping as a means to reduce permeation. The paper discusses theoretical models applicable to these issues, and presents results of experiments in two areas: permeation of mixtures of hydrogen isotopes and conversion to tritiated water

  7. First wall fusion blanket temperature variation - slab geometry

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1978-01-01

    The first wall of a fusion blanket is approximated by a slab, with the surface facing the plasma subjected to an applied heat flux, while the rear surface is convectively cooled. The relevant parameters affecting the heat transfer during the early phases of heating as well as for large times are established. Analytical solutions for the temperature variation with time and space are derived. Numerical calculations for an aluminum and stainless steel slab are performed for a wall loading of 1 MW(th)/m 2 . Both helium and water cooling are considered. (Auth.)

  8. Thermalhydraulics of flowing particle-bed-type fusion reactor blankets

    International Nuclear Information System (INIS)

    Nietert, R.E.; Abdelk-Khalik, S.I.

    1982-01-01

    An experimental investigation has been conducted to determine the heat transfer characteristics of gravity-flowing particle beds using a special heat transfer loop. Glass microspheres were allowed to flow by gravity at controlled rates through an electrically heated stainless steel tubular test section. Values of the local and average convective heat transfer coefficient as a function of the average bed velocity, particle size and heat flux were determined. Such information is necessary for the design of gravity-flowing particle-bed type fusion reactor-blankets and associated tritium recovery systems. (orig.)

  9. Magnetohydrodynamic research in fusion blanket engineering and metallurgical processing

    International Nuclear Information System (INIS)

    Tokuhiro, A.

    1991-11-01

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

  10. 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.

  11. 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.

  12. NOEL: a no-leak fusion blanket concept

    International Nuclear Information System (INIS)

    Powell, J.R.; Yu, W.S.; Fillo, J.A.; Horn, F.L.; Makowitz, H.

    1980-01-01

    Analysis and tests of a no-leak fusion blanket concept (NOEL-NO External Leak) are described. Coolant cannot leak into the plasma chamber even if large through-cracks develop in the first wall. Blanket modules contain a two-phase material, A, that is solid (several cm thick) on the inside of the module shell, and liquid in the interior. The solid layer is maintained by imbedded tubes carrying a coolant, B, below the freezing point of A. Most of the 14-MeV neutron energy is deposited as heat in the module interior. The thermal energy flow from the module interior to the shell keeps the interior liquid. Pressure on the liquid A interior is greater than the pressure on B, so that B cannot leak out if failures occur in coolant tubes. Liquid A cannot leak into the plasma chamber through first wall cracks because of the intervening frozen layer. The thermal hydraulics and neutronics of NOEL blankets have been investigated for various metallic (e.g., Li, Pb 2 , LiPb, Pb) and fused salt choices for material A

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-02

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

  14. The current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

    Veleckis, E.; Yonco, R.M.; Maroni, V.A.

    1980-01-01

    The available thermodynamic information is reviewed for three categories of materials that meet essential criteria for use as breeding blankets in D-T fuelled fusion reactors: liquid lithium, solid lithium alloys, and lithium-containing ceramics. The leading candidate, liquid lithium, which also has potential for use as a coolant, has been studied more extensively than have the solid alloys or ceramics. Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries, etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phases of the Li-LiH, Li-LiD and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li 3 N, Li 2 O, and Li 2 C 2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g. Li-Al and Li-Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li-M alloys can be estimated from lithium activity data for these alloys. There is essentially no refined thermodynamic information on the prospective ceramic blanket materials. The kinetics of tritium release from these materials is briefly discussed. Research areas are pointed out where additional thermodynamic information is needed for all three material categories. (author)

  15. JAERI/U.S. collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Mori, Takamasa; Kosako, Kazuaki; Oyama, Yukio; Nakamura, Tomoo

    1989-10-01

    Phase IIa and IIb experiments of JAERI/U.S. Collaborative Program on Fusion Blanket Neutronics have been performed using the FNS facility at JAERI. The phase IIa experimental systems consist of the Li 2 O test region, the rotating neutron target and the Li 2 CO 3 container. In phase IIb, a beryllium layer is added to the inner wall to investigate a multiplier effect. Measured parameters are source characteristics by a foil activation method and spectrum measurements using both NE-213 and proton recoil counters. The measurements inside the Li 2 O region included tritium production rates, reaction rate by foil activation and neutron spectrum measurements. Analysis for these parameters was performed by using two dimensional discrete ordinate codes DOT3.5 and DOT-DD, and a Monte Carlo code MORSE-DD. The nuclear data used were based on JENDL3/PR1 and PR2. ENDF/B-IV, V and the FNS file were used as activation cross sections. The configurations analysed for the test region were a reference, a beryllium front and a beryllium sandwiched systems in phase IIa, and a reference and a beryllium front with first wall systems in phase IIb. This document describes the results of analysis and comparison between the calculations and the measurements. The prediction accuracy of key parameters in a fusion reactor blanket are examined. The tritium production rates can be well predicted in the reference systems but are fairly underestimated in the system with a beryllium multiplier. Details of experiments and the experimental techniques are described separately in the another report. (author)

  16. 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.

  17. 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).

  18. Self-shielding characteristics of aqueous self-cooled blankets for next generation fusion devices

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.; Embrechts, M.J.

    1987-11-01

    The present study examines self-shielding characteristics for two aqueous self-cooled tritium producing driver blankets for next generation fusion devices. The aqueous Self-Cooled Blanket concept (ASCB) is a very simple blanket concept that relies on just structural material and coolant. Lithium compounds are dissolved in water to provide for tritium production. An ASCB driver blanket would provide a low technology and low temperature environment for blanket test modules in a next generation fusion reactor. The primary functions of such a blanket would be shielding, energy removal and tritium production. One driver blanket considered in this study concept relates to the one proposed for the Next European Torus (NET), while the second concept is indicative for the inboard shield design for the Engineering Test Reactor proposed by the USA (TIBER II/ETR). The driver blanket for NET is based on stainless steel for the structural material and aqueous solution, while the inboard shielding blanket for TIBER II/ETR is based on a tungsten/aqueous solution combination. The purpose of this study is to investigate self-shielding and heterogeneity effects in aqueous self-cooled blankets. It is found that no significant gains in tritium breeding can be achieved in the stainless steel blanket if spatial and energy self-shielding effects are considered, and the heterogeneity effects are also insignificant. The tungsten blanket shows a 5 percent increase in tritium production in the shielding blanket when energy and spatial self-shielding effects are accounted for. However, the tungsten blanket shows a drastic increase in the tritium breeding ratio due to heterogeneity effects. (author) 17 refs., 9 figs., 9 tabs

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

    International Nuclear Information System (INIS)

    Chen Yixue; Wu Yican

    2000-01-01

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

  20. Future of fusion implementation

    International Nuclear Information System (INIS)

    Beardsworth, E.; Powell, J.R.

    1978-01-01

    For fusion to become available for commercial use in the 21st century, R and D must be undertaken now. But it is hard to justify these expenditures with a cost/benefit oriented assessment methodology, because of both the time-frame and the uncertainty of the future benefits. Focusing on the factors most relevant for current consideration of fusion's commercial prospects, i.e., consumption levels and the outcomes for fission, solar, and coal, many possible futures of the US energy system are posited and analyzed under various assumptions about costs. The Reference Energy System approach was modified to establish both an appropriate degree of detail and explicit time dependence, and a computer code used to organize the relevant data and to perform calculations of system cost (annual and discounted present value), resource use, and residuals that are implied by the consumptions levels and technology mix in each scenario. Not unreasonable scenarios indicate benefits in the form of direct cost savings, which may well exceed R and D costs, which could be attributed to the implementation of fusion

  1. 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

  2. Thermal safety analysis for pebble bed blanket fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie

    1998-01-01

    Pebble bed blanket hybrid reactor may have more advantages than slab element blanket hybrid reactor in nuclear fuel production and nuclear safety. The thermo-hydraulic calculations of the blanket in the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor developed in China are carried out using the Code THERMIX and auxiliary code. In the calculations different fuel pebble material and steady state, depressurization and total loss of flow accident conditions are included. The results demonstrate that the conceptual design of the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor with dump tank is feasible and safe enough only if the suitable fuel pebble material is selected and the suitable control system and protection system are established. Some recommendations for due conceptual design are also presented

  3. Chemical processing of liquid lithium fusion reactor blankets

    International Nuclear Information System (INIS)

    Weston, J.R.; Calaway, W.F.; Yonco, R.M.; Hines, J.B.; Maroni, V.A.

    1979-01-01

    A 50-gallon-capacity lithium loop constructed mostly from 304L stainless steel has been operated for over 6000 hours at temperatures in the range from 360 to 480 0 C. This facility, the Lithium Processing Test Loop (LPTL), is being used to develop processing and monitoring technology for liquid lithium fusion reactor blankets. Results of tests of a molten-salt extraction method for removing impurities from liquid lithium have yielded remarkably good distribution coefficients for several of the more common nonmetallic elements found in lithium systems. In particular, the equilibrium volumetric distribution coefficients, D/sub v/ (concentration per unit volume of impurity in salt/concentration per unit volume of impurity in lithium), for hydrogen, deuterium, nitrogen and carbon are approx. 3, approx. 4, > 10, approx. 2, respectively. Other studies conducted with a smaller loop system, the Lithium Mini-Test Loop (LMTL), have shown that zirconium getter-trapping can be effectively used to remove selected impurities from flowing lithium

  4. Effect of graphite reflector on activation of fusion breeding blanket

    International Nuclear Information System (INIS)

    Lee, Cheol Woo; Lee, Young-Ouk; Lee, Dong Won; Cho, Seungyon; Ahn, Mu-Young

    2016-01-01

    Highlights: • The graphite reflector concept has been applied in the design of the Korea HCCR TBM for ITER and this concept is also a candidate design option for Korea Demo. • In the graphite reflector, C-14, B-11 and Be-10 are produced after an irradiation. Impurities in both case of beryllium and graphite is dominant in the shutdown dose after an irradiation. • Based on the evaluation, the graphite reflector is a good alternative of the beryllium multiplier in the view of induced activity and shutdown dose. But C-14 produced in the graphite reflector should be considered carefully in the view of radwaste management. - Abstract: Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. In this paper, activity analysis was performed and the effect of graphite reflector in the view of activation was compared to the beryllium multiplier. As a result, it is expected that using the graphite reflector instead of the beryllium multiplier decreases total activity very effectively. But the graphite reflector produces C-14 about 17.2 times than the beryllium multiplier. Therefore, C-14 produced in the graphite reflector is expected as a significant nuclide in the view of radwaste management.

  5. Neutron induced displacement damage in beryllium in the blanket of a (d,t)-fusion reactor

    International Nuclear Information System (INIS)

    Hermanutz, D.

    1995-09-01

    Beryllium is a favoured candidate for a neutron multiplier in solid breeder blankets of fusion reactors. This is mainly due to its low (n, 2n)-reaction threshold and because of its good thermal and mechanical properties. Its behaviour under intense neutron irradiation, however, is a crucial issue for its use in future fusion reactors. Displacement damage in beryllium so far has been calculated both with data related and methodological deficiencies. First of all, there is a need to have accurate cross-section data in order to obtain reliable spectra of primary knock-on atoms (PKA's). Furthermore, there are principal restrictions of the NRT-model in general used to calculate secondary displacements initiated by PKA's. The underlying theory of damage-energy (part of kinetic energy of PKA transferred elastically to matrix atoms) according to Lindhard is strictly valid only for medium and heavy mass ions with moderate energies in targets of the same element. In this work improved damage cross-sections and displacement rates (dpa/s) in beryllium have been calculated based on cross-section data from ENDF/B-VI (with a significantly improved (n, 2n)-evaluation) and on an appropriate treatment of damage-energy that is suitable for fusion relevant damage of light mass materials. ''This work has been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhe and is supported by the European Communities within the European Fusion Technology Program''. (orig.)

  6. Two-phase-flow cooling concept for fusion reactor blankets

    International Nuclear Information System (INIS)

    Bender, D.J.; Hoffman, M.A.

    1977-01-01

    The new two-phase heat transfer medium proposed is a mixture of potassium droplets and helium which permits blanket operation at hih temperature and low pressure, while maintaining acceptable pumping power requirements, coolant ducting size, and blanket structure fractions. A two-phase flow model is described. The helium pumping power and the primary heat transfer loop are discussed

  7. Proceedings of the third specialists' workshop on modeling tritium behaviour in ceramic fusion blankets

    International Nuclear Information System (INIS)

    Werle, H.

    1991-08-01

    The third specialists' workshop on modeling tritium behaviour in ceramic fusion blankets, hosted by Kernforschungszentrum Karlsruhe, was held June 10-11, 1991. The workshop was coordinated through the IEA Annex II implementing agreement on 'Radiation damage in fusion materials'. (orig./WL)

  8. 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

  9. 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)

  10. Fusion: Energy for the future

    International Nuclear Information System (INIS)

    1991-05-01

    Fusion, which occurs in the sun and the stars, is a process of transforming matter into energy. If we can harness the fusion process on Earth, it opens the way to assuring that future generations will not want for heat and electric power. The purpose of this booklet is to introduce the concept of fusion energy as a viable, environmentally sustainable energy source for the twenty-first century. The booklet presents the basic principles of fusion, the global research and development effort in fusion, and Canada's programs for fusion research and development

  11. 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)

  12. Recent developments in fusion first wall, blanket, and shield technology

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1983-01-01

    This brief overview of first wall, blanket and shield technology reviews the changes and trends in important design issues in first wall, blanket and shield design and related technology from the 1970's to the 1980's. The emphasis is on base technology rather than either systems engineering or materials development. The review is limited to the two primary confinement systems, tokamaks and mirrors, and production of electricity as the primary goal for development

  13. Source-to-incident flux relation for a tokamak fusion test reactor blanket module

    International Nuclear Information System (INIS)

    Imel, G.R.

    1982-01-01

    The source-to-incident 14-MeV flux relation for a blanket module on the Tokamak Fusion Test Reactor is derived. It is shown that assumptions can be made that allow an analytical expression to be derived, using point kernel methods. In addition, the effect of a nonuniform source distribution is derived, again by relatively simple point kernel methods. It is thought that the methodology developed is valid for a variety of blanket modules on tokamak reactors

  14. 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.

  15. Comparative analysis of a fusion reactor blanket in cylindrical and toroidal geometry using Monte Carlo

    International Nuclear Information System (INIS)

    Chapin, D.L.

    1976-03-01

    Differences in neutron fluxes and nuclear reaction rates in a noncircular fusion reactor blanket when analyzed in cylindrical and toroidal geometry are studied using Monte Carlo. The investigation consists of three phases--a one-dimensional calculation using a circular approximation to a hexagonal shaped blanket; a two-dimensional calculation of a hexagonal blanket in an infinite cylinder; and a three-dimensional calculation of the blanket in tori of aspect ratios 3 and 5. The total blanket reaction rate in the two-dimensional model is found to be in good agreement with the circular model. The toroidal calculations reveal large variations in reaction rates at different blanket locations as compared to the hexagonal cylinder model, although the total reaction rate is nearly the same for both models. It is shown that the local perturbations in the toroidal blanket are due mainly to volumetric effects, and can be predicted by modifying the results of the infinite cylinder calculation by simple volume factors dependent on the blanket location and the torus major radius

  16. Interactions of D-T neutrons in graphite and lithium blankets of fusion reactors

    International Nuclear Information System (INIS)

    Ofek, R.

    1986-05-01

    The present study deals with integral experiment and calculation of neutron energy spectra in bulks of graphite which is used as a reflector in blankets of fusion reactors, and lithium, the material of the blanket on which lithium is bred due to neutron interactions. The collimated beam configuration enables - due to the almost monoenergeticity and unidirectionality of the neutrons impinging on the target - to identify fine details in the measured spectra, and also facilitates the absolute normalization of the spectra. The measured and calculated spectra are generally in a good agreement and in a very good agreement at mesh points close to the system axis. A few conclusions may be drawn: a) the collimated beam source configuration is a sensitive tool for measuring neutron energy spectra with a high resolution, b) the method of unfolding proton-recoil spectra measured with a NE-213 scintillator should be improved, c) MCNP and DOT 4.2 may be used as complementary codes for neutron transport calculations of fusion blankets and deep-penetration problems, d) the updating of the cross-section libraries and checking by integral experiments is highly important for the design of fusion blankets. The present study may be regarded as an important course in the research and development of tools for the design of fusion blankets

  17. Status of fusion reactor blanket evaluation studies in France

    International Nuclear Information System (INIS)

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

    1985-03-01

    In the frame of recent CEA studies aiming at the evaluation and at the comparison of various candidate blanket concepts in moderate power conditions (Psub(n) approximately 2 MW/m 2 ), the present work examines the neutronic and thermomechanical performances of a water cooled Li 17 Pb 83 tubular blanket and those of a helium cooled canister blanket taking advantage of the excellent breeding capability of composite Beryllium/LiAlO 2 (85/15%) breeder elements. The purpose of the following discussion is to justify the impetus for these reference concepts and to summarize the state of their evaluation studies updated by the continuous assimilation of calculations and experiments in progress

  18. Limiter and first wall of the fusion reactor blanket

    International Nuclear Information System (INIS)

    Danilov, I.; Skladnov, K.; Kolganov, V.

    1994-01-01

    Previous designing of the first wall and limiter has allowed to determine their possible embodiment depending on the parameters and operation conditions of the blanket. As a rule limiter is a separate structure located on the plasma facing surface of the blanket assembly. Possible versions of the limiter/FW which may be considered: (1) limiters with mechanical attachment of the protective part; (2) limiters with the attachment with brazing; (3) limiters with common/separate cooling system; (4) limiter as a substitute of the FW. Generally the FW/limiter structure includes protective shield and its cooling system which consist of protective coating, heat accumulator, conductive layer and attachment locks

  19. 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)

  20. 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)

  1. 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

  2. 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)

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

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

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

  4. Enhanced fuel production in thorium fusion hybrid blankets utilizing uranium multipliers

    International Nuclear Information System (INIS)

    Pitulski, R.H.; Chapin, D.L.; Klevans, E.

    1979-01-01

    The multiplication of 14 MeV D-T fusion neutrons via (n,2n), (n,3n), and fission reactions by 238 U is well known and established. This study consistently evaluates the effectiveness of a depleted (tails) UO 2 multiplier on increasing the production of 233 U and tritium in a thorium/lithium fusion--fission hybrid blanket. Nuclear performance is evaluated as a function of exposure and zone thickness

  5. 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).

  6. 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).

  7. Study on fission blanket fuel cycling of a fusion-fission hybrid energy generation system

    International Nuclear Information System (INIS)

    Zhou, Z.; Yang, Y.; Xu, H.

    2011-01-01

    This paper presents a preliminary study on neutron physics characteristics of a light water cooled fission blanket for a new type subcritical fusion-fission hybrid reactor aiming at electric power generation with low technical limits of fission fuel. The major objective is to study the fission fuel cycling performance in the blanket, which may possess significant impacts on the feasibility of the new concept of fusion-fission hybrid reactor with a high energy gain (M) and tritium breeding ratio (TBR). The COUPLE2 code developed by the Institute of Nuclear and New Energy Technology of Tsinghua University is employed to simulate the neutronic behaviour in the blanket. COUPLE2 combines the particle transport code MCNPX with the fuel depletion code ORIGEN2. The code calculation results show that soft neutron spectrum can yield M > 20 while maintaining TBR >1.15 and the conversion ratio of fissile materials CR > 1 in a reasonably long refuelling cycle (>five years). The preliminary results also indicate that it is rather promising to design a high-performance light water cooled fission blanket of fusion-fission hybrid reactor for electric power generation by directly loading natural or depleted uranium if an ITER-scale tokamak fusion neutron source is achievable.

  8. Two-dimensional cross-section sensitivity and uncertainty analysis for fusion reactor blankets

    International Nuclear Information System (INIS)

    Embrechts, M.J.

    1982-02-01

    A two-dimensional sensitivity and uncertainty analysis for the heating of the TF coil for the FED (fusion engineering device) blanket was performed. The uncertainties calculated are of the same order of magnitude as those resulting from a one-dimensional analysis. The largest uncertainties were caused by the cross section uncertainties for chromium

  9. Fusion technology development: first wall/blanket system and component testing in existing nuclear facilities

    International Nuclear Information System (INIS)

    Hsu, P.Y.S.; Bohn, T.S.; Deis, G.A.; Judd, J.L.; Longhurst, G.R.; Miller, L.G.; Millsap, D.A.; Scott, A.J.; Wessol, D.E.

    1980-12-01

    A novel concept to produce a reasonable simulation of a fusion first wall/blanket test environment employing an existing nuclear facility, the Engineering Test Reactor at the Idaho National Engineering Laboratory, is presented. Preliminary results show that an asymmetric, nuclear test environment with surface and volumetric heating rates similar to those expected in a fusion first wall/blanket or divertor chamber surface appears feasible. The proposed concept takes advantage of nuclear reactions within the annulus of an existing test space (15 cm in diameter and approximately 100 cm high) to provide an energy flux to the surface of a test module. The principal reaction considered involves 3 He in the annulus as follows: n + 3 He → p + t + 0.75 MeV. Bulk heating in the test module is accomplished by neutron thermalization, gamma heating, and absorption reactions involving 6 Li in the blanket breeding region. The concept can be extended to modified core configurations that will accommodate test modules of different sizes and types. It makes possible development testing of first wall/blanket systems and other fusion components on a scale and in ways not otherwise available until actual high-power fusion reactors are built

  10. First-wall and blanket engineering development for magnetic-fusion reactors

    International Nuclear Information System (INIS)

    Baker, C.; Herman, H.; Maroni, V.; Turner, L.; Clemmer, R.; Finn, P.; Johnson, C.; Abdou, M.

    1981-01-01

    A number of programs in the USA concerned with materials and engineering development of the first wall and breeder blanket systems for magnetic-fusion power reactors are described. Argonne National Laboratory has the lead or coordinating role, with many major elements of the research and engineering tests carried out by a number of organizations including industry and other national laboratories

  11. Collection of summaries of reports on result of research at basic experiment device for nuclear fusion reactor blanket design, 1995

    International Nuclear Information System (INIS)

    1996-07-01

    This report meeting was held on May 22, 1995 at University of Tokyo by about 40 participants. As the topics on the fusion reactor engineering research in Japan, lectures were given on the present state and future of nuclear fusion networks and on the strong magnetic field tokamak using electromagnetic force-balanced coils being planned. Thereafter, the reports of the results of the researches which were carried out by using this experimental facility were made, centering around the subject related to the future conception 'The interface properties of fusion reactor materials and particle transport control'. The publication was made on the future conception of the basic experiment setup for fusion reactor blanket design, the application of high temperature superconductors to the advancement of nuclear fusion reactors, the modeling of the dynamic irradiation behavior of fusion reactor materials, the interface particle behavior in plasma-wall interaction, the behavior of tritium on the surface of breeding materials, and breeding materials and the behavior of tritium in plasma-wall interaction. (K.I.)

  12. Investigation of aqueous slurries as fusion reactor blankets

    International Nuclear Information System (INIS)

    Schuller, M.J.

    1985-01-01

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

  13. 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

  14. Corrosion characteristics of an aqueous self-cooled fusion blanket

    International Nuclear Information System (INIS)

    Bogaerts, W.F.; Embrechts, M.J.; Steiner, D.; Deutsch, L.; Jackson, D.

    1986-01-01

    A novel aqueous self-cooled blanket concept (ASCB) has recently been proposed. This blanket concept, as applied to a MARS-like tandem mirror reactor, consists of disks of spiraling tubes of Zircaloy-4 housed in a structural container of vanadium alloy (V-15 Ti-5 Cr). The Zircaloy tubes are cooled by a mixture of light and heavy water with 9 g of LiOH per 100 cm 3 of water dissolved in the coolant. A major issue for the feasibility of the integrated blanket coil concept is the chemical compatibility of the coolant and Zircaloy. Initial corrosion tests have been undertaken in order to resolve this question. Results clearly show that successful alloy heats can be prepared, for which corrosion problems will probably not be the limiting factor of the ASCB design concept. As is quite well known from fission engineering studies, small variations in the alloy compositions or in the metallurgical structure may, however, be able to cause significant alterations in the oxidation or corrosion rates. Further tests will be necessary to resolve the remaining uncertainties and to determine the behavior of successful alloy heats in the presence of trace impurities in order to address the sensitivity to localized corrosion phenomena such as pitting, stress corrosion cracking, and intergranular attack

  15. Future with fusion power

    International Nuclear Information System (INIS)

    Hirschfeld, F.

    1977-01-01

    This article reviews several current approaches to the development of nuclear fusion power sources by the year 2000. First mentioned is the only project to develop a nonpolluting, radiation-free source by using only natural and nonradioactive isotopes (nuclei of deuterium, helium 3 and boron) as ''advanced'' fuels. This system will also be capable of direct conversion of the released energy into electricity. Next described is the PACER concept, in which thermonuclear burning of deuterium occurs in fusion explosion taking place underground (e.g., in a salt dome). The released energy is absorbed in high-pressure steam which is then piped to a surface heat exchanger to provide steam for a turbogenerator. After filtration, the steam is returned. The PACER system also produces fissionable fuel. The balance of the article reviews three ''magnetic fusion'' approaches. Tokamak, mirror and theta pinch systems utilize magnetic fields to confine a plasma for either pulsed or steady-state operation. The tokamak and theta pinch are toroidal in shape, while the mirror can be thought of as a magnetic field configuration of roughly tubular shape that confines the plasma by means of higher fields at the ends than at its center. The tokamak approach accounts for about 65 percent of the magnetic fusion research and development, while theta pinches and mirrors represent about 15 percent each. Refs

  16. 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

  17. Neutronics analysis of water-cooled energy production blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Jiang Jieqiong; Wang Minghuang; Chen Zhong; Qiu Yuefeng; Liu Jinchao; Bai Yunqing; Chen Hongli; Hu Yanglin

    2010-01-01

    Neutronics calculations were performed to analyse the parameters of blanket energy multiplication factor (M) and tritium breeding ratio (TBR) in a fusion-fission hybrid reactor for energy production named FDS (Fusion-Driven hybrid System)-EM (Energy Multiplier) blanket. The most significant and main goal of the FDS-EM blanket is to achieve the energy gain of about 1 GWe with self-sustaining tritium, i.e. the M factor is expected to be ∼90. Four different fission materials were taken into account to evaluate M in subcritical blanket: (i) depleted uranium, (ii) natural uranium, (iii) enriched uranium, and (iv) Nuclear Waste (transuranic from 33 000 MWD/MTU PWR (Pressurized Water Reactor) and depleted uranium) oxide. These calculations and analyses were performed using nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library) and a home-developed code VisualBUS. The results showed that the performance of the blanket loaded with Nuclear Waste was most attractive and it could be promising to effectively obtain tritium self-sufficiency and a high-energy multiplication.

  18. Phase-IIC experiments of the JAERI/USDOE collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1992-12-01

    Neutronics experiments on two types of heterogeneous blankets have been performed as the Phase-IIC experiment of JAERI/USDOE collaborative program on fusion blanket neutronics. The experimental system was used in the same geometry as the previous Phase-IIA series which was a closed geometry using neutron source enclosure of lithium carbonate. The heterogeneous blankets selected here are the beryllium edge-on and the water coolant channel assemblies. In the former the beryllium and lithium-oxide layers are piled up alternately in the front part of test blanket. In the latter, the three simulated water cooling channels are settled in the Li 2 O blanket. These are producing steep gradient of neutron flux around material boundary. The calculation accuracy and measurement method for these features is a key of interest in the experiments. The measurements were performed for tritium production rate and the other nuclear parameters as well as the previous experiments. This report describes the experimental detail and the results enough to use for the benchmark data for testing the data and method of design calculation of fusion reactors. (author)

  19. Fusion blankets for catalyzed D--D and D--He3 reactors

    International Nuclear Information System (INIS)

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

    1977-01-01

    Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant

  20. Fusion blankets for catalyzed D--D and D--3He reactors

    International Nuclear Information System (INIS)

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

    1977-01-01

    Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant

  1. Development of vanadium base alloys for fusion first-wall/blanket applications

    International Nuclear Information System (INIS)

    Smith, D.L.; Chung, H.M.; Loomis, B.A.; Matsui, H.; Votinov, S.; VanWitzenburg, W.

    1994-01-01

    Vanadium alloys have been identified as a leading candidate material for fusion first-wall/blanket applications. Certain vanadium alloys exhibit favorable safety and environmental characteristics, good fabricability, high temperature and heat load capability, good compatibility with liquid metals and resistance to irradiation damage effects. The current focus is on vanadium alloys with (3-5)% Cr and (3-5)% Ti with a V-4Cr-4Ti alloy as the leading candidate. Preliminary results indicate that the crack-growth rates of certain alloys are not highly sensitive to irradiation. Results from the Dynamic Helium Charging Experiment (DHCE) which simulates fusion relevant helium/dpa ratios are similar to results from neutron irradiated material. This paper presents an overview of the recent results on the development of vanadium alloys for fusion first wall/blanket applications

  2. Material Issues of Blanket Systems for Fusion Reactors - Compatibility with Cooling Water -

    Science.gov (United States)

    Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

    Environmental assisted cracking (EAC) is one of the material issues for the reactor core components of light water power reactors(LWRs). Much experience and knowledge have been obtained about the EAC in the LWR field. They will be useful to prevent the EAC of water-cooled blanket systems of fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC in a water-cooled blanket does not seem to be acritical issue. However, some uncertainties about influences on water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations elucidating the uncertainties are discussed.

  3. Thermal stresses and cyclic creep-fatigue in fusion reactor blanket

    International Nuclear Information System (INIS)

    Liu, K.C.

    1977-01-01

    Thermal stresses in the first walls of fusion reactor blankets were studied in detail. ORNL multibucket modules are emphasized. Practicality of using the bucket module rather than other blanket designs is examined. The analysis shows that applying intelligent engineering judgment in design can reduce the thermal stresses significantly. Arrangement of coolant flow and distribution of temperature are reviewed. Creep-fatigue property requirements for a first wall are discussed on the basis of existing design rules and criteria. Some major questions are pointed out and experiments needed to resolve basic uncertainties relative to key design decisions are discussed

  4. The effect of optimal wall loads and blanket technologies on the cost of fusion electricity

    International Nuclear Information System (INIS)

    Knight, P.J.; Ward, D.J.

    2000-01-01

    This paper presents a discussion of trends in fusion economics based on technology, as well as, physics arguments. Based on relatively simple physics considerations, supported by detailed systems code calculations, it is shown that optimal wall loads are not high. The results of systems code calculations, focussing on the economic impact of different blanket technologies, are described. These suggest that the economically favourable thermodynamic efficiencies of some blankets capable of operating at higher temperatures may be counterbalanced by the economic penalties of shorter lifetimes

  5. One- and two-dimensional heating analyses of fusion synfuel blankets

    International Nuclear Information System (INIS)

    Tsang, J.S.K.; Lazareth, O.W.; Powell, J.R.

    1979-01-01

    Comparisons between one- and two-dimensional neutronics and heating analyses were performed on a Brookhaven designed fusion reactor blanket featuring synthetic fuel production. In this two temperature region blanket design, the structural shell is stainless steel. The interior of the module is a packed ball of high temperature ceramic material. The low temperature shell and the high temperature ceramic interior are separately cooled. Process steam (approx. 1500 0 C) is then produced in the ceramic core for the producion of H 2 and H 2 -based synthetic fuels by a high temperature electrolysis (HTE) process

  6. Rotating liquid blanket for a toroidal fusion reator

    International Nuclear Information System (INIS)

    Moir, R.W.

    1987-01-01

    A novel blanket concept is presented for toroidal geometry in which many of the limitations imposed by a first wall are avoided by not having a first wall in the usual sense. The blanket consists of a rapidly rotating, low-vapor-pressure liquid that has a sharp boundary with the vacuum region. Nozzles inject ja continuous layer of cool liquid on the inner surface. The noncentricity of the plasma is maintained so that the plasma scrape-off region intersects the rotating liqid in a localized region. This noncentricity allows sufficient space so that the scrape-off plasma layer will not bombard the nozzles, whch penetrate through the rotating liquid. This liquid ''first wall'' is bombarded by the plasma, resulting in heat deposition, sputtering, and evaporation during the short time before the exposed liquid is covered by fresh, cool liquid from the nozzles. The advantages of this reactor concept appear to be very high wall loadings (speculated to be over 10 MW/m 2 ) and long component lifetime, both crucial economic factors. The nozzles are designed for easy replacement. The reactor's disatvantage is its enormous potential for plasma contamination by impurities. (orig.)

  7. The future of fusion

    International Nuclear Information System (INIS)

    Sheffield, John

    2001-01-01

    The population of the world is increasing, mainly in the developing world, and is projected to saturate within about 100 years at up to twice the present population of 6 billion people (Bos et al., World Population Projections: 1994-95 Edition, Published for the World Bank). Studies (Goldemberg and Johansson (Eds.), Energy as an Instrument for Socio-Economic Development United Nations Development Programme, New York, 1995, p. 9; United Nations Statistical Yearbooks, 10th issue; 1965; 20th issue; 1975, 22nd issue, 1977; 32nd issue, 1987; and 39th issue 1994, United Nations Publications; Sheffield, J. Technol. Forecasting Social Change 59 (1998) 55.) show that, historically, the population growth rate has varied inversely as the annual per capita energy use in most parts of the developing world, where per capita energy use is typically less than 1 t of oil equivalent energy per year. However, in areas with more than 2-3 t of oil equivalent of energy use per year per person, the growth rate is around zero. If this trend continues, a stable world population will require, allowing for energy efficiency improvements, some 2-3 times the present annual energy use. There is an abundance of energy in the world both exploited and potential to meet this need - fossil, fission, and renewables - but it is not evenly distributed, some are costly, and there are issues of environmental pollution in present use, that may limit use. Fusion energy is a potential longer-term source with attractive environmental features. It is the least-developed energy option and still faces a challenging development path, but there are many areas of the world that would benefit hugely from its deployment from the later part of the 21st century onward, and it is important to consider how it might be deployed. Most fusion power plant options considered today show an economy of scale, owing to the fixed distance needed for shielding fusion neutrons, tritium breeding and handling the heat loads. One

  8. 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.)

  9. 233U breeding and neutron multiplying blankets for fusion reactors

    International Nuclear Information System (INIS)

    Cook, A.G.; Maniscalco, J.A.

    1975-01-01

    In this work, along with a previous paper three possible uses of 14-MeV deuterium--tritium fusion neutrons are investigated: energy production, neutron multiplication, and fissile-fuel breeding. The results presented include neutronic studies of fissioning and nonfissioning thorium systems, tritium breeding systems, various fuel options (UO 2 , UC, UC 2 , etc.), and uranium as well as refractory metal first-wall neutron-multiplying regions. A brief energy balance and an estimate of potential revenues for fusion devices are given to help illustrate the potentials of these designs

  10. A passively-safe fusion reactor blanket with helium coolant and steel structure

    Energy Technology Data Exchange (ETDEWEB)

    Crosswait, Kenneth Mitchell [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1994-04-01

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.

  11. 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.

  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.

    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

  13. First wall and blanket stresses induced by cyclic fusion core operations

    International Nuclear Information System (INIS)

    Bohachevsky, I.O.; Kostoff, R.N.

    1981-01-01

    An analysis is made of cyclic thermal loads and stresses for the complete range of operating conditions. Two critical components were examined; the solid wall adjacent to the fusion plasma (first wall) and the fuel elements in the high power density region of the blanket. Simple closed form expressions were derived for temperature increases and thermal stresses that may be evaluated conveniently and rapidly and the values compared for different systems

  14. Heat transfer models for fusion blanket first walls

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1977-01-01

    In the development of magnetically confined fusion reactors, the ability to cool the first wall, i.e., the first material surface interfacing the plasma, appears to be a critical factor involved in establishing the wall load limit. In order to understand the thermal behavior of the first wall time-dependent, one-dimensional heat conduction models are reviewed with differing modes of heat extraction and cooling

  15. Investigation of lanthanides as neutron multipliers for hybrid and fusion reactor blankets

    International Nuclear Information System (INIS)

    Sahin, Sumer

    1982-01-01

    The neutronic performance of three lanthanides ( 149 Sm, europium, and gadolinium) as neutron multiplier for the blanket of a fusion-fission (hybrid) and a pure fusion reactor has been evaluated and compared with that of beryllium and lead. During the calculations, the fission zone is made up of UO 2 rods from the LOTUS experimental hybrid facility now under construction at the Nuclear Engineering Laboratory of the Swiss Federal Institute of Technology in Lausanne. In fusion blanket the fuel zone is replaced by pure lithium. The calculations were performed for two different boundary conditions for the left boundary: (a) reflecting, representative of a typical confinement geometry, and (b) vacuum, which represents a typical blanket experiment in plane geometry. For a vacuum left boundary, threshold reactions are reduced by a factor of about 2 while 1/v-type reactions are decreased by a factor of between 5 and 10, as a consequence of the softer spectrum produced by a reflecting left boundary. In general, the results, notably tritium breeding and energy multiplication, are comparable for the lanthanide multipliers and for beryllium and lead if the left boundary is a vacuum. The use of 149 Sm is slightly less effective than europium or gadolinium and all of the lanthanides perform better for a vacuum left boundary than for the reflecting case. The analyses presented here also illustrate the importance of potential spectral shifts that can occur as the result of experimental exigencies

  16. Extension of the AUS reactor neutronics system for application to fusion blanket neutronics

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1984-03-01

    The AUS modular code scheme for reactor neutronics computations has been extended to apply to fusion blanket neutronics. A new group cross-section library with 200 neutron groups, 37 photon groups and kerma factor data has been generated from ENDF/B-IV. The library includes neutron resonance subgroup parameters and temperature-dependent data for thermal neutron scattering matrices. The validity of the overall calculation system for fusion applications has been checked by comparison with a number of published conceptual design studies

  17. 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.)

  18. Blanket materials for fusion reactors: comparisons of thermochemical performance

    International Nuclear Information System (INIS)

    Johnson, C.E.; Fischer, A.K.; Tetenbaum, M.

    1984-01-01

    Thermodynamic calculations have been made to predict the thermochemical performance of the fusion reactor breeder materials, Li 2 O, LiAlO 2 , and Li 4 SiO 4 in the temperature range 900 to 1300 0 K and in the oxygen activity range 10 -25 to 10 -5 . Except for a portion of these ranges, the performance of LiAlO 2 is predicted to be better than that of Li 2 O and Li 4 SiO 4 . The protium purge technique for enhancing tritium release is explored for the Li 2 O system; it appears advantageous at higher temperatures but should be used cautiously at lower temperatures. Oxygen activity is an important variable in these systems and must be considered in executing and interpreting measurements on rates of tritium release, the form of released tritium, diffusion of tritiated species and their identities, retention of tritium in the condensed phase, and solubility of hydrogen isotope gases

  19. 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

  20. Effect of electromagnetic coupling on MHD flow in the manifold of fusion liquid metal blanket

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Meng, Zi; Feng, Jingchao; He, Qingyun

    2014-10-15

    In fusion liquid metal (LM) blanket, magnetohydrodynamics (MHD) effects will dominate the flow patterns and the heat transfer characteristics of the liquid metal flow. Manifold is a key component in LM blanket in charge of distributing or collecting the liquid metal coolant. In this region, the complex three dimensional MHD phenomena will be occurred, and the velocity, pressure and flow rate distributions may be dramatically influenced. One important aspect is the electromagnetic coupling effect resulting from an exchange of electric currents between two neighboring fluid domains that can lead to modifications of flow distribution and pressure drop compared to that in electrical separated channels. Understanding the electromagnetic coupling effect in manifold is necessary to optimize the liquid metal blanket design. In this work, a numerical study was carried out to investigate the effect of electromagnetic coupling on MHD flow in a manifold region. The typical manifold geometry in LM blanket was considered, a rectangular supply duct entering a rectangular expansion area, finally feeding into 3 rectangular parallel channels. This paper investigated the effect of electromagnetic coupling on MHD flow in a manifold region. Different electromagnetic coupling modes with different combinations of electrical conductivity of walls were studied numerically. The flow distribution and pressure drop of these modes have been evaluated.

  1. 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

  2. Application of the integrated blanket-coil concept (IBC) to fusion reactors

    International Nuclear Information System (INIS)

    Embrechts, M.J.; Steiner, D.; Mohanti, R.; Duggan, W.

    1987-01-01

    A novel concept is proposed for combining the blanket and coil functions of a fusion reactor into a single component and several unique applications to fusion reactor embodiments are identified. The proposed concept takes advantage of the fact that lithium is a good electrical conductor in addition to being a unique tritium-breeding material capable of energy recovery and transport at high temperatures. This concept, designated the ''integrated-blanket-coil (IBC) concept'' has the potential for: allowing fusion reactor embodiments which are easier to maintain; making fusion reactors more compact with an intrinsic ultra-high mass power density (net kW/sub E//metric tonne); and enhancing the tritium breeding potential for special coil applications such as ohmic heating and bean identation. By assuming a sandwich construction for the IBC walls (i.e., a layered combination of a thin wall of structural material, insulator and structural materials) the magnetohydrodynamic (MHD)-induced pressure drops and associated pressure stresses are modest and well below design limits. Possible unique applications of the IBC concept have been investigated and include the IBC concept applied to the poloidal field (PF) coils, toroidal field (TF) coils, divertor coils, ohmic heating (OH) coils, and identation coils for bean shaping

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

    Energy Technology Data Exchange (ETDEWEB)

    Fan, L.

    2015-07-01

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

  4. 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)

  5. 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.)

  6. 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

  7. Transmutation and activation of stainless steel 316 SS in a thermal fusion reactor blanket

    International Nuclear Information System (INIS)

    Gruber, J.; Schneider, J.

    1977-10-01

    Using the program MATEXP (matrix exponential method) the influence of neutron flux is calculated for stainless steel 3s16 SS which is used as a structural material in a fusion reactor blanket (CTRD-I). The transmutations, activations and γ-dose rates are determined for an operation time of 20 years. Investigating the decay behaviour after operation time, we found that the long term activity and dose rate was mainly influenced by five nuclides: Fe55, Ni63, Ni59, Co60 and Nb94. (orig.) [de

  8. Approximated neutronic calculation for the tritium breeding ratio in fusion reactor blankets

    International Nuclear Information System (INIS)

    Santos, Raul dos

    1983-01-01

    An approximated model for the calculation of the tritium breeding ratio in conceptual thermonuclear fusion reactor blankets is presented. This model makes use of the exponential absorption concept due to the Li 6 (n, He 4 )T and Li 7 (n, n'He 4 )T reactions. The results of this approximated method are compared with reference benchmarks which were generated by the nuclear codes ANISN (discrete ordinates) and MORSE (Monte Carlo method). The maximum deviation among the results have been around 10%. (Author) [pt

  9. INDRA: a program system for calculating the neutronics and photonics characteristics of a fusion reactor blanket

    International Nuclear Information System (INIS)

    Perry, R.T.; Gorenflo, H.; Daenner, W.

    1976-01-01

    INDRA is a program system for calculating the neutronics and photonics characteristics of fusion reactor blankets. It incorporates a total of 19 different codes and 5 large data libraries. 10 of the codes are available from the code distribution organizations. Some of them, however, have been slightly modified in order to permit a convenient transfer of information from one program module to the next. The remaining 9 programs have been prepared by the authors to complete the system with respect to flexibility and to facilitate the handling of the results. (orig./WBU) [de

  10. Fusion materials: Technical evaluation of the technology of vandium alloys for use as blanket structural materials in fusion power systems

    International Nuclear Information System (INIS)

    1993-01-01

    The Committee's evaluation of vanadium alloys as a structural material for fusion reactors was constrained by limited data and time. The design of the International Thermonuclear Experimental Reactor is still in the concept stage, so meaningful design requirements were not available. The data on the effect of environment and irradiation on vanadium alloys were sparse, and interpolation of these data were made to select the V-5Cr-5Ti alloy. With an aggressive, fully funded program it is possible to qualify a vanadium alloy as the principal structural material for the ITER blanket in the available 5 to 8-year window. However, the data base for V-5Cr-5Ti is United and will require an extensive development and test program. Because of the chemical reactivity of vanadium the alloy will be less tolerant of system failures, accidents, and off-normal events than most other candidate blanket structural materials and will require more careful handling during fabrication of hardware. Because of the cost of the material more stringent requirements on processes, and minimal historical worlding experience, it will cost an order of magnitude to qualify a vanadium alloy for ITER blanket structures than other candidate materials. The use of vanadium is difficult and uncertain; therefore, other options should be explored more thoroughly before a final selection of vanadium is confirmed. The Committee views the risk as being too high to rely solely on vanadium alloys. In viewing the state and nature of the design of the ITER blanket as presented to the Committee, h is obvious that there is a need to move toward integrating fabrication, welding, and materials engineers into the ITER design team. If the vanadium allay option is to be pursued, a large program needs to be started immediately. The commitment of funding and other resources needs to be firm and consistent with a realistic program plan

  11. 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.

  12. Modeling of liquid-metal corrosion/deposition in a fusion reactor blanket

    International Nuclear Information System (INIS)

    Malang, S.; Smith, D.L.

    1984-04-01

    A model has been developed for the investigation of the liquid-metal corrosion and the corrosion product transport in a liquid-metal-cooled fusion reactor blanket. The model describes the two-dimensional transport of wall material in the liquid-metal flow and is based on the following assumptions: (1) parallel flow in a straight circular tube; (2) transport of wall material perpendicular to the flow direction by diffusion and turbulent exchange; in flow direction by the flow motion only; (3) magnetic field causes uniform velocity profile with thin boundary layer and suppresses turbulent mass exchange; and (4) liquid metal at the interface is saturated with wall material. A computer code based on this model has been used to analyze the corrosion of ferritic steel by lithium lead and the deposition of wall material in the cooler part of a loop. Three cases have been investigated: (1) ANL forced convection corrosion experiment (without magnetic field); (2) corrosion in the MARS liquid-metal-cooled blanket (with magnetic field); and (3) deposition of wall material in the corrosion product cleanup system of the MARS blanket loop

  13. 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)

  14. 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.

  15. Comparison of the leading candidate combinations of blanket materials, thermodynamic cycles, and tritium systems for full scale fusion power plants

    International Nuclear Information System (INIS)

    Fraas, A.P.

    1975-01-01

    The many possible combinations of blanket materials, tritium generation and recovery systems, and power conversion systems were surveyed and a comprehensive set of designs were generated by using a common set of ground rules that include all of the boundary conditions that could be envisioned for a full-scale commercial fusion power plant. Particular attention was given to the effects of blanket temperature on power plant cycle efficiency and economics, the interdependence of the thermodynamic cycle and the tritium recovery system, and to thermal and pressure stresses in the blanket structure. The results indicate that, of the wide variety of systems that have been considered, the most promising employs lithium recirculated in a closed loop within a niobium blanket structure and cooled with boiling potassium or cesium. This approach gives the simplest and lowest cost tritium recovery system, the lowest pressure and thermal stresses, the simplest structure with the lowest probability of a leak, the greatest resistance to damage from a plasma energy dump, and the lowest rate of plasma contamination by either outgassing or sputtering. The only other blanket materials combination that appears fairly likely to give a satisfactory tritium generation and recovery system is a lithium-beryllium fluoride-Incoloy blanket, and even this system involves major uncertainties in the effectiveness, size, and cost of the tritium recovery system. Further, the Li 2 BeF 4 blanket system has the disadvantage that the world reserves of beryllium are too limited to support a full-blown fusion reactor economy, its poor thermal conductivity leads to cooling difficulties and a requirement for a complex structure with intricate cooling passages, and this inherently leads to an expansive blanket with a relatively high probability of leaks. The other blanket materials combinations yield even less attractive systems

  16. Evaluation of alternative methods of simulating asymmetric bulk heating in fusion reactor blanket/shield components

    International Nuclear Information System (INIS)

    Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Wadkins, R.P.; Wessol, D.E.

    1981-10-01

    As a part of Phase O, Test Program Element-II of the Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program, a study was conducted by EG and G Idaho, Inc., to identify, characterize, and recommend alternative approaches for simulating fusion bulk heating in blanket/shield components. This is the report on that effort. Since the usefulness of any simulation approach depends upon the particular experiment considered, classes of problem types (thermal-hydraulic, thermomechanical, etc.) and material types (structure, solid breeder, etc.) are developed. The evaluation of the various simulation approaches is performed for the various significant combinations of problem class and material class. The simulation approaches considered are discrete-source heating, direct resistance, electromagnetic induction, microwave heating, and nuclear heating. From the evaluations performed for each experiment type, discrete - source heating emerges as a good approach for bulk heating simulation in thermal - hydraulics experiments, and nuclear heating appears to be a good approach in experiments addressing thermomechanics and combined thermal-hydraulic/thermomechanics

  17. 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

  18. Sensisivity and Uncertainty analysis for the Tritium Breeding Ratio of a DEMO Fusion reactor with a Helium cooled pebble bed blanket

    OpenAIRE

    Nunnenmann, Elena; Fischer, Ulrich; Stieglitz, Robert

    2016-01-01

    An uncertainty analysis was performed for the tritium breeding ratio (TBR) of a fusion power plant of the European DEMO type using the MCSEN patch to the MCNP Monte Carlo code. The breeding blanket was of the type Helium Cooled Pebble Bed (HCPB), currently under development in the European Power Plant Physics and Technology (PPPT) programme for a fusion power demonstration reactor (DEMO). A suitable 3D model of the DEMO reactor with HCPB blanket modules, as routinely used for blanket design c...

  19. Basic principles of lead and lead-bismuth eutectic application in blanket of fusion reactors

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Pinaev, S.S.; Muraviev, E.V.; Romanov, P.V.

    2005-01-01

    High magnetohydrodynamic pressure drop is an important issue for liquid metal blanket concepts. To decrease magnetohydrodynamic resistance authors propose to form insulating coatings on internal surface of blanket ducts at any moment of fusion reactor exploitation. It may be achieved easily if lead or lead-bismuth eutectic is used and technology of oxidative potential handling is applied. A number of experiments carried out in NNSTU show the availability of the proposed technology. It bases on formation of the insulating coatings that consist of the oxides of components of the structural materials and of the coolant components. In-situ value of the insulating coatings characteristics ρδ is ∼ 10 -5 Ohm·m 2 for steels and 5,0x10 -6 - 5,0x10 -5 Ohm·m 2 for vanadium alloys. Thermal cycling is possible during exploitation of a blanket. The experimental research of the insulating coatings properties during thermal cycling have shown that the coatings formed into the lead and lead-bismuth coolants save there insulating properties. Experience of many years is an undoubted advantage of the lead-bismuth coolant and less of the lead coolant in comparison with lithium. Russian Federation possesses of experience of exploitation of the research and industrial facilities, of experience of creation of the pumps, steamgenerators and equipment with heavy liquid metal coolants. The unique experience of designing, assembling and exploitation of the fission reactors with lead-bismuth coolant is also available. The problem of technology of lead and lead-bismuth coolants for power high temperature radioactive facilities has been solved. Accidents, emergency situations such as leakage of steamgenerators or depressurization of gas system in facilities with lead and lead-bismuth coolants have been explored and suppressed. (author)

  20. Phase IIC experiments of the USDOE/JAERI collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Youssef, M.Z.

    1992-12-01

    Effort in Phase IIC of the US/JAERI Collaborative Program on Fusion Neutronics was focused on performing integral experiments and post analyses on blankets that include the actual heterogeneities found in several blanket designs. Two geometrical arrangements were considered for the blanket assembly, namely multi-layers of Li 2 O and beryllium in an edge-on, horizontally alternating configuration for a front depth of 30 cm, followed by the Li 2 O breeding zone (Be edge-on, BEO, experiment), and vertical water coolant channels arrangement (WCC experiment). The objectives are to examine the accuracy of predicting tritium production. In the BEO system, it was shown that, with the zonal method to measure tritium production from natural lithium (Tn), the calculated-to-measured values (C/E) are 0.95-1.05 (JAERI) and 0.98-0.9 (U.S.), which is consistent with the results obtained in other Phases of the Program (Phases IIA and IIb)). In the WCC experiment, there is a noticeable change in C/E values for T 6 near the coolant channels where steep gradients in T 6 production are observed. The C/E values obtained with the Li-foil detectors are on the average closer to unity than those obtained by the Li-glass method. As for T 7 , the values obtained by NE213 method are within ±15% in JAERI's calculations, but larger values (∼20-25%) are obtained in the U.S. calculations due to the differences of cross-sections data files. Around heterogeneities, the prediction accuracy for T 7 is better than for T 6 . (J.P.N.)

  1. Catalyzed deuterium-deuterium and deuterium-tritium fusion blankets for high temperature process heat production

    International Nuclear Information System (INIS)

    Ragheb, M.M.H.; Salimi, B.

    1982-01-01

    Tritiumless blanket designs, associated with a catalyzed deuterium-deuterium (D-D) fusion cycle and using a single high temperature solid pebble or falling bed zone, for process heat production, are proposed. Neutronics and photonics calculations, using the Monte Carlo method, show that an about 90% heat deposition fraction is possible in the high temperature zone, compared to a 30 to 40% fraction if a deuterium-tritium (D-T) fusion cycle is used with separate breeding and heat deposition zones. Such a design is intended primarily for synthetic fuels manufacture through hydrogen production using high temperature water electrolysis. A system analysis involving plant energy balances and accounting for the different fusion energy partitions into neutrons and charged particles showed that plasma amplification factors in the range of 2 are needed. In terms of maximization of process heat and electricity production, and the maximization of the ratio of high temperature process heat to electricity, the catalyzed D-D system outperforms the D-T one by about 20%. The concept is thought competitive to the lithium boiler concept for such applications, with the added potential advantages of lower tritium inventories in the plasma, reduced lithium pumping (in the case of magnetic confinement) and safety problems, less radiation damage at the first wall, and minimized risks of radioactive product contamination by tritium

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

    International Nuclear Information System (INIS)

    Deng Baiquan; Huang Jinhua; Peng Lilin; Yan Jiancheng

    2003-01-01

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

  3. Al-based anti-corrosion and T-permeation barrier development for future DEMO blankets

    International Nuclear Information System (INIS)

    Krauss, W.; Konys, J.; Holstein, N.; Zimmermann, H.

    2011-01-01

    In the Helium-Cooled-Liquid-Lead (HCLL) design of Test-Blanket-Modules (TBM's) for a future fusion power plant Pb-15.7Li is used as liquid breeder which is in direct contact with the structure material, e.g. EUROFER steel. Compatibility testing showed that high corrosion attack appears and that the dissolved steel components form precipitates with a high risk of system blockages. A reliable operation needs coatings as corrosion barriers. The earlier developed Hot-Dip Aluminisation (HDA) process has shown that Al-based scales can act as anti-corrosion as well as T-permeation barriers. Meanwhile two advanced electro-chemically based processes for deposition of Al-scales were successfully developed. The first (ECA = Electro-Chemical Al-deposition) is working with an organic electrolyte and the second one (ECX = Electro-Chemical-X-metal-deposition) is based on ionic liquids. Coatings in the μm-range were deposited homogeneously with exact controllable thicknesses. Metallurgical investigations showed the successful generation of protective scales and compatibility testing demonstrated the barrier function.

  4. Al-based anti-corrosion and T-permeation barrier development for future DEMO blankets

    Energy Technology Data Exchange (ETDEWEB)

    Krauss, W., E-mail: wolfgang.krauss@kit.edu [Karlsruhe Institute of Technology, Hermann von Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Konys, J.; Holstein, N.; Zimmermann, H. [Karlsruhe Institute of Technology, Hermann von Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2011-10-01

    In the Helium-Cooled-Liquid-Lead (HCLL) design of Test-Blanket-Modules (TBM's) for a future fusion power plant Pb-15.7Li is used as liquid breeder which is in direct contact with the structure material, e.g. EUROFER steel. Compatibility testing showed that high corrosion attack appears and that the dissolved steel components form precipitates with a high risk of system blockages. A reliable operation needs coatings as corrosion barriers. The earlier developed Hot-Dip Aluminisation (HDA) process has shown that Al-based scales can act as anti-corrosion as well as T-permeation barriers. Meanwhile two advanced electro-chemically based processes for deposition of Al-scales were successfully developed. The first (ECA = Electro-Chemical Al-deposition) is working with an organic electrolyte and the second one (ECX = Electro-Chemical-X-metal-deposition) is based on ionic liquids. Coatings in the {mu}m-range were deposited homogeneously with exact controllable thicknesses. Metallurgical investigations showed the successful generation of protective scales and compatibility testing demonstrated the barrier function.

  5. A vanadium alloy for the application in a liquid metal blanket of a fusion reactor

    Science.gov (United States)

    Borgstedt, H. U.; Grundmann, M.; Konys, J.; Perić, Z.

    1988-07-01

    The vanadium alloy V3Ti1Si has been corrosion tested in liquid lithium and the eutectic alloy Pb-17Li at 550°C. This alloy has a comparable corrosion resistance to the alloy V15Cr5Ti in lithium. In this molten metal it is superior to stainless steel AISI 316. In the Pb-17Li melt it is even superior to martensitic steels. The alloy has only a weak tendency to be dissolved. It is sensitive to an exchange of non-metallic elements, which causes the formation of a hardened surface layer. These chemical effects are influenced by the mass and surface ratios of the vanadium alloy to the molten metals and other structural materials. These ratios are unfavorable in the two test loops. The effects might be less pronounced in a vanadium alloy/liquid metal fusion reactor blanket.

  6. Heat transfer in the lithium-cooled blanket of a pulsed fusion reactor

    International Nuclear Information System (INIS)

    Cort, G.E.; Krakowski, R.A.

    1978-01-01

    The transient temperature distribution in the lithium-cooled blanket of a pulsed fusion reactor has been calculated using a finite-element heat-conduction computer program. An auxiliary program was used to predict the coolant transient velocity in a network of parallel and series flow passages with constant driving pressure and varying magnetic field. The coolant velocity was calculated by a Runge-Kutta numerical integration of the conservation equations. The lithium coolant was part of the finite-element heat-conduction mesh with the velocity terms included in the total matrix. The matrix was solved implicitly at each time step for the nodal point temperatures. Slug flow was assumed in the coolant passages and the Boussinesq analogy was used to calculate turbulent heat transfer when the magnetic field was not present

  7. Tritium breeding experiments in a fusion blanket assembly using a low-intensity neutron generator

    International Nuclear Information System (INIS)

    Dalton, A.W.; Woodley, H.J.; McGregor, B.J.

    1987-01-01

    Experiments have been carried out to determine the accuracy with which tritium production rates (TPRs) can be measured in a fusion blanket assembly of non-spherical geometry by a non-central low intensity D-T neutron source (2x10 10 neutrons per second). The tritium production was determined for samples of lithium carbonate containing high enrichments of 6 Li(96%) and 7 Li(99.9%). The measured data were used to check the accuracy with which the TPRs could be numerically predicted using current nuclear data and calculational methods. The numerical predictions from tritium production from the 7 Li samples agreed within the experimental errors of the measurements, but 6 Li measurements which differ by more than 20 per cent from the predicted values were observed in the lower half of the assembly

  8. 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)

  9. Feasibility study of LiF-BeF2 and chloride salts as blanket coolants for fusion power reactors

    International Nuclear Information System (INIS)

    Imamura, Y.

    1977-09-01

    The feasibility of using molten salts, in particular, nonberyllium-bearing chloride salts, as blanket coolants for Tokamak fusion reactors has been examined for the nucleonic and thermal/hydraulic aspects. It is concluded that the chloride salts, i.e., LiCl--KCl, LiCl--PbCl 2 and LiCl--SnCl 2 , can be used as the blanket coolant for a static lithium metal blanket provided that large blanket thickness can be tolerated, along with the use of U-238 for neutron multiplication in the cases of LiCl--KCl or LiCl--SnCl 2 cooled blankets. However, to make the appraisal complete, the tritium recovery and corrosion problems must be examined extensively, based on data not yet at hand. As for LiF--BeF 2 , it is observed that although the salt mixture can be used for a single fluid blanket with satisfactory nuclear performance, careful attention should be paid to the cooling capability

  10. Blanket safety by GEMSAFE methodology

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Saito, Masaki

    2001-01-01

    General Methodology of Safety Analysis and Evaluation for Fusion Energy Systems (GEMSAFE) has been applied to a number of fusion system designs, such as R-tokamak, Fusion Experimental Reactor (FER), and the International Thermonuclear Experimental Reactor (ITER) designs in the both stages of Conceptual Design Activities (CDA) and Engineering Design Activities (EDA). Though the major objective of GEMSAFE is to reasonably select design basis events (DBEs) it is also useful to elucidate related safety functions as well as requirements to ensure its safety. In this paper, we apply the methodology to fusion systems with future tritium breeding blankets and make clear which points of the system should be of concern from safety ensuring point of view. In this context, we have obtained five DBEs that are related to the blanket system. We have also clarified the safety functions required to prevent accident propagations initiated by those blanket-specific DBEs. The outline of the methodology is also reviewed. (author)

  11. 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

  12. 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.)

  13. Future directions in fusion research

    International Nuclear Information System (INIS)

    Clarke, J.F.

    1987-01-01

    The author discusses his analysis to quantify the priority of fusion R and D in the United States. The conclusion is that this priority has been essentially constant for 35 years with only two exceptions. He identifies four basic problems that must be solved. These problems are: to improve the scientific understanding of confinement concepts if we are going to have an energy source that can be utilized some day; to understand the physics of burning plasmas; to develop the materials for fusion use to realize the environmental potential of fusion; and to develop fusion nuclear technology. A response to these problems is given, based on the author's argument for international collaboration in fusion research

  14. Fabrication and performance of AIN insulator coatings for application in fusion reactor blankets

    International Nuclear Information System (INIS)

    Natesan, K.

    1995-09-01

    The liquid-metal blanket concept for fusion reactors requires an coating on the first-wall structural material to minimize the magnetohydrodynamic pressure drop that occurs during the flow of liquid metal in a magnetic field. Based on the thermodynamics of interactions betwen the coating and the liquid lithium on one side and the structural V-base alloy on the other side, an AIN coating was selected as a candidate. Detailed investigations were conducted on the fabrication, metallurgical microstructure, compatibility in liquid Li, and electrical characteristics of AIN material obtained from several sources. Lithium compatibility was studied in static systems by exposing AIN-coated specimens to liquid Li for several time periods. Electrical resistance was measured at room temperature on the specimens before and after exposure to liquid Li. The results obtained in this study indicate that AIN is a viable coating from the standpoint of chemical compatibility in Li, electrical insulation, and ease of fabrication; for these reasons, the coating should be examined further for fusion reactor applications

  15. U. S. Fusion Energy Future

    International Nuclear Information System (INIS)

    Schmidt, John A.; Jassby, Dan; Larson, Scott; Pueyo, Maria; Rutherford, Paul H.

    2000-01-01

    Fusion implementation scenarios for the US have been developed. The dependence of these scenarios on both the fusion development and implementation paths has been assessed. A range of implementation paths has been studied. The deployment of CANDU fission reactors in Canada and the deployment of fission reactors in France have been assessed as possible models for US fusion deployment. The waste production and resource (including tritium) needs have been assessed. The conclusion that can be drawn from these studies is that it is challenging to make a significant impact on energy production during this century. However, the rapid deployment of fission reactors in Canada and France support fusion implementation scenarios for the US with significant power production during this century. If the country can meet the schedule requirements then the resource needs and waste production are found to be manageable problems

  16. Investigation on welding and cutting methods for blanket support legs of fusion experimental reactors

    International Nuclear Information System (INIS)

    Tokami, Ikuhide; Nakahira, Masataka; Kurasawa, Toshimasa; Sato, Satoshi; Furuya, Kazuyuki; Hatano, Toshihisa; Takatsu, Hideyuki; Kuroda, Toshimasa.

    1996-07-01

    A toroidally-and poloidally-divided modular blanket has been proposed for a fusion experimental reactor, such as ITER, to enhance its maintainability as well as improve its fabricability. The blanket module, typically the size of 1 m wide, 1-2 m high and 0.4 m deep and the weight of 4 ton, will be supported by support legs which are extruded from back of the module and connected to a 70-100 mm thick strong back plate. The support leg has to withstand large electromagnetic force during plasma disruption and provide the way for in-situ module replacement by remote handling. For the connection method of the support leg to the back plate, a welding approach has been investigated here in terms of its high reliability against the large electromagnetic loads. For the welding approach, the support leg needs to be 70 mm thick, and the working space for welding/cutting heads are limited to 100 mm x 150 mm adjacent to the support leg. Based on a comparison of several welding methods, e.g. NGTIG, NGMIG and laser, NGTIG has been selected as a reference due to its well-established technology and the least R and D required. As for the cutting method, a plasma cutting has been given the highest priority to be pursued because of its compactness and high speed. Through preliminary design studies, the possibility of small welding/cutting heads that will work in the limited space has been shown, and maintenance route for in-situ module replacement with pre-and postfixture of the module has been investigated. Also preliminary R and Ds have resulted in; 1)the welding distortion is predictable according to the shape of weld groove and adjustable to meet the placement requirement of the module first wall, 2)the plasma cut surface can be rewelded without machining, 3)the welding/cutting time will meet the requirement of maintenance time. (author)

  17. 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

  18. Preliminary analyses of neutronics schemes for three kinds waste transmutation blankets of fusion-fission hybrid

    International Nuclear Information System (INIS)

    Zhang Mingchun; Feng Kaiming; Li Zaixin; Zhao Fengchao

    2012-01-01

    The neutronics schemes of the helium-cooled waste transmutation blanket, sodium-cooled waste transmutation blanket and FLiBe-cooled waste transmutation blanket were preliminarily calculated and analysed by using the spheroidal tokamak (ST) plasma configuration. The neutronics properties of these blankets' were compared and analyzed. The results show that for the transmutation of "2"3"7Np, FLiBe-cooled waste transmutation blanket has the most superior transmutation performance. The calculation results of the helium-cooled waste transmutation blanket show that this transmutation blanket can run on a steady effective multiplication factor (k_e_f_f), steady power (P), and steady tritium production rate (TBR) state for a long operating time (9.62 years) by change "2"3"7Np's initial loading rate of the minor actinides (MA). (authors)

  19. Impact of fusion-fission hybrids on world nuclear future

    International Nuclear Information System (INIS)

    Abdel-Khalick, S.; Jansen, P.; Kessler, G.; Klumpp, P.

    1980-08-01

    An investigation has been conducted to examine the impact of fusion-fission hybrids on world nuclear future. The primary objectives of this investigation have been: (1) to determine whether hybrids can allow us to meet the projected nuclear component of the world energy demand within current estimates of uranium resources without fast breeders, and (2) to identify the preferred hybrid concept from a resource standpoint. The results indicate that hybrids have the potential to lower the world uranium demand to values well below the resource base. However, the time window for hybrid introduction is quite near and narrow (2000-2020). If historical market penetration rates are assumed, the demand will not be met within the resource base unless hybrids are coupled to the breeders. The results also indicate that from a resource standpoint hybrids which breed their own tritium and have a low blanket energy multiplication are preferable. (orig.) [de

  20. Impact of fusion-fission hybrids on world nuclear future

    International Nuclear Information System (INIS)

    Abdel-Khalik, S.I.

    1980-01-01

    An investigation has been conducted to examine the impact of fusion-fission hybrids on world nuclear future. The primary objectives of this investigation have been (1) to determine whether hybrids can allow us to meet the projected nuclear component of the world energy demand within current estimates of uranium resources with or without fast breeders, and (2) to identify the preferred hybrid concept from a resource standpoint. The results indicate that hybrids have the potential to lower the world uranium demand to values well below the resource base. However, the time window for hybrid introduction is quite near and narrow (2000-2020). If historical market penetration rates are assumed, the demand will not be met within the resource base unless hybrides are coupled to the breeders. The results also indicate that from a resource standpaint hybrids which breed their own tritium and have a low blanket energy multiplication are preferable. (orig.) [de

  1. Impact of fusion-fission hybrids on world nuclear future

    International Nuclear Information System (INIS)

    Abdel-Khalik, S.I.; Jansen, P.; Kessler, G.; Klumpp, P.

    1981-01-01

    An investigation has been conducted to examine the impact of fusion-fission hybrids on world nuclear future. The primary objectives of this investigation have been: (1) to determine whether hybrids can allow us to meet the projected nuclear component of the world energy demand within current estimates of uranium resources with or without fast breeders, and (2) to identify the preferred hybrid concept from a resource standpoint. The results indicate that hybrids have the potential to lower the world uranium demand to values well below the resource base. However, the time window for hybrid introduction is quite near and narrow (2000-2020). If historical market penetration rates are assumed, the demand will not be met within the resource base unless hybrids are coupled to the breeders. The results also indicate that from a resource standpoint hybrids which breed their own tritium and have a low blanket energy multiplication are preferable. (orig.) [de

  2. Phase IIA and IIB experiments of JAERI/U.S.DOE collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1989-12-01

    Phase IIA and IIB experiments on fusion blanket neutronics has been performed on a basis of JAERI/USDOE collaborative program. In the Phase II experimental series, a D-T neutron source and a test blanket were contained by a lithium-carbonate enclosure to adjust the incident neutron spectrum to the test blanket so as to simulate that of a fusion reactor. First two series of the Phase II, IIA and IIB, focused especially on influences of beryllium configurations for neutron multiplying zone to neutronic parameters. Measured parameters were tritium production rate using Li-glass and NE213 scintillators, and Li-metal foil and Lithium-oxide block with liquid scintillation technique; neutron spectrum using NE213 scintillator and proton recoil proportional counter; reaction rate using foil activation technique. These parameters were compared among six different beryllium configurations of the experimental system. Consistency between different techniques for each measured parameter was also tested among different experimental systems and confirmed to be within experimental errors. This report describes, in detail, experimental conditions, assemblies, equipments and neutron source in Part I. The part II compiles all information required for a calculational analysis of this experiment, e.g., dimensions of the target room, target assembly, experimental assembly, their material densities and numerical data of experimental results. This compilation provides benchmark data to test calculation models and computing code systems used for a nuclear design of a fusion reactor. (author)

  3. 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.

  4. Supercritical CO2 Brayton power cycles for DEMO (demonstration power plant) fusion reactor based on dual coolant lithium lead blanket

    International Nuclear Information System (INIS)

    Linares, José Ignacio; Cantizano, Alexis; Moratilla, Beatriz Yolanda; Martín-Palacios, Víctor; Batet, Lluis

    2016-01-01

    This paper presents an exploratory analysis of the suitability of supercritical CO 2 Brayton power cycles as alternative energy conversion systems for a future fusion reactor based on a DCLL (dual coolant lithium-lead) blanket, as prescribed by EUROfusion. The main issue dealt is the optimization of the integration of the different thermal sources with the power cycle in order to achieve the highest electricity production. The analysis includes the assessment of the pumping consumption in the heating and cooling loops, taking into account additional considerations as control issues and integration of thermal energy storage systems. An exergy analysis has been performed in order to understand the behavior of each layout. Up to ten scenarios have been analyzed assessing different locations for thermal sources heat exchangers. Neglecting the worst four scenarios, it is observed less than 2% of variation among the other six ones. One of the best six scenarios clearly stands out over the others due to the location of the thermal sources in a unique island, being this scenario compatible with the control criteria. In this proposal 34.6% of electric efficiency (before the self-consumptions of the reactor but including pumping consumptions and generator efficiency) is achieved. - Highlights: • Supercritical CO 2 Brayton cycles have been proposed for BoP of DCLL fusion reactor. • Integration of different available thermal sources has been analyzed considering ten scenarios. • Neglecting the four worst scenarios the electricity production varies less than 2%. • Control and energy storage integration issues have been considered in the analysis. • Discarding the vacuum vessel and joining the other sources in an island is proposed.

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

    International Nuclear Information System (INIS)

    Hofer, D.; Kamlah, M.

    2005-01-01

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

  6. Joining of SiC/SiCf ceramic matrix composites for fusion reactor blanket applications

    International Nuclear Information System (INIS)

    Colombo, P.; Riccardi, B.; Donato, A.; Scarinci, G.

    2000-01-01

    Using a preceramic polymer, joints between SiC/SiC f ceramic matrix composites were obtained. The polymer, upon pyrolysis at high temperature, transforms into a ceramic material and develops an adhesive bonding with the composite. The surface morphology of 2D and 3D SiC/SiC f composites did not allow satisfactory results to be obtained by a simple application of the method initially developed for monolithic SiC bodies, which employed the use of a pure silicone resin. Thus, active or inert fillers were mixed with the preceramic polymer, in order to reduce its volumetric shrinkage which occurs during pyrolysis. In particular, the joints realized using the silicone resin with Al-Si powder as reactive additive displayed remarkable shear strength (31.6 MPa maximum). Large standard deviation for the shear strength has nevertheless been measured. The proposed joining method is promising for the realization of fusion reactor blanket structures, even if presently the measured strength values are not fully satisfactory

  7. Study of MHD problems in liquid metal blankets of fusion reactors

    International Nuclear Information System (INIS)

    Michael, I.

    1984-12-01

    This study describes in a concise form the state of knowledge regarding MHD problems to be expected in case of use of liquid metal in the blankets of fusion reactors with magnetic confinement. MHD pressure losses and MHD friction coefficients in the straight channel, in bent sections and in case of variation of the channel cross section play a major role because the high MHD flow resistances call for high pumping powers. Influencing the velocity profile transverse to the main flow direction of the liquid metal by application of an external, strong magnetic field bears consequences on the release and transport of corrosion products in the liquid metal circuit and on the heat transfer. Possibilities of reducing the MHD effects are discussed. However, it becomes obvious that an account of the lack of experimental results there are still major gaps in the knowledge of MHD effects occurring in strong magnetic fields. These gaps can be greatly reduced by implementation of an experimental program as proposed in this report. (orig.) [de

  8. Structural effects on fusion reactor blankets due to liquid metals in magnetic fields

    International Nuclear Information System (INIS)

    Lehner, J.R.; Reich, M.; Powell, J.R.

    1976-01-01

    The transient stress distribution caused in the blanket structure when the plasma current suddenly switches off in a time short compared to the L/R decay time of the liquid metal blanket was studied. Poloidal field of the plasma will induce a current to flow in the liquid metal and blanket walls. Since the resistance of the liquid lithium will be much less than that of the metal walls, the current can be considered as flowing around the blanket near the cross section perimeter, but in the lithium

  9. 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.)

  10. Thermonuclear fusion: Current status and future prospects

    International Nuclear Information System (INIS)

    Bruhns, H.; Maisonnier, Ch.

    1992-01-01

    Thermonuclear Fusion holds great promises for becoming an important energy source for the future. Fusion research and development is undertaken in al major countries of the world. The European Community pursues fusion in a large programme which embraces all R and D in the field of magnetic confinement fusion in the Member States, and to which Sweden and Switzerland are fully associated. The long-term objective of the programme is the joint creation of safe, environmentally sound prototype reactors. The main R and D line of the Community Fusion Programme is fusion by toroidal magnetic confinement on the basis of the Tokamak concept. Some related concepts are also studied which possibly could offer advantages for a reactor, and keep-in-touch activities exist for other approaches. Several small and medium sized specialised devices in Associated Laboratories have been built by the Community Fusion Programme as well as the Joint European Torus (JET Joint Undertaking) which is the largest and the most successful fusion device in the world. Recently, fusion power in the megawatt range has been achieved in JET. The long timescale and the large effort needed for the development of fusion as an energy source have been important elements to foster international collaboration. Engineering Design Activities for an International Thermonuclear Experimental Reactor (ITER) are undertaken, under the auspices of the IAEA, by the European Community, Japan, the Russian Federation and the United States of America. The objective of ITER is to achieve self-sustained thermonuclear burn and its control under long-pulse operation and to provide basic data for the engineering of a demonstration fusion reactor. (author)

  11. 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

  12. Radiation facilities for fusion-reactor first-wall and blanket structural-materials development

    International Nuclear Information System (INIS)

    Klueh, R.L.; Bloom, E.E.

    1981-12-01

    Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed

  13. Extensive neutronic sensitivity-uncertainty analysis of a fusion reactor shielding blanket

    International Nuclear Information System (INIS)

    Hogenbirk, A.

    1994-01-01

    In this paper the results are presented of an extensive neutronic sensitivity-uncertainty study performed for the design of a shielding blanket for a next-step fusion reactor, such as ITER. A code system was used, which was developed at ECN Petten. The uncertainty in an important response parameter, the neutron heating in the inboard superconducting coils, was evaluated. Neutron transport calculations in the 100 neutron group GAM-II structure were performed using the code ANISN. For the sensitivity and uncertainty calculations the code SUSD was used. Uncertainties due to cross-section uncertainties were taken into account as well as uncertainties due to uncertainties in energy and angular distributions of scattered neutrons (SED and SAD uncertainties, respectively). The subject of direct-term uncertainties (i.e. uncertainties due to uncertainties in the kerma factors of the superconducting coils) is briefly touched upon. It is shown that SAD uncertainties, which have been largely neglected until now, contribute significantly to the total uncertainty. Moreover, the contribution of direct-term uncertainties may be large. The total uncertainty in the neutron heating, only due to Fe cross-sections, amounts to approximately 25%, which is rather large. However, uncertainty data are scarce and the data may very well be conservative. It is shown in this paper that with the code system used, sensitivity and uncertainty calculations can be performed in a straightforward way. Therefore, it is suggested that emphasis is now put on the generation of realistic, reliable covariance data for cross-sections as well as for angular and energy distributions. ((orig.))

  14. Magnetohydrodynamic pressure drop and flow balancing of liquid metal flow in a prototypic fusion blanket manifold

    Science.gov (United States)

    Rhodes, Tyler J.; Smolentsev, Sergey; Abdou, Mohamed

    2018-05-01

    Understanding magnetohydrodynamic (MHD) phenomena associated with the flow of electrically conducting fluids in complex geometry ducts subject to a strong magnetic field is required to effectively design liquid metal (LM) blankets for fusion reactors. Particularly, accurately predicting the 3D MHD pressure drop and flow distribution is important. To investigate these topics, we simulate a LM MHD flow through an electrically non-conducting prototypic manifold for a wide range of flow and geometry parameters using a 3D MHD solver, HyPerComp incompressible MHD solver for arbitrary geometry. The reference manifold geometry consists of a rectangular feeding duct which suddenly expands such that the duct thickness in the magnetic field direction abruptly increases by a factor rexp. Downstream of the sudden expansion, the LM is distributed into several parallel channels. As a first step in qualifying the flow, a magnitude of the curl of the induced Lorentz force was used to distinguish between inviscid, irrotational core flows and boundary and internal shear layers where inertia and/or viscous forces are important. Scaling laws have been obtained which characterize the 3D MHD pressure drop and flow balancing as a function of the flow parameters and the manifold geometry. Associated Hartmann and Reynolds numbers in the computations were ˜103 and ˜101-103, respectively, while rexp was varied from 4 to 12. An accurate model for the pressure drop was developed for the first time for inertial-electromagnetic and viscous-electromagnetic regimes based on 96 computed cases. Analysis shows that flow balance can be improved by lengthening the distance between the manifold inlet and the entrances of the parallel channels by utilizing the effect of flow transitioning to a quasi-two-dimensional state in the expansion region of the manifold.

  15. Composition Optimization of Lithium-Based Ternary Alloy Blankets for Fusion Reactors

    Science.gov (United States)

    Jolodosky, Alejandra

    The goal of this dissertation is to examine the neutronic properties of a novel type of fusion reactor blanket material in the form of lithium-based ternary alloys. Pure liquid lithium, first proposed as a blanket for fusion reactors, is utilized as both a tritium breeder and a coolant. It has many attractive features such as high heat transfer and low corrosion properties, but 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 presents plant safety concerns including degradation of the concrete containment structure. The work of this thesis began as a collaboration with Lawrence Livermore National Laboratory in an effort to develop a lithium-based ternary alloy that can maintain the beneficial properties of lithium while reducing the reactivity concerns. The first studies down-selected alloys based on the analysis and performance of both neutronic and activation characteristics. First, 3-D Monte Carlo calculations were performed to evaluate two main neutronics performance parameters for the blanket: tritium breeding ratio (TBR), and energy multiplication factor (EMF). Alloys with adequate results based on TBR and EMF calculations were considered for activation analysis. Activation simulations were executed with 50 years of irradiation and 300 years of cooling. It was discovered that bismuth is a poor choice due to achieving the highest decay heat, contact dose rates, and accident doses. In addition, it does not meet the waste disposal ratings (WDR). The straightforward approach to obtain Monte Carlo TBR and EMF results required 231 simulations per alloy and became computationally expensive, time consuming, and inefficient. Consequently, alternate methods were pursued. A collision history-based methodology recently developed for the Monte Carlo code Serpent, calculates perturbation effects on practically

  16. 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

  17. Effects of neutron source ratio on nuclear characteristics of D-D fusion reactor blankets and shields

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Nakao, Yasuyuki; Ohta, Masao

    1978-01-01

    An examination is made of the dependence shown by the nuclear characteristics of the blanket and shield of D-D fusion reactors on S sub( d d)/S sub( d t), the ratio between the 2.45 MeV neutrons resulting from the D-D reaction and those of 14.06 MeV from the D-T reaction. Also, an estimate is presented of this neutron source ratio S sub( d d)/S sub( d t) for the case of D-D reactors, taken as an example. It is shown that an increase of S sub( d d)/S sub( d t) reduces the amount of nuclear heating per unit source neutron, while at the same time improving the shielding characteristics. This is accountable to lowering of the energy and penetrability of incident neutrons into the blanket brought about by the increase of S sub( d d)/S sub( d t). The value of S sub( d d)/S sub( d t) in a steady state D-D fusioning plasma core is estimated to be 1.46 -- 1.72 for an ion temperature ranging from 60 -- 180 keV. The reductions obtained on H sub( t)sup( b) (total heating in the blanket), H sub( t)sup( m g)/H sub( t)sup( b) (shielding indicator = ratio between total heating in superconducting magnet and that in the blanket) and phi sup( m g)/phi sup( w) (ratio of fast neutron fluxes between that at the magnet inner surface and that at the first wall inner surface) brought about by increasing S sub( d d)/S sub( d t) from unity to the value cited above do not differ to any appreciable extent, whichever is adopted among the design models considered here, the differences being at most about 10, 15 and 25%, respectively, for these three parameters. These results would broaden the validity of the conclusion derived in the previous paper for the case of S sub( d d)/S sub( d t) = 1.0, that the blanket-shield concept would appear to be the most suitable for D-D fusion reactors. (author)

  18. 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)

  19. Neutronic calculation and cross section sensitivity analysis of the Livermore mirror fusion/fission hybrid reactor blanket

    International Nuclear Information System (INIS)

    Ku, L.P.; Price, W.G. Jr.

    1977-08-01

    The neutronic calculation for the Livermore mirror fusion/fission hybrid reactor blanket was performed using the PPPL cross section library. Significant differences were found in the tritium breeding and plutonium production in comparison to the results of the LLL calculation. The cross section sensitivity study for tritium breeding indicates that the response is sensitive to the cross section of 238 U in the neighborhood of 14 MeV and 1 MeV. The response is also sensitive to the cross sections of iron in the vicinity of 14 MeV near the first wall. Neutron transport in the resonance region is not important in this reactor model

  20. On the role of fusion neutron source with thorium blanket in forming the nuclide composition of the nuclear fuel cycle of the Russian Federation

    Energy Technology Data Exchange (ETDEWEB)

    Shmelev, A. N.; Kulikov, G. G., E-mail: ggkulikov@mephi.ru [National Research Nuclear University (Moscow Engineering Physics Institute) (Russian Federation)

    2016-12-15

    The possible role of available thorium resources of the Russian Federation in utilization of thorium in the closed (U–Pu)-fuel cycle of nuclear power is considered. The efficiency of application of fusion neutron sources with thorium blanket for economical use of available thorium resources is demonstrated. The objective of this study is the search for a solution of such major tasks of nuclear power as reduction of the amount of front-end operations in the nuclear fuel cycle and enhancement of its protection against uncontrolled proliferation of fissile materials with the smallest possible alterations in the fuel cycle. The earlier results are analyzed, new information on the amount of thorium resources of the Russian Federation is used, and additional estimates are made. The following basic results obtained on the basis of the assumption of involving fusion reactors with Th-blanket in future nuclear power for generation of the light uranium fraction {sup 232+233+234}U and {sup 231}Pa are formulated. (1) The fuel cycle would shift from fissile {sup 235}U to {sup 233}U, which is more attractive for thermal power reactors. (2) The light uranium fraction is the most “protected” in the uranium fuel component, and being mixed with regenerated uranium, it would become reduced-enrichment uranium fuel, which would relieve the problem of nonproliferation of the fissile material. (3) The addition of {sup 231}Pa into the fuel would stabilize its neutron-multiplying properties, thus making it possible to implement a long fuel residence time and, as a consequence, increase the export potential of the whole nuclear power technology. (4) The available thorium resource in the vicinity of Krasnoufimsk is sufficient for operation of the large-scale nuclear power industry of the Russian Federation with an electric power of 70 GW for more than one quarter of a century. The general conclusion is that involvement of a small number of fusion reactors with Th-blanket in the future

  1. Fusion looks to the future - again

    International Nuclear Information System (INIS)

    Waldrop, M.M.

    1984-01-01

    The $46 million budget cut in the US magnetic fusion program introduced a new approach that abandons the race to build a working power reactor in favor of a long-term emphasis on science, technology, and international cooperation. Administration policies which favor private funding for demonstration projects and general concern over the deficit have changed the overall fusion policy, although there is some concern among research groups that the program will become unfocused without its detailed timetable. If this happens, they see the program becoming even more vulnerable to future budget cuts. 2 references

  2. A cross-section sensitivity and uncertainty analysis on fusion reactor blankets with SAD/SED effect

    International Nuclear Information System (INIS)

    Furuta, Kazuo; Oka, Yoshiaki; Kondo, Shunsuke

    1986-01-01

    A cross-section sensitivity and uncertainty analysis on four types of fusion reactor blankets has been performed, based on cross-section covariance matrices. The design parameters investigated in the analysis include the tritium breeding ratio, the neutron heating and the fast neutron leakage flux from the inboard shield. Uncertainities in Secondary Angular Distribution (SAD) and Secondary Energy Distribution (SED) of scattered neutrons have been considered for lithium. The collective standard deviation, due to uncertainties in the evaluated cross-section data presently available, is 2-4% in the tritium breeding ratio, 2-3% in the neutron heating, and 10-20% in the fast neutron leakage flux. Contributions from SAD/SED uncertainties are significant for some parameters, such as those investigated in the present study. SAD/SED uncertainties should be considered in the sensitivity and uncertainty analysis on nuclear design of fusion reactors. (orig.)

  3. 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

  4. Initial progress in the first wall, blanket, and shield Engineering Test Program for magnetically confined fusion-power reactors

    International Nuclear Information System (INIS)

    Herman, H.; Baker, C.C.; Maroni, V.A.

    1981-10-01

    The first wall/blanket/shield (FW/B/S) Engineering Test Program (ETP) progressed from the planning stage into implementation during July, 1981. The program, generic in nature, comprises four Test Program Elements (TPE's), the emphasis of which is on defining the performance parameters for the Fusion Engineering Device (FED) and the major fusion device to follow FED. These elements are: (1) nonnuclear thermal-hydraulic and thermomechanical testing of first wall and component facsimiles with emphasis on surface heat loads and heat transient (i.e., plasma disruption) effects; (2) nonnuclear and nuclear testing of FW/B/S components and assemblies with emphasis on bulk (nuclear) heating effects, integrated FW/B/S hydraulics and mechanics, blanket coolant system transients, and nuclear benchmarks; (3) FW/B/S electromagnetic and eddy current effects testing, including pulsed field penetration, torque and force restraint, electromagnetic materials, liquid metal MHD effects and the like; and (4) FW/B/S Assembly, Maintenance and Repair (AMR) studies focusing on generic AMR criteria, with the objective of preparing an AMR designers guidebook; also, development of rapid remote assembly/disassembly joint system technology, leak detection and remote handling methods

  5. 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)

  6. Swiss fusion blanket experiments: Final report, November 1, 1985-October 31, 1987

    International Nuclear Information System (INIS)

    Woodruff, G.L.

    1987-01-01

    The major thrust of this project related to the effort to transfer the Lithium Blanket Module (LBM) to the Nuclear Engineering Laboratory of the Swiss Institute of Technology at Lausanne, and to the subsequent support with analytical calculations of a variety of experiments performed with the LBM. 12 refs

  7. Methods to enhance blanket power density in low-power fusion devices

    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 breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow testing of advanced reactor blanket modules on INTOR at representative conditions. The tentative approach adopted for this task consists of three parts. First, the requirements for augmented heating of the test module are outlined for different applications of interest. Second, methods are identified which have potential for augmenting the heating power in a test module, and this list of methods is narrowed to those which appear to be most useful. Finally, these methods are examined in more detail to determine the practical benefits of employing each

  8. Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

    International Nuclear Information System (INIS)

    Misra, B.; Maroni, V.A.

    1978-01-01

    A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

  9. Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

    International Nuclear Information System (INIS)

    Misra, B.; Maroni, V.A.

    1977-01-01

    A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

  10. Materials science problems of blankets in Russian concept of fusion reactor

    International Nuclear Information System (INIS)

    Solonin, M.I.

    1998-01-01

    Structural materials, beryllium and tritium breeding materials proposed for blanket of Russian reactor DEMO and Test Modules for ITER are discussed. Main requirements for the materials are concerned with basis current designs of blankets and modules and possibility meet of ones for presence and developed alloys and materials discussed considered. Main properties and results of test of ferrite-martensite and vanadium alloys for DEMO and Test Modules are cited. Beryllium compositions used as component of first wall and neutron multiplier are discussed. Liquid lithium and ceramic (lithium orthosilicate) are treated as tritium breeding materials. Russian development of reactor experimental unit for tritium breeding zone using beryllium, lithium ceramic and ferrite-martensite alloys for structural materials is presented. (orig.)

  11. 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

  12. Study of tritium removal from fusion reactor blankets of molten salt and lithium--aluminum

    International Nuclear Information System (INIS)

    Talbot, J.B.

    1976-03-01

    The sorption of tritium by molten lithium--bismuth (Li--Bi, approx. 15 at. % lithium) and solid equiatomic lithium--aluminum (Li--Al) was investigated experimentally to evaluate the potential applications of both materials in a controlled thermonuclear reactor. The Li--Bi alloy was proposed to countercurrently extract tritium from a molten salt (Li 2 BeF 4 ) blanket. However, because of the low solubility ( 0 C, the extraction process is not attractive

  13. 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

  14. Cross section sensitivity study for fusion blankets incorporating lead neutron multiplier

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.

    1983-01-01

    In the recent European INTOR design, lead has been considered for incorporation in the blanket as either an explicit or implicit neutron multiplier. The blanket employs either Li 2 SiO 3 or Li 17 Pb 83 as tritium breeding material. Nucleonic analysis was performed for this blanket using the DLC37 and DLC41 cross section libraries. The reaction rates were estimated using the reaction cross sections provided with both libraries. In addition to that, they were estimated using the MACKLIB-IV response library. The calculated tritium breeding ratio was found to be 5% less and 15% more in the calculations with DLC41 and DLC41 plus MACKLIB-IV libraries, respectively, than in the calculation with the DLC37 library. The Fe, Pb, and Li cross sections given by the ENDF/B-IV and V were reviewed. A sensitivity study of these cross section uncertainties shows that the tritium breeding ratio is relatively insensitive to the above mentioned partial cross sections. The calculated tritium breeding ratio can be known within +-2%. (Auth.)

  15. Supercritical CO2 Brayton power cycles for DEMO fusion reactor based on Helium Cooled Lithium Lead blanket

    International Nuclear Information System (INIS)

    Linares, José Ignacio; Herranz, Luis Enrique; Fernández, Iván; Cantizano, Alexis; Moratilla, Beatriz Yolanda

    2015-01-01

    Fusion energy is one of the most promising solutions to the world energy supply. This paper presents an exploratory analysis of the suitability of supercritical CO 2 Brayton power cycles (S-CO 2 ) for low-temperature divertor fusion reactors cooled by helium (as defined by EFDA). Integration of three thermal sources (i.e., blanket, divertor and vacuum vessel) has been studied through proposing and analyzing a number of alternative layouts, achieving an improvement on power production higher than 5% over the baseline case, which entails to a gross efficiency (before self-consumptions) higher than 42%. In spite of this achievement, the assessment of power consumption for the circulating heat transfer fluids results in a penalty of 20% in the electricity production. Once the most suitable layout has been selected an optimization process has been conducted to adjust the key parameters to balance performance and size, achieving an electrical efficiency (electricity without taking into account auxiliary consumptions due to operation of the fusion reactor) higher than 33% and a reduction in overall size of heat exchangers of 1/3. Some relevant conclusions can be drawn from the present work: the potential of S-CO 2 cycles as suitable converters of thermal energy to power in fusion reactors; the significance of a suitable integration of thermal sources to maximize power output; the high penalty of pumping power; and the convenience of identifying the key components of the layout as a way to optimize the whole cycle performance. - Highlights: • Supercritical CO 2 Brayton cycles have been proposed for BoP of HCLL fusion reactor. • Low temperature sources have been successfully integrated with high temperature ones. • Optimization of thermal sources integration improves 5% the electricity production. • Assessment of pumping power with sources and sink loops results on 20% of gross power. • Matching of key parameters has conducted to 1/3 of reduction in heat

  16. 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.

  17. 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

  18. G8 decision on fusion would herald nuclear future

    CERN Multimedia

    Starck, Peter

    2005-01-01

    Nuclear fusion as a future abundant energy source would receive a boost if G8 leaders agree next month on the site for the world's first fusion test reactor, two nuclear scientists said on Wednesday (1 page)

  19. Heavy-ion fusion: Future promise and future directions

    International Nuclear Information System (INIS)

    Dudziak, D.J.; Saylor, W.W.; Pendergrass, J.H.

    1986-01-01

    The previous several papers in this heavy-ion fusion special session have described work that has taken place as part of the Heavy-Ion Fusion Systems Assessment (HIFSA) project. Key technical issues in the design and costing of targets, accelerator systems, beam transport, reactor and balance-of-plant, and systems integration have been identified and described. The HIFSA systems model was used to measure the relative value of improvements in physics understanding and technology developments in many different areas. The result of this study has been to, within the limits of our 1986 imagination and creativity, define the ''most attractive'' future heavy-ion fusion (HIF) power plant at some time in the future (beyond the year 2020 in this case). The project has specifically avoided narrowing the focus to a point facility design; thus, the generic systems modeling capability developed in the process allows for a relative comparison among design options. The authors describe what are thought to be achievable breakthroughs and what the relative significance of the breakthroughs will be, although the specific mechanism for achieving some breakthroughs may not be clear at this point

  20. Fusion reactors as a future energy source

    International Nuclear Information System (INIS)

    Seifritz, W.

    A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors, (8) fuel supply for fusion reactors, (9) safety and environmental impact, and (10) accidents, and (11) waste removal and storage

  1. Analysis of in-situ tritium recovery from solid fusion-reactor blankets

    International Nuclear Information System (INIS)

    Smith, D.L.; Clemmer, R.G.; Jankus, V.Z.; Rest, J.

    1980-01-01

    The proposed concept for in-situ tritium recovery from the STARFIRE blanket involves circulation of a low pressure (approx. 0.05 MPa) helium through formed channels in the highly porous solid breeding material. Tritium generated within the grains must diffuse to the grain boundaries, migrate through the grain boundaries to the particle surface and then percolate through the packed bed to the helium purge channel. Highly porous α-LiAlO 2 with a bimodal pore distribution is proposed for the breeding material to facilitate the tritium release

  2. Study of tritium removal from fusion reactor blankets of molten salt and lithium--aluminum

    International Nuclear Information System (INIS)

    Talbot, J.B.

    1976-03-01

    The sorption of tritium by molten lithium--bismuth (Li--Bi, approximately 15 at. percent lithium) and solid equiatomic lithium--aluminum (Li--Al) was investigated experimentally to evaluate the potential applications of both materials in a controlled thermonuclear reactor. The Li--Bi alloy was proposed to countercurrently extract tritium from a molten salt (Li 2 BeF 4 ) blanket. However, because of the low solubility (less than 10 ppb) at temperatures ranging from 500 to 700 0 C, the extraction process is not attractive

  3. 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.

  4. 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)

  5. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2005-03-01

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

  6. European research and development programme for water-cooled lithium-lead blankets: present status and future work

    International Nuclear Information System (INIS)

    Giancarli, L.; Leroy, P.; Proust, E.; Raepsaet, X.

    1992-01-01

    The European R and D programme in support of the development of water-cooled Pb-17Li blankets for DEMO aims at improving the data base concerning tritium behaviour and compatibility between blanket materials. The four main areas of the experimental programme are structural material corrosion by Pb-17Li, tritium extraction and permeation control.=, Pb-17Li physico-chemistry, and water/Pb-17Li interaction. This paper describes the most significant results obtained to date in the various experiments performed in Europe and the future programme required to complete the data base by 1994. 28 refs

  7. Failure initiation and propagation of Li4SiO4 pebbles in fusion blankets

    International Nuclear Information System (INIS)

    Zhao Shuo; Gan Yixiang; Kamlah, Marc

    2013-01-01

    Lithium orthosilicate (Li 4 SiO 4 ) pebbles are considered to be a candidate as solid tritium breeder in the helium cooled pebble bed (HCPB) blanket. These ceramic pebbles might be crushed during thermomechanical loading in the blanket. In this work, the failure initiation and propagation of pebbles in pebble beds is investigated using the discrete element method (DEM). Pebbles are simplified as mono-sized elastic spheres. Every pebble has a contact strength in terms of critical strain energy, which is derived from a validated strength model and crush test data for pebbles from a specific batch of Li 4 SiO 4 pebbles. Pebble beds are compressed uniaxially and triaxially in DEM simulations. When the strain energy absorbed by a pebble exceeds its critical energy it fails. The failure initiation is defined as a given small fraction of pebbles crushed. It is found that the load level for failure initiation can be very low. For example, if failure initiation is defined as soon as 0.02% of the pebbles have been crushed, the pressure required for uniaxial loading is about 2.5 MPa. Therefore, it is essential to study the influence of failure propagation on the macroscopic response of pebble beds. Thus a reduction ratio defined as the size ratio of a pebble before and after its failure is introduced. The macroscopic stress–strain relation is investigated with different reduction ratios. A typical stress plateau is found for a small reduction ratio.

  8. Fusion energy - an abundant energy source for the future

    DEFF Research Database (Denmark)

    Fusion energy is the fundamental energy source of the Universe, as the energy of the Sun and the stars are produced by fusion of e.g. hydrogen to helium. Fusion energy research is a strongly international endeavor aiming at realizing fusion energy production in power plants on Earth. Reaching...... this goal, mankind will have a sustainable base load energy source with abundant resources, having no CO2 release, and with no longlived radioactive waste. This presentation will describe the basics of fusion energy production and the status and future prospects of the research. Considerations...... of integration into the future electricity system and socio-economic studies of fusion energy will be presented, referring to the programme of Socio-Economic Research on Fusion (SERF) under the European Fusion Energy Agreement (EFDA)....

  9. 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

  10. An induction-based magnetohydrodynamic 3D code for finite magnetic Reynolds number liquid-metal flows in fusion blankets

    International Nuclear Information System (INIS)

    Kawczynski, Charlie; Smolentsev, Sergey; Abdou, Mohamed

    2016-01-01

    Highlights: • A new induction-based magnetohydrodynamic code was developed using a finite difference method. • The code was benchmarked against purely hydrodynamic and MHD flows for low and finite magnetic Reynolds number. • Possible applications of the new code include liquid-metal MHD flows in the breeder blanket during unsteady events in the plasma. - Abstract: Most numerical analysis performed in the past for MHD flows in liquid-metal blankets were based on the assumption of low magnetic Reynolds number and involved numerical codes that utilized electric potential as the main electromagnetic variable. One limitation of this approach is that such codes cannot be applied to truly unsteady processes, for example, MHD flows of liquid-metal breeder/coolant during unsteady events in plasma, such as major plasma disruptions, edge-localized modes and vertical displacements, when changes in plasmas occur at millisecond timescales. Our newly developed code MOONS (Magnetohydrodynamic Object-Oriented Numerical Solver) uses the magnetic field as the main electromagnetic variable to relax the limitations of the low magnetic Reynolds number approximation for more realistic fusion reactor environments. The new code, written in Fortran, implements a 3D finite-difference method and is capable of simulating multi-material domains. The constrained transport method was implemented to evolve the magnetic field in time and assure that the magnetic field remains solenoidal within machine accuracy at every time step. Various verification tests have been performed including purely hydrodynamic flows and MHD flows at low and finite magnetic Reynolds numbers. Test results have demonstrated very good accuracy against known analytic solutions and other numerical data.

  11. An induction-based magnetohydrodynamic 3D code for finite magnetic Reynolds number liquid-metal flows in fusion blankets

    Energy Technology Data Exchange (ETDEWEB)

    Kawczynski, Charlie; Smolentsev, Sergey, E-mail: sergey@fusion.ucla.edu; Abdou, Mohamed

    2016-11-01

    Highlights: • A new induction-based magnetohydrodynamic code was developed using a finite difference method. • The code was benchmarked against purely hydrodynamic and MHD flows for low and finite magnetic Reynolds number. • Possible applications of the new code include liquid-metal MHD flows in the breeder blanket during unsteady events in the plasma. - Abstract: Most numerical analysis performed in the past for MHD flows in liquid-metal blankets were based on the assumption of low magnetic Reynolds number and involved numerical codes that utilized electric potential as the main electromagnetic variable. One limitation of this approach is that such codes cannot be applied to truly unsteady processes, for example, MHD flows of liquid-metal breeder/coolant during unsteady events in plasma, such as major plasma disruptions, edge-localized modes and vertical displacements, when changes in plasmas occur at millisecond timescales. Our newly developed code MOONS (Magnetohydrodynamic Object-Oriented Numerical Solver) uses the magnetic field as the main electromagnetic variable to relax the limitations of the low magnetic Reynolds number approximation for more realistic fusion reactor environments. The new code, written in Fortran, implements a 3D finite-difference method and is capable of simulating multi-material domains. The constrained transport method was implemented to evolve the magnetic field in time and assure that the magnetic field remains solenoidal within machine accuracy at every time step. Various verification tests have been performed including purely hydrodynamic flows and MHD flows at low and finite magnetic Reynolds numbers. Test results have demonstrated very good accuracy against known analytic solutions and other numerical data.

  12. Fusion and its future in Illinois

    International Nuclear Information System (INIS)

    Baker, C.C.

    1984-08-01

    This report was prepared by the Illinois Fusion Power Task Force under the sponsorship of the Governor's Commission on Sciences and Technology. The report presents the findings and recommendations of the Task Force, an explanation of the basic concepts of fusion, a summary of national and international programs and a description of ongoing fusion activities in Illinois

  13. Comparative study of the more promising combinations of blanket materials, power conversion systems, and tritium recovery and containment systems for fusion reactors

    International Nuclear Information System (INIS)

    Fraas, A.P.

    1975-11-01

    The many possible combinations of blanket materials, tritium generation and recovery systems, and power conversion systems were surveyed first by reviewing the principal design studies that have been prepared and then by examining a comprehensive set of designs generated by using a common set of ground rules that included all of the boundary conditions that could be envisioned. The results indicate that, of the wide variety of systems that have been considered, by far the most promising employs lithium recirculated in a closed loop within a niobium blanket structure and cooled with boiling potassium or cesium. This approach gives the simplest and lowest cost tritium recovery system, the lowest pressure and thermal stresses, the simplest structure with the lowest probability of a leak, the greatest resistance to damage from a plasma energy dump, and the lowest rate of plasma contamination by either outgassing or sputtering. The only other blanket materials combination that appears fairly likely to give a satisfactory tritium generation and recovery system is an Li 2 BeF 4 -Incoloy blanket, and even this system involves major uncertainties in the effectiveness, size, and cost of the tritium recovery system. Further, the Li 2 BeF 4 blanket system has the disadvantage that the world reserves of beryllium are too limited to support a full-blown fusion reactor economy, its poor thermal conductivity leads to cooling difficulties and a requirement for a complex structure with intricate cooling passages, and this inherently leads to an expensive blanket with a relatively high probability of leaks. The other blanket materials combinations yield even less attractive systems

  14. 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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  16. Mechanical properties of materials in fusion reactor first-wall and blanket systems

    Energy Technology Data Exchange (ETDEWEB)

    Bloom, E.E.

    1979-01-01

    With respect to the effects of irradiation on mechanical properties, the most significant difference between fast fission and fusion reactor spectra is the relatively large amount of helium produced by (n,..cap alpha..) transmutations in the latter. Relevant information on the effects of large amounts of helium (with concomitant displacement damage) comes from irradiation of alloys containing nickel in mixed spectrum reactors. At helium levels of interest for fusion reactor development, properties are degraded to unacceptable levels above Tm/2. Below this temperature, strength and ductility are retained and fractures remain transgranular. Importantly, the properties remain sensitive to composition and structure. A comparison of the response of bcc refractory alloys to that of stainless steel at equivalent damage levels shows the same general trends in properties with homologous temperature. The refractory alloys do offer potential for higher temperature applications because of their melting temperatures.

  17. Self-consistent Analysis of a Blanket and Shielding of a Fusion Reactor Concept

    International Nuclear Information System (INIS)

    Kim, Suk Kwon; Hong, B. G.; Lee, D. W.; Kim, D. H.; Lee, Y. O.

    2008-01-01

    To develop the concept of a DEMO reactor, a tokamak reactor system analysis code has been developed at KAERI (Korea Atomic Energy Research Institute). The system analysis code incorporates prospects of the development of plasma physics and the technologies in a simple mathematical model and it helps to develop the concept of a fusion reactor and to identify the necessary R and D areas for a realization of the concept. In the system code, a plant power balance equation and a plasma power balance equation are solved to find plant parameters which satisfy the plasma physics and technology constraints, simultaneously. The outcome of the system analysis is to identify which areas of plasma physics and technologies and to what extent they should be developed for a realization of given fusion reactor concepts

  18. Mechanical properties of materials in fusion reactor first-wall and blanket systems

    International Nuclear Information System (INIS)

    Bloom, E.E.

    1979-01-01

    With respect to the effects of irradiation on mechanical properties, the most significant difference between fast fission and fusion reactor spectra is the relatively large amount of helium produced by (n,α) transmutations in the latter. Relevant information on the effects of large amounts of helium (with concomitant displacement damage) comes from irradiation of alloys containing nickel in mixed spectrum reactors. At helium levels of interest for fusion reactor development, properties are degraded to unacceptable levels above Tm/2. Below this temperature, strength and ductility are retained and fractures remain transgranular. Importantly, the properties remain sensitive to composition and structure. A comparison of the response of bcc refractory alloys to that of stainless steel at equivalent damage levels shows the same general trends in properties with homologous temperature. The refractory alloys do offer potential for higher temperature applications because of their melting temperatures

  19. Progress in the development of the blanket structural material for fusion reactors

    International Nuclear Information System (INIS)

    Scott, J.L.; Bloom, E.E.; Grossbeck, M.L.; Maziasz, P.J.; Wiffen, F.W.; Gold, R.E.; Holmes, J.J.; Reuther, P.C. Jr.; Rosenwasser, S.N.

    1981-01-01

    The Alloy Development for Irradiation Performance Program has become more focused since the last Fusion Reactor Technology Conference two years ago. Since austenitic stainless steels and ferritic steels are candidate structural materials for the near-term reactors ETF and INTOR and austenitic stainless steel is also the preferred structural material for the steady-state commercial fusion reactor, STARFIRE, a vigorous experimental program is under way to identify the best alloy from each of these alloy classes and to provide the engineering data base in a timely manner. In addition the comprehensive program that includes high-strength Fe-Ni-Cr alloys, reactive and refractory metals, and advanced concepts continues in an orderly fashion

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  1. Future directions in inertial confinement fusion

    International Nuclear Information System (INIS)

    Bodner, S.E.

    1992-01-01

    The author discusses future directions for the ICF program. At this time there is still uncertainty on a number of key issues necessary to decide on what type of a National Ignition Facility should be constructed. Mechanisms are in place to answer these questions. The author offers his opinions of where the program is likely to proceed. Technology wise indications are that direct drive heating has the best chance of reaching ignition and high gain. This has the advantage of making all three major user programs happy, namely weapons physics, weapons effects, and electrical energy. The demand for and price of energy in the country will have a major impact on the way the program is developed. From the laser fusion side the most promising drivers at present seem to be KrF lasers, and a major concern for these systems is whether the peak to valley nonuniformities can be reduced to the 1 to 2% level when delivered to the target in order to avoid driving instabilities

  2. Reduction of circulation power for helium-cooled fusion reactor blanket using additive CO{sub 2} gas

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yeon-Gun [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Department of Nuclear and Energy Engineering, Jeju National University, 102 Jejudaehakno, Jeju-si 690-756, Jeju (Korea, Republic of); Park, Il-Woong [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Lee, Dong Won [Nuclear Fusion Engineering Development Center, Korea Atomic Energy Research Institute, Daedeokdaero 989 beon-gil, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Kim, Eung-Soo, E-mail: kes7741@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of)

    2015-11-15

    Helium (He) cooling requires large circulation power to remove high heat from plasma side and nuclear heating by high energy neutron in fusion reactors due to its low density. Based on the recent findings that the heat transfer capability of the light gas can be enhanced by mixing another heavier gas, this study adds CO{sub 2} to a reference helium coolant and evaluates the cooling performance of the binary mixture for various compositions. To assess the cooling performance, computational fluid dynamic (CFD) analyses on the KO HCML (Korea Helium Cooled Molten Lithium) TBM are conducted. As a result, it is revealed that the binary mixing of helium, which has favorable thermophysical properties but the density, with a heavier noble gas or an unreactive gas significantly reduces the required circulation power by an order of magnitude with meeting the thermal design requirements. This is attributed to the fact that the density can be highly increased with small amount of a heavier gas while other gas properties are kept relatively comparable. The optimal CO{sub 2} mole fraction is estimated to be 0.4 and the circulation power, in this case, can be reduced to 13% of that of pure helium. This implies that the thermal efficiency of a He-cooled blanket system can be fairly enhanced by means of the proposed binary mixing.

  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. Some initial considerations on the suitability of Ferritic/ martensitic stainless steels as first wall and blanket materials in fusion reactors

    International Nuclear Information System (INIS)

    Butterworth, G.J.

    1982-01-01

    The constitution of stainless iron alloys and the characteristic properties of alloys in the main ferritic, martensitic and austenitic groups are discussed. A comparison of published data on the mechanical, thermal and irradiation properties of typical austenitic and martensitic/ferritic steels shows that alloys in the latter groups have certain advantages for fusion applications. The ferromagnetism exhibited by martensitic and ferritic alloys has, however, been identified as a potentially serious obstacle to their utilisation in magnetic confinement devices. The paper describes measurements performed in other laboratories on the magnetic properties of two representative martensitic alloys 12Cr-1Mo and 9Cr-2Mo. These observations show that a modest bias magnetic field of magnitude 1 - 2 tesla induces a state of magnetic saturation in these materials. They would thus behave as essentially paramagnetic materials having a relative permeability close to unity when saturated by the toroidal field of a tokamak reactor. The results of computations by the General Atomic research group to assess the implications of such magnetic behaviour on reactor design and operation are presented. The results so far indicate that the ferromagnetism of martensitic/ferritic steels would not represent a major obstacle to their utilisation as first wall or blanket materials. (author)

  5. Neutronic design of pulse operation simulating device for in-pile functional test of fusion blanket by MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu; Nakamichi, Masaru; Kawamura, Hiroshi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan)

    2000-03-01

    The pulse operation of a fusion reactor can be simulated in a fission reactor by controlling the neutron flux entering a test section by using a rotating 'hollow cylinder with window' made of hafnium. The rotating cylinder is installed between the test section and the fixed outer neutron absorber cylinder and is also made of hafnium with an opening in the direction to the core center. For gathering engineering data for the tritium breeding blanket such as characteristics of temperature change, tritium release and recovery, etc., it is desirable that the ratio of minimum to maximum thermal neutron fluxes is greater than 1:10. Design calculations were performed for the test assembly which considered local neutronic effects and the mechanical constraints of the device. From the results of these calculations, the ratio of minimum to maximum thermal neutron flux under irradiation would be about 1:10 using a pulse operation simulating device which has a thickness of 6.5 mm and a 150deg window angle for the rotating hollow cylinder and 5.0 mm in thickness of fixed neutron absorber. (author)

  6. TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey James [Univ. of California, Berkeley, CA (United States)

    2011-11-30

    This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated

  7. Nuclear fusion an energetic option to the future

    International Nuclear Information System (INIS)

    Medialdea Utande, A.; Sanchez Sanz, J.

    2007-01-01

    Nuclear fusion is the energy source of the Sun and the rest of starts. The great availability of deuterium on Earth, the inherent safety of the reactions involved and the intrinsic environmental respect make fusion an attractive energy source for the future of making of man king. International promising contributions are making Fusion Science and Technology progress by leaps and bounds to achieve its long term goal of cost-effective energy-producing plasmas. (Author)

  8. Past, present and future of the fusion reactors

    International Nuclear Information System (INIS)

    Rosenbaum P, M.

    1992-01-01

    Among the alternate technologies that have acquired a special interest in the present decade, we find the nuclear fusion. Within this, the fusion reactors by magnetic confinement of the Tokamak type have shown an increasing technological progress during this period. For this reason, a new strategy, coordinated at international level, has been implemented for the specific development of the nuclear fusion reactors, aimed to face those scientific and technological aspects which still remain, and which will determine their future economic feasibility. (Author)

  9. Laser Fusion: status, future, and tritium control

    International Nuclear Information System (INIS)

    Coyle, P.E.

    1978-11-01

    At Livermore the 10 kJ, 20 to 30 TW Shiva facility is now operational and producing regular new fusion results. Design work has begun on a 200 to 300 TW laser designed to carry the program through the first breakeven demonstration experiments in the mid-1980's. Confidence in reaching this goal is based on the significant progress we have made in state-of-the-art, high-power Nd:glass laser technology, in experimental laser fusion and laser plasma interaction physics, and in theoretical and analytical computer codes which reliably model and predict experimental results. For all of these experiments, a variety of fusion targets are being fabricated in the laboratory, and the control and handling of tritium is now a regular and routine part of ongoing inertial fusion experiments. Target design with gains of about 1000 have been studied and the means to mass produce such pellets at low cost are also being developed

  10. The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

    International Nuclear Information System (INIS)

    Zheng Youqi; Zu Tiejun; Wu Hongchun; Cao Liangzhi; Yang Chao

    2012-01-01

    In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.

  11. 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

  12. Thermo-mechanical Modelling of Pebble Beds in Fusion Blankets and its Implementation by a Return-Mapping Algorithm

    International Nuclear Information System (INIS)

    Gan, Yixiang; Kamlah, Marc

    2008-01-01

    In this investigation, a thermo-mechanical model of pebble beds is adopted and developed based on experiments by Dr. Reimann at Forschungszentrum Karlsruhe (FZK). The framework of the present material model is composed of a non-linear elastic law, the Drucker-Prager-Cap theory, and a modified creep law. Furthermore, the volumetric inelastic strain dependent thermal conductivity of beryllium pebble beds is taken into account and full thermo-mechanical coupling is considered. Investigation showed that the Drucker-Prager-Cap model implemented in ABAQUS can not fulfill the requirements of both the prediction of large creep strains and the hardening behaviour caused by creep, which are of importance with respect to the application of pebble beds in fusion blankets. Therefore, UMAT (user defined material's mechanical behaviour) and UMATHT (user defined material's thermal behaviour) routines are used to re-implement the present thermo-mechanical model in ABAQUS. An elastic predictor radial return mapping algorithm is used to solve the non-associated plasticity iteratively, and a proper tangent stiffness matrix is obtained for cost-efficiency in the calculation. An explicit creep mechanism is adopted for the prediction of time-dependent behaviour in order to represent large creep strains in high temperature. Finally, the thermo-mechanical interactions are implemented in a UMATHT routine for the coupled analysis. The oedometric compression tests and creep tests of pebble beds at different temperatures are simulated with the help of the present UMAT and UMATHT routines, and the comparison between the simulation and the experiments is made. (authors)

  13. Economic and environmental performance of future fusion plants in comparison

    International Nuclear Information System (INIS)

    Hamacher, T.; Saez, R.M.; Lako, P.

    2001-01-01

    If the good performance of fusion as technology with no CO 2 emission during normal operation and rather low external costs, reflecting the advantageous environmental and safety characteristics, are considered in future energy regulations, fusion can win considerable market shares in future electricity markets. The economic performance was elaborated for Western Europe for the time period till 2100. The software tool MARKAL widely used in energy research was used to simulate and optimise the development of the Western European energy system. Two different scenarios were considered, the main difference was the interest rate for investments. Stringent CO 2 -emission strategies lead to considerable market shares for fusion. As a comprehensive indicator of the environmental and safety performance of fusion plants the external costs following the ExternE method was used. External costs of fusion are rather low, much below the cost of electricity, and are in the same range as photovoltaics and wind energy. (author)

  14. Program plan for the DOE Office of Fusion Energy First Wall/Blanket/Shield Engineering Technology Program. Volume I. Summary, objectives and management. Revision 2

    International Nuclear Information System (INIS)

    1982-08-01

    This document defines a plan for conducting selected aspects of the engineering testing required for magnetic fusion reactor FWBS components and systems. The ultimate product of this program is an established data base that contributes to a functional, reliable, maintainable, economically attractive, and environmentally acceptable commercial fusion reactor first wall, blanket, and shield system. This program plan updates the initial plan issued in November of 1980 by the DOE/Office of Fusion Energy (unnumbered report). The plan consists of two parts. Part I is a summary of activities, responsibilities and program management including reporting and interfaces with other programs. Part II is a compilation of the Detailed Technical Plans for Phase I (1982 to 1984) developed by the participants during Phase 0 of the program

  15. Thermal analysis of a helium-cooled, tube-bank blanket module for a tandem mirror fusion reactor

    International Nuclear Information System (INIS)

    Werner, R.W.

    1983-01-01

    A blanket module concept for the central cell of a tandem mirror reactor is described which takes advantage of the excellent heat transfer and low pressure drop characteristics of tube banks in cross-flow. The blanket employs solid Li 2 O as the tritium breeding material and helium as the coolant. The lithium oxide is contained in tubes arranged within the submodules as a two-pass, cross-flow heat exchanger. Primarily, the heat transfer and thermal-hydraulic aspects of the blanket design study are described in this paper. In particular, the analytical model used for selection of the best tube-bank design parameters is discussed in some detail

  16. Thermal analysis of a helium-cooled, tube-bank blanket module for a tandem-mirror fusion reactor

    International Nuclear Information System (INIS)

    Werner, R.W.; Hoffman, M.A.; Johnson, G.L.

    1983-01-01

    A blanket module concept for the central cell of a tandem mirror reactor is described which takes advantage of the excellent heat transfer and low pressure drop characteristics of tube banks in cross-flow. The blanket employs solid Li 2 O as the tritium breeding material and helium as the coolant. The lithium oxide is contained in tubes arranged within the submodules as a two-pass, cross-flow heat exchanger. Primarily, the heat transfer and thermal-hydraulic aspects of the blanket design study are described in this paper. In particular, the analytical model used for selection of the best tube-bank design parameters is discussed in some detail

  17. Evaluation on sweep gas pressure drop in fusion blanket mock-up for in-pile test

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi; Nagakura, Masaaki; Kanzawa, Toru.

    1993-03-01

    In the ITER/CDA (Conceptual Design Activity) of a tritium breeding blanket, Japan have proposed the pebble-typed blanket. The in-pile mock-up test will be preparing in JMTR (Japan Materials Testing Reactor) for Japanese engineering design with the pebble-typed blanket. Therefore, the He sweep gas pressure drop in the pebble bed was measured for the design of the mock-up used on in-pile test. From the results of this test, it was clear that the pressure drop was predicted on Kozeny- Carman's equation within +25 ∼ -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author)

  18. 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

  19. Study on the temperature control mechanism of the tritium breeding blanket for CFETR

    Science.gov (United States)

    Liu, Changle; Qiu, Yang; Zhang, Jie; Zhang, Jianzhong; Li, Lei; Yao, Damao; Li, Guoqiang; Gao, Xiang; Wu, Songtao; Wan, Yuanxi

    2017-12-01

    The Chinese fusion engineering testing reactor (CFETR) will demonstrate tritium self- sufficiency using a tritium breeding blanket for the tritium fuel cycle. The temperature control mechanism (TCM) involves the tritium production of the breeding blanket and has an impact on tritium self-sufficiency. In this letter, the CFETR tritium target is addressed according to its missions. TCM research on the neutronics and thermal hydraulics issues for the CFETR blanket is presented. The key concerns regarding the blanket design for tritium production under temperature field control are depicted. A systematic theory on the TCM is established based on a multiplier blanket model. In particular, a closed-loop method is developed for the mechanism with universal function solutions, which is employed in the CFETR blanket design activity for tritium production. A tritium accumulation phenomenon is found close to the coolant in the blanket interior, which has a very important impact on current blanket concepts using water coolant inside the blanket. In addition, an optimal tritium breeding ratio (TBR) method based on the TCM is proposed, combined with thermal hydraulics and finite element technology. Meanwhile, the energy gain factor is adopted to estimate neutron heat deposition, which is a key parameter relating to the blanket TBR calculations, considering the structural factors. This work will benefit breeding blanket engineering for the CFETR reactor in the future.

  20. 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

  1. JET:Preparing the future in fusion

    Czech Academy of Sciences Publication Activity Database

    Mlynář, Jan; Ongena, J.; Ďuran, Ivan; Hron, Martin; Pánek, Radomír; Petržílka, Václav; Žáček, František

    2004-01-01

    Roč. 54, suppl.C (2004), C28-C38 ISSN 0011-4626. [Symposium on Plasma Physics and Technology /21st/. Praha, 14.06.2004-17.06.2004] R&D Projects: GA ČR GA202/04/0360 Institutional research plan: CEZ:AV0Z2043910 Keywords : fusion, tokamak, JET EFDA, ITER Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.292, year: 2004

  2. Neutronics optimization of LiPb-He dual-cooled fuel breeding blanket for the fusion-driven sub-critical system

    International Nuclear Information System (INIS)

    Zheng Shanliang; Wu Yican

    2002-01-01

    The concept of the liquid Li 17 Pb 83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR > 1.05) and annual output of 100 kg or more fissile 239 Pu (FBR > 0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimized calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio (Br = Tbr + Fbr) is listed corresponding to different cases

  3. Summary report of the IAEA advisory group meeting on nuclear data for neutron multiplication in fusion-reactor first-wall and blanket materials

    International Nuclear Information System (INIS)

    Muir, D.W.; Pashchenko, A.B.

    1992-09-01

    The present Report contains the Summary of the IAEA Advisory Group Meeting on Nuclear Data for Neutron Multiplication in Fusion-Reactor First-Wall and Blanket Materials, which was hosted by the Southwest Institute of Nuclear physics and Chemistry (SWINPC) at Chengdu, China and held from 19-21 November 1990. This AGM was organized by the IAEA Nuclear Data Section (NDS), with the cooperation and assistance of local organizers at the SWINPC. The papers which the participants prepared for and presented at the meeting will be published as an INDC report. (author)

  4. Heavy-ion fusion: future promise and future directions

    International Nuclear Information System (INIS)

    Dudziak, D.J.; Saylor, W.W.; Pendergrass, J.H.

    1986-01-01

    The previous papers in this heavy-ion fusion special session have described work performed as part of the Heavy-Ion Fusion Systems Assessment (HIFSA) Project. Key technical issues in the design and costing of targets, induction linacs, beam transport, reactor, balance of plant, and systems integration have been identified and described. The HIFSA systems model was used to measure the relative value of improvements in physics understanding and technology developments in many different areas. Within the limits of our 1986 knowledge and imagination, this study defines the most attractive heavy-ion fusion (HIF) power plant concepts. The project has deliberately avoided narrowing the focus to a point facility design; thus, the generic systems modeling capability developed in the process allows for relative comparisons among design options. We will describe what are thought to be achievable breakthroughs and what the relative significance of the breakthroughs will be, although the specific mechanism for achieving some breakthroughs may not be clear at this point. This degree of optimism concerning such breakthroughs is probably at least as conservative as that used in other fusion assessments

  5. A preliminary study of a D-T tokamak fusion reactor with advanced blanket using the compact fusion advanced Brayton (CFAB) cycle

    International Nuclear Information System (INIS)

    Yoshikawa, K.; Ishikawa, M.; Umoto, J.; Fukuyama, A.; Mitarai, O.; Okamoto, M.; Sekimoto, H.; Nagatsu, M.

    1995-01-01

    Preliminary key issues for a synchrotron radiation-enhanced compact fusion advanced Brayton (CFAB) cycle fusion reactor similar to the CFAR (compact fusion advanced Rankine) cycle reactor are presented. These include plasma operation windows as a function of the first wall reflectivity and related issues, to estimate an allowance for deterioration of the first wall reflectivity due to dpa effects. It was found theoretically that first wall reflectivities down to 0.8 are still adequate for operation at an energy confinement scaling of 3 times Kaye-Goldston. Measurements of the graphite first wall reflectivities at Nagoya University indicate excellent reflectivities in excess of 90% for CC-312, PCC-2S, and PD-330S in the submillimeter regime, even at high temperatures in excess of 1000K. Some engineering issues inherent to the CFAB cycle are also discussed briefly in comparison with the CFAR cycle which uses hazardous limited-resource materials but is capable of using mercury as coolant for high heat removal. The CFAB cycle using helium coolant is found to achieve higher net plant conversion efficiencies in excess 60% using a non-equilibrium magnetohydrodynamic disk generator in the moderate pressure range, even at the cost of a relatively large pumping power, and at the penalty of high temperature materials, although excellent heat removal characteristics in the moderate pressure range need to be guaranteed in the future. (orig.)

  6. The near future of nuclear fusion

    International Nuclear Information System (INIS)

    Grenier, E.

    2000-01-01

    In this short article the author tries to find an explanation about the withdraw of the Usa from the ITER project. According to the author the American economy is not so flourishing as it seems, it is only the IT (information technologies) sector that is attracting huge foreign investments. Moreover state agencies such as Nasa have difficulties to recruit talented people because they are lured by easy money earned in the new economy. The whole of the ITER project has been reduced: the volume of plasma will be 840 m 3 (instead of 2000 m 3 ) for a fusion power of 500 MW (instead of 1500 MW). Canada is standing for hosting the building site. (A.C.)

  7. Sensitivity and uncertainty analysis for the tritium breeding ratio of a DEMO fusion reactor with a helium cooled pebble bed blanket

    Directory of Open Access Journals (Sweden)

    Nunnenmann Elena

    2017-01-01

    Full Text Available An uncertainty analysis was performed for the tritium breeding ratio (TBR of a fusion power plant of the European DEMO type using the MCSEN patch to the MCNP Monte Carlo code. The breeding blanket was of the type Helium Cooled Pebble Bed (HCPB, currently under development in the European Power Plant Physics and Technology (PPPT programme for a fusion power demonstration reactor (DEMO. A suitable 3D model of the DEMO reactor with HCPB blanket modules, as routinely used for blanket design calculations, was employed. The nuclear cross-section data were taken from the JEFF-3.2 data library. For the uncertainty analysis, the isotopes H-1, Li-6, Li-7, Be-9, O-16, Si-28, Si-29, Si-30, Cr-52, Fe-54, Fe-56, Ni-58, W-182, W-183, W-184 and W-186 were considered. The covariance data were taken from JEFF-3.2 where available. Otherwise a combination of FENDL-2.1 for Li-7, EFF-3 for Be-9 and JENDL-3.2 for O-16 were compared with data from TENDL-2014. Another comparison was performed with covariance data from JEFF-3.3T1. The analyses show an overall uncertainty of ± 3.2% for the TBR when using JEFF-3.2 covariance data with the mentioned additions. When using TENDL-2014 covariance data as replacement, the uncertainty increases to ± 8.6%. For JEFF-3.3T1 the uncertainty result is ± 5.6%. The uncertainty is dominated by O-16, Li-6 and Li-7 cross-sections.

  8. The role of fusion as a future power source

    International Nuclear Information System (INIS)

    Kintner, E.E.; Hirsch, R.L.

    1977-01-01

    potentials of fusion power in relation to nuclear fission, solar and other future energy sources can be assessed in general terms. The probability of success in fusion development, while not susceptible to measurement, continues to improve. Fusion can be expected to play an increasingly important role in energy supply world-wide in the early decades of the 21st century. If a commercial scale demonstration reactor (greater than or equal to 500 MWe) operates successfully by 2000, it is reasonable to anticipate as many as 20 to 100 large (1000 MWe) plants by 2020 and an increasing percentage of fusion electrical generating stations thereafter

  9. Evaluation on sweep gas pressure drop in fusion blanket mock-up for in-pile test

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Nagakura, Masaaki; Kanzawa, Toru.

    1993-03-01

    In the ITER/CDA (Conceptual Design Activity) of a tritium breeding blanket, Japan have proposed the pebble-typed blanket. The in-pile mock-up test will be preparing in JMTR (Japan Materials Testing Reactor) for Japanese engineering design with the pebble-typed blanket. Therefore, the He sweep gas pressure drop in the pebble bed was measured for the design of the mock-up used on in-pile test. From the results of this test, it was clear that the pressure drop was predicted on Kozeny- Carman's equation within +25 [approx] -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author).

  10. Evaluation on sweep gas pressure drop in fusion blanket mock-up for in-pile test

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Sagawa, Hisashi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Nagakura, Masaaki; Kanzawa, Toru

    1993-03-01

    In the ITER/CDA (Conceptual Design Activity) of a tritium breeding blanket, Japan have proposed the pebble-typed blanket. The in-pile mock-up test will be preparing in JMTR (Japan Materials Testing Reactor) for Japanese engineering design with the pebble-typed blanket. Therefore, the He sweep gas pressure drop in the pebble bed was measured for the design of the mock-up used on in-pile test. From the results of this test, it was clear that the pressure drop was predicted on Kozeny- Carman`s equation within +25 {approx} -60 %, and that the pressure drop was not affected by moisture concentration (< 100 ppm). (author).

  11. 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

  12. Use of high temperature superconductors for future fusion magnet systems

    Energy Technology Data Exchange (ETDEWEB)

    Fietz, W H [Forschungszentrum Karlsruhe, Institut fuer Technische Physik, Karlsruhe (Germany); Celentano, G; Della Corte, A [Superconductivity Division, ENEA - Frascati Research Center, Frascati (Italy); Goldacker, W; Heller, R; Komarek, P; Kotzyba, G; Nast, R; Obst, B; Schlachter, S I; Schmidt, C; Zahn, G [Forschungszentrum Karlsruhe, Institut fuer Technische Physik, Karlsruhe (Germany); Pasztor, G; Wesche, R [Centre de Recherches en Physique des Plasmas, Villingen (Switzerland); Salpietro, E; Vostner, A [European Fusion Development Agreement, Close Support Unit, Garching (Germany)

    2005-01-01

    With the construction of ITER the feasibility of a fusion machine will be demonstrated. To commercialize fusion it is essential to keep losses as small as possible in future fusion power plants. One major component where losses can be strongly reduced is the cooling system. For example in ITER where efficiency is not a major goal, a cooling power of 64 kW at 4.4 K is foreseen taking more than 20 MW electric power. Considering the size of future commercial fusion machines this consumption of electric power for cooling will even be higher. With a magnet system working at 20 K a fusion machine would work more efficient by a factor of 5-10 with respect to electric power consumption for cryogenics. Even better than that, would be a machine with a magnet system operating at 65 K to 77 K. In this case liquid nitrogen could be used as coolant saving money for investment and operation costs. Such an increase in the operating temperature of the magnet system can be achieved by the use of High- Temperature Superconductors (HTS). In addition the use of HTS would allow much smaller efforts for thermal shielding and alternative thermal insulation concepts may be possible, e.g. for an HTS bus bar system. This contribution will give an overview about status, promises and challenges of HTS conductors on the way to an HTS fusion magnet system beyond ITER. (author)

  13. New design of cable-in-conduit conductor for application in future fusion reactors

    Science.gov (United States)

    Qin, Jinggang; Wu, Yu; Li, Jiangang; Liu, Fang; Dai, Chao; Shi, Yi; Liu, Huajun; Mao, Zhehua; Nijhuis, Arend; Zhou, Chao; Yagotintsev, Konstantin A.; Lubkemann, Ruben; Anvar, V. A.; Devred, Arnaud

    2017-11-01

    The China Fusion Engineering Test Reactor (CFETR) is a new tokamak device whose magnet system includes toroidal field, central solenoid (CS) and poloidal field coils. The main goal is to build a fusion engineering tokamak reactor with about 1 GW fusion power and self-sufficiency by blanket. In order to reach this high performance, the magnet field target is 15 T. However, the huge electromagnetic load caused by high field and current is a threat for conductor degradation under cycling. The conductor with a short-twist-pitch (STP) design has large stiffness, which enables a significant performance improvement in view of load and thermal cycling. But the conductor with STP design has a remarkable disadvantage: it can easily cause severe strand indentation during cabling. The indentation can reduce the strand performance, especially under high load cycling. In order to overcome this disadvantage, a new design is proposed. The main characteristic of this new design is an updated layout in the triplet. The triplet is made of two Nb3Sn strands and one soft copper strand. The twist pitch of the two Nb3Sn strands is large and cabled first. The copper strand is then wound around the two superconducting strands (CWS) with a shorter twist pitch. The following cable stages layout and twist pitches are similar to the ITER CS conductor with STP design. One short conductor sample with a similar scale to the ITER CS was manufactured and tested with the Twente Cable Press to investigate the mechanical properties, AC loss and internal inspection by destructive examination. The results are compared to the STP conductor (ITER CS and CFETR CSMC) tests. The results show that the new conductor design has similar stiffness, but much lower strand indentation than the STP design. The new design shows potential for application in future fusion reactors.

  14. Role of nuclear fusion in future energy systems and the environment under future uncertainties

    International Nuclear Information System (INIS)

    Tokimatsu, Koji; Fujino, Jun'ichi; Konishi, Satoshi; Ogawa, Yuichi; Yamaji, Kenji

    2003-01-01

    Debates about whether or not to invest heavily in nuclear fusion as a future innovative energy option have been made within the context of energy technology development strategies. This is because the prospects for nuclear fusion are quite uncertain and the investments therefore carry the risk of quite large regrets, even though investment is needed in order to develop the technology. The timeframe by which nuclear fusion could become competitive in the energy market has not been adequately studied, nor has roles of the nuclear fusion in energy systems and the environment. The present study has two objectives. One is to reveal the conditions under which nuclear fusion could be introduced economically (hereafter, we refer to such introductory conditions as breakeven prices) in future energy systems. The other objective is to evaluate the future roles of nuclear fusion in energy systems and in the environment. Here we identify three roles that nuclear fusion will take on when breakeven prices are achieved: (i) a portion of the electricity market in 2100, (ii) reduction of annual global total energy systems cost, and (iii) mitigation of carbon tax (shadow price of carbon) under CO 2 constraints. Future uncertainties are key issues in evaluating nuclear fusion. Here we treated the following uncertainties: energy demand scenarios, introduction timeframe for nuclear fusion, capacity projections of nuclear fusion, CO 2 target in 2100, capacity utilization ratio of options in energy/environment technologies, and utility discount rates. From our investigations, we conclude that the presently designed nuclear fusion reactors may be ready for economical introduction into energy systems beginning around 2050-2060, and we can confirm that the favorable introduction of the reactors would reduce both the annual energy systems cost and the carbon tax (the shadow price of carbon) under a CO 2 concentration constraint

  15. 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

  16. Welding and cutting characteristics of blanket/first wall module to back plate for fusion experimental reactor

    International Nuclear Information System (INIS)

    Sato, Shinichi; Osaki, Toshio; Koga, Shinji

    1996-01-01

    The first wall and the blanket of the International Thermonuclear Experimental Reactor (ITER) are used under severe conditions such as the neutron irradiation by plasma, surface thermal load, the electromagnetic force at the time of plasma disruption and others. Consequently, from the viewpoint of the necessity for disassembling and maintenance, those are divided into modules in toroidal and poloidal directions. In this study, as to the welding of the back plate and the legs supporting blanket modules, which are installed in a vacuum vessel, the characteristic test paying attention to the deformation at the time of welding was carried out, and the optimal welding conditions and the characteristics of welding deformation and others were clarified. Moreover, when water jet method was used for cutting the welded parts of the supporting legs, the properties of the cut parts, the time for cutting and others were examined. The performance required for the welded parts of blanket modules with back plate is shown. The basic test of welding conditions using plate models, partial model test and whole model test are reported. The test of water jet cutting for the maintenance of shielding blanket modules is described. (K.I.)

  17. A computational procedure for coupled electromagnetic-structural dynamic problems and its application to a fusion reactor blanket

    International Nuclear Information System (INIS)

    Jordan, T.

    1993-09-01

    A method is presented in order to couple quasistationary electromagnetics and the dynamics of structure and fluid. This method allows to compute forces, strains and stresses in structures subjected to transient magnetic fields. An important application was to determine the dynamic loading of the self-cooled liquid metal blanket during a plasma distruption. (orig./HP) [de

  18. 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.)

  19. Progress of and future plans for the L-4 Blanket Project

    International Nuclear Information System (INIS)

    Daenner, W.; Ioki, K.; Cardella, A.

    2001-01-01

    The ITER L-4 Blanket Project has achieved substantial progress over the last two years. The qualification of materials so far considered as reference for the shield module fabrication has been completed, as well as the developments for joining the triplex First Wall structure. Several Primary Wall, baffle, and limiter mock-ups have been manufactured and tested showing comfortable margins against the loads expected in ITER. Shield prototypes have been manufactured by conventional and advanced technology, which have finally demonstrated the manufacturing feasibility. More recently, activities for the qualification of the module attachment system have been started, and first results from materials and mock-up tests have become available. Several test campaigns are still to be finished to complete the data base for the design. In the meantime, further activities have been initiated to adapt the R and D programme to the ITER-FEAT design features, with the aim to further reduce the cost. (author)

  20. Fusion power in a future low carbon global electricity system

    DEFF Research Database (Denmark)

    Cabal, H.; Lechón, Y.; Bustreo, C.

    2017-01-01

    Fusion is one of the technologies that may contribute to a future, low carbon, global energy supply system. In this article we investigate the role that it may play under different scenarios. The global energy model ETM (originally EFDA TIMES Model) has been used to analyse the participation...

  1. 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.

  2. The fusion reactor wall is getting hot. A challenge towards the future for numerical modelling (4). Chap. 4. What is really happening in the wall?

    International Nuclear Information System (INIS)

    Murata, Isao; Konno, Chikara

    2008-01-01

    In fusion plasmas, a lot of fast neutrons with a kinetic energy of 14 MeV are generated through D-T fusion reactions. These neutrons travel deep into the first wall and are absorbed in the blanket through nuclear reactions. In the present chapter, the authors discuss what happens in the blanket with the help of computerized simulation. (T.I.)

  3. 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.)

  4. Progress of nuclear fusion research and review on development of fusion reactors

    International Nuclear Information System (INIS)

    1976-01-01

    Set up in October 1971, the ad hoc Committee on Survey of Nuclear Fusion Reactors has worked on overall fusion reactor aspects and definition of the future problems under four working groups of core, nuclear heat, materials and system. The presect volume is intended to provide reference materials in the field of fusion reactor engineering, prepared by members of the committee. Contents are broadly the following: concept of the nuclear fusion reactor, fusion core engineering, fusion reactor blanket engineering, fusion reactor materials engineering, and system problems in development of fusion reactors. (Mori, K.)

  5. 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

  6. Personnel Safety for Future Magnetic Fusion Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee Cadwallader

    2009-07-01

    The safety of personnel at existing fusion experiments is an important concern that requires diligence. Looking to the future, fusion experiments will continue to increase in power and operating time until steady state power plants are achieved; this causes increased concern for personnel safety. This paper addresses four important aspects of personnel safety in the present and extrapolates these aspects to future power plants. The four aspects are personnel exposure to ionizing radiation, chemicals, magnetic fields, and radiofrequency (RF) energy. Ionizing radiation safety is treated well for present and near-term experiments by the use of proven techniques from other nuclear endeavors. There is documentation that suggests decreasing the annual ionizing radiation exposure limits that have remained constant for several decades. Many chemicals are used in fusion research, for parts cleaning, as use as coolants, cooling water cleanliness control, lubrication, and other needs. In present fusion experiments, a typical chemical laboratory safety program, such as those instituted in most industrialized countries, is effective in protecting personnel from chemical exposures. As fusion facilities grow in complexity, the chemical safety program must transition from a laboratory scale to an industrial scale program that addresses chemical use in larger quantity. It is also noted that allowable chemical exposure concentrations for workers have decreased over time and, in some cases, now pose more stringent exposure limits than those for ionizing radiation. Allowable chemical exposure concentrations have been the fastest changing occupational exposure values in the last thirty years. The trend of more restrictive chemical exposure regulations is expected to continue into the future. Other issues of safety importance are magnetic field exposure and RF energy exposure. Magnetic field exposure limits are consensus values adopted as best practices for worker safety; a typical

  7. Personnel Safety for Future Magnetic Fusion Power Plants

    International Nuclear Information System (INIS)

    Cadwallader, Lee

    2009-01-01

    The safety of personnel at existing fusion experiments is an important concern that requires diligence. Looking to the future, fusion experiments will continue to increase in power and operating time until steady state power plants are achieved; this causes increased concern for personnel safety. This paper addresses four important aspects of personnel safety in the present and extrapolates these aspects to future power plants. The four aspects are personnel exposure to ionizing radiation, chemicals, magnetic fields, and radiofrequency (RF) energy. Ionizing radiation safety is treated well for present and near-term experiments by the use of proven techniques from other nuclear endeavors. There is documentation that suggests decreasing the annual ionizing radiation exposure limits that have remained constant for several decades. Many chemicals are used in fusion research, for parts cleaning, as use as coolants, cooling water cleanliness control, lubrication, and other needs. In present fusion experiments, a typical chemical laboratory safety program, such as those instituted in most industrialized countries, is effective in protecting personnel from chemical exposures. As fusion facilities grow in complexity, the chemical safety program must transition from a laboratory scale to an industrial scale program that addresses chemical use in larger quantity. It is also noted that allowable chemical exposure concentrations for workers have decreased over time and, in some cases, now pose more stringent exposure limits than those for ionizing radiation. Allowable chemical exposure concentrations have been the fastest changing occupational exposure values in the last thirty years. The trend of more restrictive chemical exposure regulations is expected to continue into the future. Other issues of safety importance are magnetic field exposure and RF energy exposure. Magnetic field exposure limits are consensus values adopted as best practices for worker safety; a typical

  8. Introducing the book 'Cold fusion and the future'

    International Nuclear Information System (INIS)

    Rothwell, Jed

    2006-01-01

    Cold fusion will be the ideal source of energy, provided its introduction can be handled properly. A few cells have shown power density and temperatures suitable for real-world applications. Once these cells can be replicated on demand, commercial development will be straightforward. Manufacturing should not be too demanding, so thousands of companies will compete, and costs will fall quickly. The transition from fossil fuel to cold fusion will be rapid. Many extraordinary new applications will become possible, and seemingly intractable problems such as global warming may be fixed. Some examples will be presented. Public support is essential to funding research, and commercialization. 'Cold Fusion and the Future' is the title of a new book by this author. This paper discusses a few of the topics in the book

  9. Alternate applications of fusion power: development of a high-temperature blanket for synthetic-fuel production

    International Nuclear Information System (INIS)

    Howard, P.A.; Mattas, R.F.; Krajcinovic, D.; DePaz, J.; Gohar, Y.

    1981-11-01

    This study has shown that utilization of the unique features of a fusion reactor can result in a novel and potentially economical method of decomposing steam into hydrogen and oxygen. Most of the power of fusion reactors is in the form of energetic neutrons. If this power could be used to produce high temperature uncontaminated steam, a large fraction of the energy needed to decomposee the steam could be supplied as thermal energy by the fusion reaction. Proposed high temperature electrolysis processes require steam temperature in excess of 1000 0 C for high efficiency. The design put forth in this study details a system that can accomplish that end

  10. 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.

  11. Monte Carlo analysis of the effects of a blanket-shield penetration on the performance of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Tang, J.S.; Alsmiller, R.G. Jr.; Barnes, J.M.

    1977-05-01

    Adjoint Monte Carlo calculations have been carried out using the three-dimensional radiation transport code, MORSE, to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm 2 rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections obtained by collapsing the DLC-37 cross-section library. Nuclear heating and radiation damage rates were estimated using the latest available nuclear response functions. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196 depending on the location. Increases in the radiation damage also result in the TF coils. The atomic displacement rates increase from factors of 2 to 138 and the hydrogen and helium gas production rates increase from factors of 11 to 7600 and from 15 to 9700, respectively

  12. 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

  13. 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

  14. 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

  15. IEC fusion: The future power and propulsion system for space

    International Nuclear Information System (INIS)

    Hammond, Walter E.; Coventry, Matt; Miley, George H.; Nadler, Jon; Hanson, John; Hrbud, Ivana

    2000-01-01

    Rapid access to any point in the solar system requires advanced propulsion concepts that will provide extremely high specific impulse, low specific power, and a high thrust-to-power ratio. Inertial Electrostatic Confinement (IEC) fusion is one of many exciting concepts emerging through propulsion and power research in laboratories across the nation which will determine the future direction of space exploration. This is part of a series of papers that discuss different applications of the Inertial Electrostatic Confinement (IEC) fusion concept for both in-space and terrestrial use. IEC will enable tremendous advances in faster travel times within the solar system. The technology is currently under investigation for proof of concept and transitioning into the first prototype units for commercial applications. In addition to use in propulsion for space applications, terrestrial applications include desalinization plants, high energy neutron sources for radioisotope generation, high flux sources for medical applications, proton sources for specialized medical applications, and tritium production

  16. System code assessment with thermal-hydraulic experiment to develop helium cooled breeding blanket for nuclear fusion reactor

    International Nuclear Information System (INIS)

    Yum, S. B.; Park, I. W.; Park, G. C.; Lee, D. W.

    2012-01-01

    By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He Cooled Molten Lithium (HCML) Test Blanket Module (TBM) for testing in the International Thermonuclear Experimental Reactor (ITER). A performance analysis for the thermal-hydraulics and a safety analysis for an accident caused by a loss of coolant for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs Multicomponent Mixture Analysis), which was developed by the Gas Cooled Reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM First Wall (FW) mock-up made from the same material as tho KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 11, 19, and 29 bar, and under various ranges of flow rate from 0.63 to 2.44kg/min with a constant wall temperature condition. In the present study, a thermal-hydraulic test was performed with the newly constructed helium supplying system, In which the design pressure and temperature were 9 MPa and 500 .deg. C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 8 MPa pressure, 0.2 kg/s flow rate, which are almost same conditions of the KO TBM FW. One-side of the mock-up was heated with a constant heat flux of 0.5 MW/m 2 using a graphite heating system, KoHLT-2 (Korea Heat Load Test Facility-2). The wall temperatures were measured using installed thermocouples, and they show a strong parity with the code results simulated under the same test conditions

  17. Status and strategy of fusion materials development in China

    International Nuclear Information System (INIS)

    Huang, Q.Y.; Wu, Y.C.; Li, J.G.; Wan, F.R.; Chen, J.L.; Luo, G.N.; Liu, X.; Chen, J.M.; Xu, Z.Y.; Zhou, X.G.; Ju, X.; Shan, Y.Y.; Yu, J.N.; Zhu, S.Y.; Zhang, P.Y.; Yang, J.F.; Chen, X.J.; Dong, S.M.

    2009-01-01

    The liquid metal and solid ceramic pebble breeder blankets have become the most promising blankets for ITER-TBMs or DEMO reactors in China and the world due to their potential advantages. In recent years the corresponding research work on fusion reactor materials mainly focuses on structural materials, plasma facing materials and the functional materials for the blanket such as breeder, coating and flow channel insert etc. for the successful application of fusion energy in the near future. The R and D on those materials in the two kinds of blankets is being carried out widely in China, including fabrication and manufacturing techniques, physical/mechanical properties assessment before and after irradiation, joining techniques for structural materials, compatibility evaluation, and the development and verification of the criteria for fusion material designs. The progress on main R and D activities of fusion reactor materials in China is introduced and prospected in the paper.

  18. Fusion-fission dynamics and perspectives of future experiments

    International Nuclear Information System (INIS)

    Zagrebaev, V.I.; Itkis, M.G.; Oganessian, Yu.Ts.

    2003-01-01

    The paper is focused on reaction dynamics of superheavy-nucleus formation and decay at beam energies near the Coulomb barrier. The aim is to review the things we have learned from recent experiments on fusion-fission reactions leading to the formation of compound nuclei with Z ≥ 102 and from their extensive theoretical analysis. Major attention is paid to the dynamics of formation of very heavy compound nuclei taking place in strong competition with the process of fast fission (quasifission). The choice of collective degrees of freedom playing a fundamental role and finding the multidimensional driving potential and the corresponding dynamic equation regulating the whole process are discussed. A possibility of deriving the fission barriers of superheavy nuclei directly from performed experiments is of particular interest here. In conclusion, the results of a detailed theoretical analysis of available experimental data on the 'cold' and 'hot' fusion-fission reactions are presented. Perspectives of future experiments are discussed along with additional theoretical studies in this field needed for deeper understanding of the fusion-fission processes of very heavy nuclear systems

  19. 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

  20. 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.

  1. Thermal Hydraulic Analysis of K-DEMO Single Blanket Module for Preliminary Accident Analysis using MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

  2. Quantification of design margins and safety factors based on the prediction uncertainty in tritium production rate from fusion integral experiments of the USDOE/JAERI collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Youssef, M.Z.; Konno, C.; Maekawa, F.; Ikeda, Y.; Kosako, K.; Nakagawa, M.; Mori, T.; Maekawa, H.

    1995-01-01

    Several fusion integral experiments were performed within a collaboration between the USA and Japan on fusion breeder neutronics aimed at verifying the prediction accuracy of key neutronics parameters in a fusion reactor blanket based on current neutron transport codes and basic nuclear databases. The focus has been on the tritium production rate (TRP) as an important design parameter to resolve the issue of tritium self-sufficiency in a fusion reactor. In this paper, the calculational and experimental uncertainties (errors) in local TPR in each experiment performed i were interpolated and propagated to estimate the prediction uncertainty u i in the line-integrated TPR and its standard deviation σ i . The measured data are based on Li-glass and NE213 detectors. From the quantities u i and σ i , normalized density functions (NDFs) were constructed, considering all the experiments and their associated analyses performed independently by the UCLA and JAERI. Several statistical parameters were derived, including the mean prediction uncertainties u and the possible spread ±σ u around them. Design margins and safety factors were derived from these NDFs. Distinction was made between the results obtained by UCLA and JAERI and between calculational results based on the discrete ordinates and Monte Carlo methods. The prediction uncertainties, their standard deviations and the design margins and safety factors were derived for the line-integrated TPR from Li-6 T 6 , and Li-7 T 7 . These parameters were used to estimate the corresponding uncertainties and safety factor for the line-integrated TPR from natural lithium T n . (orig.)

  3. Nuclear fusion, an energy source of the future

    International Nuclear Information System (INIS)

    Koeppendoerfer, W.

    1994-01-01

    The paper discusses the possibility to obtain energy by nuclear fusion. It deals successively with: The physical bases of nuclear fusion, research and development with a view to harnessing nuclear fusion, properties of a fusion reactor, and programme and timetable to economic exploitation. (orig./UA) [de

  4. Conceptual Design of Main Cooling System for a Fusion Power Reactor with Water Cooled Lithium-Lead Blanket. TW1-TRP-PPCS1, Deliverable 8

    International Nuclear Information System (INIS)

    Natalizio, Antonio; Collen, Jan

    2002-06-01

    The HTS (Heat Transfer System) conceptual design developed for the PPCS (Power-Plant Conceptual Study) plant model is compliant with the single failure criterion - i.e., the failure of a single active component (e.g., pump) will not cause the reactor to shutdown. The system effective availability (capacity factor), however, is only marginally better than that of the SEAFP design, as the number of loops could not be decreased further, due to coolant inventory limitations. The PPCS Plant Model A has about 70 % more fusion power than the SEAFP model. Therefore, keeping the same number of loops as in the SEAFP model would have implied a 70 % larger inventory. To improve plant availability and safety, however, the number of blanket and first wall loops have been reduced from eight to six, implying a further increase in loop inventory of about 25 %. For these and other reasons, the coolant inventory, at risk from a loss-of-coolant accident, has increased significantly, relative to the SEAFP design (∼130 vs. 50 m 3 ). The proposed heat transport system conceptual design meets, or exceeds, all project specifications

  5. EXOTIC: Development of ceramic tritium breeding materials for fusion reactor blankets. The behaviour of tritium in: lithium aluminate, lithium oxide, lithium silicates, lithium zirconates

    Energy Technology Data Exchange (ETDEWEB)

    Kwast, H [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Stijkel, H [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Muis, R [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Conrad, R [Commission of the European Communities, Petten (Netherlands). Joint Reseach Centre

    1995-12-01

    This report describes the results of six EXOTIC experiments comprising a total of 48 capsules. Samples of the candidate tritium breeding materials LiAlO{sub 2}, Li{sub 2}ZrO{sub 3}, Li{sub 4}SiO{sub 4}, Li{sub 6}Zr{sub 2}O{sub 7}, Li{sub 8}ZrO{sub 6}, Li{sub 2}O and Li{sub 2}SiO{sub 3} have been irradiated at different temperature levels and up to a maximum lithium burnup of about 3%. Tritium residence times of the various breeding materials have been determined from temperature transients performed during irradiation. After irradiation the tritium inventory has been determined from small samples of the various materials. From the out-of-pile tritium release experiments activation energies were determined. These activities have been performed at ECN within the framework of the European Fusion Technology Programme on Breeding Blankets. (orig.).

  6. Flow characteristics analysis of purge gas in unitary pebble beds by CFD simulation coupled with DEM geometry model for fusion blanket

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Youhua [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Chen, Lei [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); Luo, Guangnan [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

    2017-01-15

    Highlights: • A unitary pebble bed was built to analyze the flow characteristics of purge gas based on DEM-CFD method. • Flow characteristics between particles were clearly displayed. • Porosity distribution, velocity field distribution, pressure field distribution, pressure drop and the wall effects on velocity distribution were studied. - Abstract: Helium is used as the purge gas to sweep tritium out when it flows through the lithium ceramic and beryllium pebble beds in solid breeder blanket for fusion reactor. The flow characteristics of the purge gas will dominate the tritium sweep capability and tritium recovery system design. In this paper, a computational model for the unitary pebble bed was conducted using DEM-CFD method to study the purge gas flow characteristics in the bed, which include porosity distribution between pebbles, velocity field distribution, pressure field distribution, pressure drop as well as the wall effects on velocity distribution. Pebble bed porosity and velocity distribution with great fluctuations were found in the near-wall region and detailed flow characteristics between pebbles were displayed clearly. The results show that the numerical simulation model has an error with about 11% for estimating pressure drop when compared with the Ergun equation.

  7. Challenges and the future of the fusion energy

    International Nuclear Information System (INIS)

    Gross, R.A.

    1982-01-01

    The need to develop new large energy resources is discussed. One of three inexhaustible energy resource possibilities is fusion energy, whose history and scientific goals are described. The current world-wide research and development program for fusion is outlined. As an example of today's perception of what fusion energy will be like, a commercial tokamak fusion electric powerplant is described. Special attention is devoted to some of the challenging material problems that face fusion power development. (Author) [pt

  8. Calculation of design uncertainties for the development of fusion reactor blankets, taking into account uncertainties in nuclear data

    International Nuclear Information System (INIS)

    Hogenbirk, A.

    1994-07-01

    The use is demonstrated of the newly developed ECN-SUSD sensitivity/uncertainty code system. With ECN-SUSD it is possible to calculate uncertainties in response parameters in fixed source calculations due to cross section uncertainties (using MF33) as well as to uncertainties in angular distributions (using MF34). It is shown that the latter contribution, which is generally neglected because of the lack of MF34-data in modern evaluations (except for EFF), is large in fusion reactor shielding calculations. (orig.)

  9. Intelligible seminar on fusion reactors. (12) Next step toward the realization of fusion reactors. Future vision of fusion energy research and development

    International Nuclear Information System (INIS)

    Okano, Kunihiko; Kurihara, Kenichi; Tobita, Kenji

    2006-01-01

    In the last session of this seminar the progress of research and development for the realization of fusion reactors and future vision of fusion energy research and development are summarized. The some problems to be solved when the commercial fusion reactors would be realized, (1) production of deuterium as the fuel, (2) why need the thermonuclear reactors, (3) environmental problems, and (4) ITER project, are described. (H. Mase)

  10. Implications of the second law for future directions in controlled fusion research

    International Nuclear Information System (INIS)

    Roth, J.R.; Miley, G.H.

    1980-01-01

    Many existing energy related technologies have developed under the influence of social, economic, or state of the art constraints, and they cannot be viewed as optimum systems according to the second law of thermodynamics. Controlled fusion research presents an opportunity to optimize a nascent technology with respect to second law considerations in order to develop a practical energy source. In its present state of development, fusion research offers several independent approaches that may result in a net power producing fusion reactor. This paper discusses how second law considerations might be used to narrow the range of choices that must be made among various fusion fuel cycles. From a second law point of view, the most desirable fusion reactors are those for which the energy of charged particles can be converted directly into d.c. electrical power, while still allowing the energy that could be recovered by an efficient high-temperature 'blanket' to be transported largely by radiation. Fusion research in all major industrialized countries is developing the deuterium-tritium (D-T) fuel cycle for first-generation fusion power plants. It will be shown that other fuel cycles have significant advantages over the D-T fuel cycle according to second law principles. (author)

  11. Fusion power system: technology and engineering considerations

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1976-01-01

    Engineering concepts are discussed for the following topics: (1) blanket environment, (2) blanket materials, (3) tritium breeding, (4) heat removal problems, (5) materials selection for radiation shields, (6) afterheat, and (7) fusion blanket design

  12. Divertor development for a future fusion power plant

    International Nuclear Information System (INIS)

    Norajitra, Prachai

    2011-01-01

    Nuclear fusion is considered as a future source of sustainable energy supply. In the first chapter, the physical principle of magnetic plasma confinement, and the function of a tokamak are described. Since the discovery of the H-mode in ASDEX experiment ''Divertor I'' in 1982, the divertor has been an integral part of all modern tokamaks and stellarators, not least the ITER machine. The goal of this work is to develop a feasible divertor design for a fusion power plant to be built after ITER. This task is particularly challenging because a fusion power plant formulates much greater demands on the structural material and the design than ITER in terms of neutron wall load and radiation. First several divertor concepts proposed in the literature e.g. the Power Plant Conceptual Study (PPCS) using different coolants are reviewed and analyzed with respect to their performance. As a result helium cooled divertor concept exhibited the best potential to come up to the highest safety requirements and therefore has been chosen for the design process. From the third chapter the necessary steps towards this goal are described. First, the boundary conditions for the arrangement of a divertor with respect to the fusion plasma are discussed, as this determines the main thermal and neutronic load parameters. Based on the loads material selection criteria are inherently formulated. In the next step, the reference design is defined in accordance with the established functional design specifications. The developed concept is of modular nature and consists of cooling fingers of tungsten using an impingement cooling in order to achieve a heat dissipation of 10 MW/m 2 . In the next step, the design was subjected to the thermal-hydraulic and thermo-mechanical calculations in order to analyze and improve the performance and the manufacturing technologies. Based on these results, a prototype was produced and experimentally tested on their cooling capacity, their thermo-cyclic loading

  13. Peaceful fusion

    Energy Technology Data Exchange (ETDEWEB)

    Englert, Matthias [IANUS, TU Darmstadt (Germany)

    2014-07-01

    Like other intense neutron sources fusion reactors have in principle a potential to be used for military purposes. Although the use of fissile material is usually not considered when thinking of fusion reactors (except in fusion-fission hybrid concepts) quantitative estimates about the possible production potential of future commercial fusion reactor concepts show that significant amounts of weapon grade fissile materials could be produced even with very limited amounts of source materials. In this talk detailed burnup calculations with VESTA and MCMATH using an MCNP model of the PPCS-A will be presented. We compare different irradiation positions and the isotopic vectors of the plutonium bred in different blankets of the reactor wall with the liquid lead-lithium alloy replaced by uranium. The technical, regulatory and policy challenges to manage the proliferation risks of fusion power will be addressed as well. Some of these challenges would benefit if addressed at an early stage of the research and development process. Hence, research on fusion reactor safeguards should start as early as possible and accompany the current research on experimental fusion reactors.

  14. Vapor condensation on the surface of a liquid blanket jet in an inertial-confinement fusion reactor

    International Nuclear Information System (INIS)

    Takahashi, Minoru; Inoue, Akira; Fujinuma, Hajime; Tsukui, Jun.

    1991-01-01

    As the fundamental study on lithium jet cooling of an inertial-confinement fusion reactor, the experiment was performed to investigate for the steady condensation of saturated steam on a vertical downward water jet. The experimental parameters were the nozzle diameter of 3 and 5 mm, the jet length of 60∼316 mm, the outlet velocity of 2∼12 m/s, the outlet temperature of 30∼70degC, and the pressure of 0.03∼0.44 MPa, which corresponds to the Reynolds number of 1.35 x 10 4 ∼2.71 x 10 5 and the Prandtl number of 1.0∼5.2. As the Reynolds number or the jet length is increased, the Stanton number decreases and then increases again. As the steam pressure is increased, it increases monotonously. These characteristics of condensation heat transfer have been classical into four regions based on the criteria for jet break-up and surface disturbance, or entrainment. The empirical correlations for the Stanton number have been obtained for these regions, and the validity was confirmed by comparing them with the previous correlations. (author)

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

    International Nuclear Information System (INIS)

    Alvani, Carlo; Casadio, Sergio; Contini, Vittoria; Giorgi, Rossella; Mancini, Maria Rita; Tsuchiya, Kunihiko; Kawamura, Hiroshi

    2005-07-01

    This is a report of the Working Group in the Subtask on Solid Breeder Blankets under the Implementing Agreement on a Co-operative Programme on Nuclear Technology of Fusion Reactors (International Energy Agency (IEA)). This Working Group (Task F and WG-F) was performed from 2000 to 2004 by a collaboration of European Union (EU) and Japan (JA). In this report, lithium depletion effects on the reaction of lithium titanate (Li 2 TiO 3 ) with hydrogen (H 2 ) in thermo-chemical environment were discussed. The reaction of Li 2 TiO 3 ceramics with H 2 was studied in a thermo-chemical environment simulating (excepting irradiation) that of the hottest pebble-bed zone of breeding-blanket actually designed for fusion power plants. This 'reduction' as performed at 900degC in Ar+0.1%H, purge gas (He+0.1%H 2 being the designed reference') was found to be enhanced by TiO 2 doping of the specimens of simulate 6 Li-burn-up expected to reach 20% at their end-of-life. The reaction rates, however, were so slow to be not significantly extrapolated to the breeder material service time (years). In Ar+3%H 2 , faster reaction rates allowed a better identification of the process evolution (kinetics) by Temperature-Programmed Reduction' (TPR) and 'Oxidation' (TPO), and combined TG-DTA thermal analysis. The reduction of pure Li 4/5 TiO 12/5 spinel phase to Li 4/5 TiO 12/5-y was found to reach in one day the steady state at the O-vacancy concentration y=0.2. Complimentary microscopy (SEM) and spectroscopy (XRD, XPS) techniques were used to characterize the reaction products among which the presence of the orthorhombic Li v TiO 2 (0 ≤ v ≤ 1/2) and Li 2 TiO 3 could be diagnosed. So that the complete spinel reduction to Li 1/2 TiO 2 was obtained according to a scheme involving the Li 1/2 TiO 2 -Li 4/5 TiO 12/5 spinel phase solid solution for which y=3v/(10-5v). The reduction rate of pure meta-titanate to Li 2 TiO 3-x was found much lower (x approx. = 0.01) and even possibly due to the presence

  16. Material Challenges For Plasma Facing Components in Future Fusion Reactors

    International Nuclear Information System (INIS)

    Linke, J; Pintsuk, G.; Rödig, M.

    2013-01-01

    Increasing attention is directed towards thermonuclear fusion as a possible future energy source. Major advantages of this energy conversion technology are the almost inexhaustible resources and the option to produce energy without CO2-emissions. However, in the most advanced field of magnetic plasma confinement a number of technological challenges have to be met. In particular high-temperature resistant and plasma compatible materials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PFMs) and components (PFCs) in such fusion devices, i.e. the first wall (FW), the limiters and the divertor, are strongly affected by the plasma wall interaction processes and the applied intense thermal loads during plasma operation. On the one hand, these mechanisms have a strong influence on the plasma performance; on the other hand, they have major impact on the lifetime of the plasma facing armour. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm-2; the limiters and the divertor are expected to be exposed to power densities being at least one order of magnitude above the FW-level, i.e. up to 20 MWm-2 for next step tokamaks such as ITER or DEMO. These requirements are responsible for high demands on the selection of qualified PFMs and heat sink materials as well as reliable fabrication processes for actively cooled plasma facing components. The technical solutions which are considered today are mainly based on the PFMs beryllium, carbon or tungsten joined to copper alloys or stainless steel heat sinks. In addition to the above mentioned quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJm-2 are a serious concern for next step tokamak devices. The most frequent events are so-called Edge Localized Modes (type I ELMs) and plasma disruptions

  17. Liquid-metal MHD flow in a duct whose cross section changes from a rectangle to a trapezoid, with applications in fusion blanket designs

    International Nuclear Information System (INIS)

    Walker, J.S.

    1986-04-01

    This paper treats the liquid-metal MHD flow in a semi-infinite rectangular duct and a semi-infinite trapezoidal duct, which are connected by a finite-length transition duct. There is a strong, transverse, uniform magnetic field. The walls parallel to the magnetic field (sides) remain parallel, while the walls intersecting the magnetic field are twisted in the transition duct to provide the change in cross sectional shape. The left side has a constant height, while the height of the right side increases or decreases in the transition duct. This geometry gives a skewed velocity profile with a high velocity near the left side, provided the right side is relatively thick. All walls are thin and electrically conducting, but the sides are considerably thicker than the other walls. The application is to fusion-reactor blankets in which a high velocity near the first wall (separating the plasma chamber from the coolant) improves the thermal performance. Junctions of different ducts with walls parallel to the magnetic field are treated for the first time. In expansions, contractions and other geometric transition ducts, as well as in straight ducts with axially varying magnetic fields, the fluid flow and electric currents are concentrated in boundary layers adjacent to the sides and in the side. At a junction with a straight duct with a uniform magnetic field, the flow and current must transfer from the boundary layers adn sides to the core regions. These transfers at junctions play a key role in any three-dimensional flow

  18. Experimental study of gaseous lithium deuterides and lithium oxides. Implications for the use of lithium and Li2O as breeding materials in fusion reactor blankets

    International Nuclear Information System (INIS)

    Ihle, H.R.; Wu, C.H.; Kudo, H.

    1980-01-01

    In addition to LiH, which has been studied extensively by optical spectroscopy, the existence of a number of other stable lithium hydrides has been predicted theoretically. By analysis of the saturated vapour over dilute solutions of the hydrogen isotopes in lithium, using Knudsen effusion mass spectrometry, all lithium hydrides predicted to be stable were found. Solutions of deuterium in lithium were used predominantly because of practical advantages for mass spectrometric measurements. The heats of dissociation of LiD, Li 2 D, LiD 2 and Li 2 D 2 , and the binding energies of their singly charged positive ions were determined, and the constants of the gas/liquid equilibria were calculated. The existence of these lithium deuterides in the gas phase over solutions of deuterium in lithium leads to enrichment of deuterium in the gas above 1240 K. The enrichment factor, which increases exponentially with temperature and is independent of concentration for low concentrations of deuterium in the liquid, was determined by Rayleigh distillation experiments. It was found that it is thermodynamically possible to separate deuterium from lithium by distillation. One of the alternatives to the use of lithium in (D,T)-fusion reactors as tritium-breeding blanket material is to employ solid lithium oxide. This has a high melting point, a high lithium density and still favourable tritium-breeding properties. Because of its rather high volatility, an experimental study of the vaporization of Li 2 O was undertaken by mass spectrometry. It vaporizes to give lithium and oxygen, and LiO, Li 2 O, Li 3 O and Li 2 O 2 . The molecule Li 3 O was found as a new species. Heats of dissociation, binding energies of the various ions and the constants of the gas/solid equilibria were determined. The effect of using different materials for the Knudsen cells and the relative thermal stabilities of lithium-aluminium oxides were also studied. (author)

  19. Liquid-metal MHD flow in a duct whose cross section changes from a rectangle to a trapezoid, with applications in fusion blanket designs

    Energy Technology Data Exchange (ETDEWEB)

    Walker, J.S.

    1986-04-01

    This paper treats the liquid-metal MHD flow in a semi-infinite rectangular duct and a semi-infinite trapezoidal duct, which are connected by a finite-length transition duct. There is a strong, transverse, uniform magnetic field. The walls parallel to the magnetic field (sides) remain parallel, while the walls intersecting the magnetic field are twisted in the transition duct to provide the change in cross sectional shape. The left side has a constant height, while the height of the right side increases or decreases in the transition duct. This geometry gives a skewed velocity profile with a high velocity near the left side, provided the right side is relatively thick. All walls are thin and electrically conducting, but the sides are considerably thicker than the other walls. The application is to fusion-reactor blankets in which a high velocity near the first wall (separating the plasma chamber from the coolant) improves the thermal performance. Junctions of different ducts with walls parallel to the magnetic field are treated for the first time. In expansions, contractions and other geometric transition ducts, as well as in straight ducts with axially varying magnetic fields, the fluid flow and electric currents are concentrated in boundary layers adjacent to the sides and in the side. At a junction with a straight duct with a uniform magnetic field, the flow and current must transfer from the boundary layers adn sides to the core regions. These transfers at junctions play a key role in any three-dimensional flow.

  20. 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

  1. Safety and environmental impact of the dual coolant blanket concept. SEAL subtask 6.2, final report

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Dammel, F.; Gabel, K.; Jordan, T.; Schmuck, I.

    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 dual coolant type representing the liquid breeder line. In the dual coolant concept the breeder material (Pb-17Li) is circulated to external heat exchangers to carry away the bulk of the generated heat and to extract the tritium. Additionally, the heavily loaded first wall is cooled by high pressure helium gas. The safety and environmental impact of the dual coolant blanket concept has been assessed as part of the blanket concept selection excercise, a European concerted action, aiming at selecting the two most promising concepts for futher development. The topics investigated are: (a) Blanket materials and toxic materials inventory, (b) energy sources for mobilisation, (c) fault tolerance, (d) tritium and activation products release, and (e) waste generation and management. No insurmountable safety problems have been identified for the dual coolant blanket. 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 longterm Programme' (SEAL). The unresolved issues pertaining to the dual coolant blanket which would need further investigations in future programmes are outlined herein. (orig.) [de

  2. European Fusion Materials Research Program - Recent Results and Future Strategy

    International Nuclear Information System (INIS)

    Diegele, E.; Andreani, R.; Laesser, R.; Schaaf, B. van der

    2005-01-01

    The paper reviews the objectives and the status of the current EU long-term materials program. It highlights recent results, discusses some of the key issues and major existing problems to be resolved and presents an outlook on the R and D planned for the next few years. The main objectives of the Materials Development program are the development and qualification of reduced activation structural materials for the Test Blanket Modules (TBMs) in ITER and of low activation structural materials resistant to high fluence neutron irradiation for in-vessel components such as breeding blanket, divertor and first wall in DEMO. The EU strategy assumes: (i) ITER operation starting in 2015 with DEMO relevant Test Blanket Modules to be installed from day one of operation, (ii) IFMIF operation in 2017 and (iii) DEMO final design activities in 2022 to 2025. The EU candidate structural material EUROFER for TBMs has to be fully code qualified for licensing well before 2015. In parallel, research on materials for operation at higher temperatures is conducted following a logical sequence, by supplementing EUROFER with the oxide dispersion strengthened ferritic steels and, thereafter, with fibre-reinforced Silicon Carbide (SiC f /SiC). Complementary, tungsten alloys are developed as structural material for high temperature applications such as gas-cooled divertors

  3. Blanket maintenance by remote means using the cassette blanket approach

    International Nuclear Information System (INIS)

    Werner, R.W.

    1978-01-01

    Induced radioactivity in the blanket and other parts of a fusion reactor close to the plasma zone will dictate remote assembly, disassembly, and maintenance procedures. Time will be of the essence in these procedures. They must be practicable and certain. This paper discusses the reduction of a complicated Tokamak reactor to a simpler assembly via the use of a vacuum building in which to house the reactor and the introduction in this new model of cassette blanket modules. The cassettes significantly simplify remote handling

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

    Science.gov (United States)

    Saeidi, Sheida

    Two important components of a liquid breeder blanket of a fusion power reactor are the liquid breeder/coolant and the steel structure that the liquid is enclosed in. One candidate combination for such components is Lead-Lithium (PbLi) eutectic alloy and advanced Reduced Activation Ferritic/Martensitic (RAFM) steel. The research performed here is aimed at: (1) better understanding of corrosion processes in the system including RAFM steel and flowing PbLi in the presence of a strong magnetic field and (2) prediction of corrosion losses in conditions of a Dual Coolant Lead Lithium (DCLL) blanket, which is at present the key liquid metal blanket concept in the US. To do this, numerical and analytical tools have been developed and then applied to the analysis of corrosion processes. First, efforts were taken to develop a computational suite called TRANSMAG (Transport phenomena in Magnetohydrodynamic Flows) as an analysis tool for corrosion processes in the PbLi/RAFM system, including transport of corrosion products in MHD laminar and turbulent flows. The computational approach in TRANSMAG is based on simultaneous solution of flow, energy and mass transfer equations with or without a magnetic field, assuming mass transfer controlled corrosion and uniform dissolution of iron in the flowing PbLi. Then, the new computational tool was used to solve an inverse mass transfer problem where the saturation concentration of iron in PbLi was reconstructed from the experimental data resulting in the following correlation: CS = e 13.604--12975/T, where T is the temperature of PbLi in K and CS is in wppm. The new correlation for saturation concentration was then used in the analysis of corrosion processes in laminar flows in a rectangular duct in the presence of a strong transverse magnetic field. As shown in this study, the mass loss increases with the magnetic field such that the corrosion rate in the presence of a magnetic field can be a few times higher compared to purely

  5. Towards fusion power

    International Nuclear Information System (INIS)

    Venkataraman, G.

    1975-01-01

    An attempt has been made to present general but broad review of the recent developments in the field of plasma physics and its application to fusion power. The first chapter describes the fusion reactions and fusion power systems. The second chapter deals in detail with production and behaviour of plasma, screening, oscillations, instability, energy losses, temperature effects, etc. Magnetic confinements, including pinch systems, toroidal systems such as Tokamac and stellarator, minor machine, etc. are discussed in detail in chapter III. Laser produced plasma, laser implosion and problems associated with it and future prospects are explained in chapter IV. Chapter V is devoted entirely to the various aspects of hybrid systems. The last chapter throws light on problems of fusion technology, such as plasma heating, vacuum requirements, radiation damage, choice of materials, blanket problems, hazards of fusion reactions, etc. (K.B.)

  6. Advanced smart tungsten alloys for a future fusion power plant

    Science.gov (United States)

    Litnovsky, A.; Wegener, T.; Klein, F.; Linsmeier, Ch; Rasinski, M.; Kreter, A.; Tan, X.; Schmitz, J.; Mao, Y.; Coenen, J. W.; Bram, M.; Gonzalez-Julian, J.

    2017-06-01

    The severe particle, radiation and neutron environment in a future fusion power plant requires the development of advanced plasma-facing materials. At the same time, the highest level of safety needs to be ensured. The so-called loss-of-coolant accident combined with air ingress in the vacuum vessel represents a severe safety challenge. In the absence of a coolant the temperature of the tungsten first wall may reach 1200 °C. At such a temperature, the neutron-activated radioactive tungsten forms volatile oxide which can be mobilized into atmosphere. Smart tungsten alloys are being developed to address this safety issue. Smart alloys should combine an acceptable plasma performance with the suppressed oxidation during an accident. New thin film tungsten-chromium-yttrium smart alloys feature an impressive 105 fold suppression of oxidation compared to that of pure tungsten at temperatures of up to 1000 °C. Oxidation behavior at temperatures up to 1200 °C, and reactivity of alloys in humid atmosphere along with a manufacturing of reactor-relevant bulk samples, impose an additional challenge in smart alloy development. First exposures of smart alloys in steady-state deuterium plasma were made. Smart tungsten-chroimium-titanium alloys demonstrated a sputtering resistance which is similar to that of pure tungsten. Expected preferential sputtering of alloying elements by plasma ions was confirmed experimentally. The subsequent isothermal oxidation of exposed samples did not reveal any influence of plasma exposure on the passivation of alloys.

  7. Industry perspectives on future directions in the fusion program

    International Nuclear Information System (INIS)

    Maniscalco, J.A.; Bell, J.M.

    1985-01-01

    Industry is the ultimate recipient of the product of the fusion development program. However, budget trends are causing the program to lose it's focus on the energy goal, thus diminishing opportunities for a meaningful industrial role at a time when technical progress has been remarkable and scientific feasibility is being demonstrated. The findings of the Magnetic Fusion Advisory Committee Panel charged to report on industrial participation in fusion energy development are summarized. A recommendation of this panel was to increase intellectual involvement of industry in the development of fusion. Opportunities to achieve this include forming partnerships with national laboratories and universities, assigning industry responsibility for a complete scope of work, and assigning industry a major role in system studies and reactor design. These opportunities can be implemented without requiring large budget increases. Increasing the involvement of industry in the fusion program will provide both long and short-term benefits

  8. Fusion systems engineering

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    Research during this report period has covered the following areas: (1) fusion reactor systems studies, (2) development of blanket processing technology for fusion reactors, (3) safety studies of fusion concepts, (4) MACKLIB-IV, a new library of nuclear response functions, (5) energy storage and power supply requirements for commercial fusion reactors, (6) blanket/shield design evaluation for commercial fusion reactors, and (7) cross section measurements, evaluations, and techniques

  9. Multiple Module Simulation of Water Cooled Breeding Blankets in K-DEMO Using Thermal-Hydraulic Analysis Code MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A preliminary concept for the Korean fusion demonstration reactor (K-DEMO) has been studied by the National Fusion Research Institute (NFRI) based on the National Fusion Roadmap of Korea. The feasibility studies have been performed in order to establish the conceptual design guidelines of the breeding blanket. As a part of the NFRI research, Seoul National University (SNU) is conducting thermal design, evaluation and validation of the water-cooled breeding blanket for the K-DEMO reactor. The purpose of this study is to extend the capability of MARS-KS to the overall blanket system analysis which includes 736 blanket modules in total. The strategy for the multi-module blanket system analysis using MARS-KS is introduced and the analysis result of the 46 blanket modules of single sector was summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for thermal analysis of the conceptual design of the K-DEMO breeding blanket. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering the pressure drop that occurs in each module. For a feasibility test of the proposed methodology, 46 blankets in a sector, which are connected with each other through the common headers for the sector inlet and outlet, were simulated. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation. Because of parallelization using the MPI system, the computational time could be reduced significantly. In future, this methodology will be extended to an efficient simulation of multiple sectors, and further validation for transient simulation will be carried out for more practical applications.

  10. IFMIF suitability for evaluation of fusion functional materials

    International Nuclear Information System (INIS)

    Casal, N.; Sordo, F.; Mota, F.; Jordanova, J.; Garcia, A.; Ibarra, A.; Vila, R.; Rapisarda, D.; Queral, V.; Perlado, M.

    2011-01-01

    The International Fusion Materials Irradiation Facility (IFMIF) is a future neutron source based on the D-Li stripping reaction, planned to test candidate fusion materials at relevant fusion irradiation conditions. During the design of IFMIF special attention was paid to the structural materials for the blanket and first wall, because they will be exposed to the most severe irradiation conditions in a fusion reactor. Also the irradiation of candidate materials for solid breeder blankets is planned in the IFMIF reference design. This paper focuses on the assessment of the suitability of IFMIF irradiation conditions for testing functional materials to be used in liquid blankets and diagnostics systems, since they are been also considered within IFMIF objectives. The study has been based on the analysis and comparison of the main expected irradiation parameters in IFMIF and DEMO reactor.

  11. European blanket development for a demo reactor

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Anzidei, L.

    1994-01-01

    There are four breeding blanket concepts for a fusion DEMO reactor under development within the framework of the fusion technology programme of the European Union (EU). This paper describes the design of these concepts, the accompanying R + D programme and the status of the development. (authors). 8 figs., 1 tab

  12. First Wall, Blanket, Shield Engineering Technology Program

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1982-01-01

    The First Wall/Blanket/Shield Engineering Technology Program sponsored by the Office of Fusion Energy of DOE has the overall objective of providing engineering data that will define performance parameters for nuclear systems in advanced fusion reactors. The program comprises testing and the development of computational tools in four areas: (1) thermomechanical and thermal-hydraulic performance of first-wall component facsimiles with emphasis on surface heat loads; (2) thermomechanical and thermal-hydraulic performance of blanket and shield component facsimiles with emphasis on bulk heating; (3) electromagnetic effects in first wall, blanket, and shield component facsimiles with emphasis on transient field penetration and eddy-current effects; (4) assembly, maintenance and repair with emphasis on remote-handling techniques. This paper will focus on elements 2 and 4 above and, in keeping with the conference participation from both fusion and fission programs, will emphasize potential interfaces between fusion technology and experience in the fission industry

  13. Inertial confinement fusion: present status and future potential

    International Nuclear Information System (INIS)

    Hogan, W.J.

    1984-01-01

    Power from inertial confinement fusion holds much promise for society. This paper points out many of the benefits relative to combustion of hydrocarbon fuels and fission power. Potential problems are also identified and put in perspective. The progress toward achieving inertial fusion power is described and results of recent work at the Lawrence Livermore National Laboratory are presented. Key phenomenological uncertainties are described and experimental goals for the Nova laser system are given. Several ICF reactor designs are discussed

  14. Recent developments concerning the fusion; Developpements recents sur la fusion

    Energy Technology Data Exchange (ETDEWEB)

    Jacquinot, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint Paul lez Durance (France); Andre, M. [CEA/DAM Ile de France, 91 - Bruyeres Le Chatel (France); Aymar, R. [ITER Joint Central Team Garching, Muenchen (Germany)] [and others

    2000-09-04

    Organized the 9 march 2000 by the SFEN, this meeting on the european program concerning the fusion, showed the utility of the exploitation and the enhancement of the actual technology (JET, Tore Supra, ASDEX) and the importance of the Europe engagement in the ITER program. The physical stakes for the magnetic fusion have been developed with a presentation of the progresses in the knowledge of the stability limits. A paper on the inertial fusion was based on the LMJ (Laser MegaJoule) project. The two blanket concepts chosen in the scope of the european program on the tritium blankets, have been discussed. These concepts will be validated by irradiation tests in the ITER-FEAT and adapted for a future reactor. (A.L.B.)

  15. Stat-of-the art of nuclear fusion and its future outlook in

    International Nuclear Information System (INIS)

    Abdelaziz, M.E.; Elnadi, A.M.; Masoud, M.; Elshaer, M.A.; Khalil, S.M.

    1993-01-01

    The study in this project is carried out with the objective of being able to present a clear view for the state-of-the art of nuclear fusion as one of the most promising coming energy source and its future outlook in Egypt. The study introduce a summary of the world energy problem and the advantages of thermonuclear fusion energy compared to other energy sources. A description of the two main techniques of confining plasma in the fusion experiments, namely the magnetic and the inertial confinement. These techniques are discussed and investigated through linear pinches and tokamaks. Tokamaks showed to be a promising machines for achieving the controlled thermonuclear fusion power reactor. Recent development of the research on laser fusion together with fast progress in pellet and laser technology suggest that it may be possible to achieve laser fusion power reactor. The story of the strange phenomena of cold fusion, muon-catalyzed fusion, and cold fusion in condensed matter are also studied and showed to be non promising. The project study in details the future fusion reactor, its nuclear engineering and its safety and environmental aspects. The study is based on the magnetic fusion using the tokamak configuration. The positive safety and environmental aspects of fusion reactors, if exist, is also investigated. Status of plasma physics and nuclear fusion activities and strategies in the developing countries (including egypt and the arab countries) are reviewed, besides, some national programmes are proposed. In addition, the status of international activities in plasma technology and its application are represented. Future outlook for egyptian programmes on different plasma technologies are studied. Finally, conclusions and recommendations are presented which summarized the principle achiements and future research opportunities in nuclear fusion activities. In fact, it must be emphasized that fusion is an exciting and challenging field of research -the most

  16. 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

  17. Mirror fusion--fission hybrids

    International Nuclear Information System (INIS)

    Lee, J.D.

    1978-01-01

    The fusion-fission concept and the mirror fusion-fission hybrid program are outlined. Magnetic mirror fusion drivers and blankets for hybrid reactors are discussed. Results of system analyses are presented and a reference design is described

  18. Hydrogen production in early generation fusion power plant and its socio-economic implication

    International Nuclear Information System (INIS)

    Konishi, S.; Yamamoto, Y.

    2007-01-01

    Full text: This paper describes technical possibility of high temperature blanket for the early generation of fusion power plant and its application to hydrogen production. Its anticipated implication and strategy from the socio-economic aspects will be also discussed. Material and energy balances, such as fuel supply and delivery of product energy from fusion plants, as well as waste discharge and accident scenario that lead to environmental impact, are characterized by blanket concepts. Thus blankets are considered to dominate the feature of fusion energy that should respond to the requirements of the sponsors, i.e., public and future market. Fusion blanket concept based on the combinations of LiPb and SiC materials are regarded as a candidate for ITER/TBM, and at the same time, applied in various DEMO designs encompassing high temperature output. Recent developments of SiC-LiPb blanket in Japan, EU, US or China suggests staged development paths starting from TBMs and targeting high temperature blanket and efficient energy output from early generation plants. These strategies are strongly affected by the views of these parties on fusion energy, from the aspects of socio-economics. Hydrogen production process with the high temperature blanket is one of the most important issues, because temperature range much higher than is possible with current or near future fission plants are needed, suggesting market possibility different from that of fission. Fuel cycles, particularly lithium supply and TBR control will be also important. Self-sustained fusion fuel cycle requires technical capability to maintain the lithium contents. Liquid blanket has an advantage in continuous and real-time control TBR in a plant, but large amount of lithium-6 and initial tritium supply remains as issues. As for the environmental effect, normal operation release, assumed accidental scenario, and rad-waste will be the key issue to dominate social acceptance of fusion. (author)

  19. Hydrogen production in early generation fusion power plant and its socio-economic implication

    International Nuclear Information System (INIS)

    Konishi, Satoshi; Yamamoto, Yasushi

    2008-01-01

    This paper describes technical possibility of high temperature blanket for the early generation of fusion power plant and its application to hydrogen production. Its anticipated implication and strategy from the socio-economic aspects will be also discussed. Material and energy balances, such as fuel supply and delivery of product energy from fusion plants, as well as waste discharge and accident scenario that lead to environmental impact, are characterized by blanket concepts. Thus blankets are considered to dominate the feature of fusion energy that should respond to the requirements of the sponsors, i.e., public and future market. Fusion blanket concept based on the combinations of LiPb and SiC materials are regarded as a candidate for ITER/TBM, and at the same time, applied in various DEMO designs encompassing high temperature output. Recent developments of SiC-LiPb blanket in Japan, EU, US or China suggests staged development paths starting from TBMs and targeting high temperature blanket and efficient energy output from early generation plants. These strategies are strongly affected by the views of these parties on fusion energy, from the aspects of socio-economics. Hydrogen production process with the high temperature blanket is one of the most important issues, because temperature range much higher than is possible with current or near future fission plants are needed, suggesting market possibility different from that of fission. Fuel cycles, particularly lithium supply and TBR control will be also important. Self-sustained fusion fuel cycle requires technical capability to maintain the lithium contents. Liquid blanket has an advantage in continuous and real-time control TBR in a plant, but large amount of lithium-6 and initial tritium supply remains as issues. As for the environmental effect, normal operation release, assumed accidental scenario, and rad-waste will be the key issue to dominate social acceptance of fusion. (author)

  20. Neutron dosimetry for the TFTR Lithium-Blanket-Module program

    International Nuclear Information System (INIS)

    Harker, Y.D.; Tsang, F.Y.; Caffrey, A.J.; Homeyer, W.G.; Engholm, B.A.

    1981-01-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-a-kind neutronics experiment involving a prototypical fusion reactor blanket module with a distributed neutron source from the plasma of the TFTR at Princeton Plasma Physics Laboratory. The objectives of the LBM program are: (1) to test the capabilities of neutron transport codes when applied to fusion test reactor blanket conditions, and (2) to obtain tritium breeding performance data on a typical design concept of a fusion-reactor blanket. This paper addresses the issues relative to the measurement of neutron fields in the LBM, presents the results of preliminary design studies concerning neutron measurements and also presents the results of blanket mockup experiments performed at the Idaho National Engineering Laboratory

  1. 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)

  2. Present status and future prospects for direct drive laser fusion

    International Nuclear Information System (INIS)

    Bodner, S.E.

    1986-01-01

    If one assumes that the best short wavelength laser will have an efficiency of 5--7%, and if one assumes that reasonable cost electricity requires that the product of laser efficiency and pellet gain be greater than 10--15, then pellet grains for laser fusion must be at least 150--300. The only laser fusion concept with any potential for energy applications then seems to be directly driven targets with moderately thin shells and 1/4 micron KrF laser light. This direct drive concept has potential pellet energy gains of 200--300

  3. Fusion systems engineering

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    Information is given on each of the following topics: (1) fusion reactor systems studies, (2) development of blanket processing technology for fusion reactors, (3) safety studies of CTR concepts, and (4) cross section measurements and techniques

  4. 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

  5. ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

    International Nuclear Information System (INIS)

    WONG, CPC; MALANG, S; NISHIO, S; RAFFRAY, R; SAGARA, S

    2002-01-01

    OAK A271 ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS. First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability

  6. Possible futures for the development of a fusion demonstration plant

    International Nuclear Information System (INIS)

    Nichols, S.P.

    1976-01-01

    As indicated by the Fusion Planning Bulletins, the Division of Controlled Thermonuclear Research is becoming involved in planning with alternative scenarios. The Center for Energy Studies at the University of Texas at Austin has been involved with such planning for several years and has examined various scenarios for fusion power development using the Partitive Analytical Forecasting (PAF) technique. The most recent studies compare the long-term plan presented in WASH-1290, Fusion Power by Magnetic Confinement, with other plans that have been proposed, such as the plan proposed by Kulcinski and Conn of the University of Wisconsin. The study indicates that some of the alternative plans do have possibilities to shorten the required time for the completion of a demonstration fusion plant without increased costs or a decrease in the likelihood of success. The current efforts of the project are in the planning exercises recently completed by committees set up by the DCTR. Further comparisons of alternative scenarios will be performed as part of this effort

  7. (D,T) Driven thorium hybrid blankets

    International Nuclear Information System (INIS)

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

    1983-01-01

    Recently, a project has started, with the aim to establish the neutronic performance and the basic design of an experimental fusionfission (hybrid) reactor facility, called AYMAN, in cylinderical geometry. The fusion reactor will have to be simulated by a (D,T) neutron generator. Fissile and fertile fuel will have to surround the neutron generator as a cylinderical blanket to simulate the boundary conditions of the hybrid blanket in a proper way. This geometry is consistent with Tandem Mirror Hybrid Blanket design and with most of the ICF blanket designs. A similar experimental installation will become operational around 1984 at the Swiss Federal Institute of Technology in Lausanne, Switzerland known under the project LOTUS. Due to the limited dimensions of the experimental cavity of the LOTUS-hybrid reactor, the LOTUS blankets have to be designed in plane geometry. Also, the bulky form of the Haefely neutron generator of the LOTUS facility obliges one to design a blanket in the plane geometry. This results in a vacuum left boundary conditions for the LOTUS blanket. The importance of a reflecting left boundary condition on the overall neutronic performance of a hybrid blanket has been analyzed in previous work in detail

  8. 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

  9. 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

  10. An Assessment of the Economics of Future Electric Power Generation Options and the Implications for Fusion

    International Nuclear Information System (INIS)

    Delene, Jerry G.; Sheffield, John; Williams, Kent A.; Reid, R. Lowell; Hadley, Stan

    2001-01-01

    This study examines the potential range of electric power costs for some major alternatives to fusion electric power generation when it is ultimately deployed in the middle of the 21st century and, thus, offers a perspective on the cost levels that fusion must achieve to be competitive. The alternative technologies include coal burning, coal gasification, natural gas, nuclear fission, and renewable energy. The cost of electricity (COE) from the alternatives to fusion should be in a 30 to 53 mills/kW.h (1999 dollars) range if carbon sequestration is not needed, 30 to 61 mills/kW.h if sequestration is required, or as high as 83 mills/kW.h for the worst-case scenario for cost uncertainty. The reference COE range for fusion was estimated at 65 to 102 mills/kW.h for 1- to 1.3-GW(electric) scale power plants, based on the tokamak concept. Tokamak fusion costs will have to be reduced and/or cost-effective alternative nontokamak concepts devised before fusion will be competitive with the alternatives for the future production of electricity. Fortunately, there are routes to achieve this goal. Recent results from fusion experiments and developments in technology and engineering solutions indicate that lower cost fusion power plants are possible at the 1-GW(electric) level. Another general route for fusion to reduce costs is to go to large plant sizes [multigigawatts (electric)

  11. Laser fusion and future energy sources - some recent results

    International Nuclear Information System (INIS)

    Hora, H.

    1979-01-01

    While the laser fusion is at present producing more genuine fusion neutrons than the tokamak with magnetic confinement, if use of short laser pulses is preferred, the then appearing nonlinear effect causes considerable complications. Nonlinear processes for the preferred geometry of perpendicular incidence can avoid the problems of resonance absorption, while parametric instabilities have no quantitative influence on the energy balance. The early stages of interaction show the generation of thick 'cold' compressing plasma blocks which can be used for a nonlinear force fast pusher compression of high efficiency (low entropy production). A short time interaction results in a fast thermalization of the plasma corona by soliton decay and this provides the necessary condition for Nuckolls' gasdynamic ablation compression. For longer duration of high intensity irradiation, a pulsation of reflectivity and thermalization will complicate the interaction

  12. Materials technology for fusion - Current status and future requirements

    International Nuclear Information System (INIS)

    Gold, R.E.; Bloom, E.E.; Clinard, F.W. Jr.; Smith, D.L.; Stevenson, R.D.; Wolfer, W.G.

    1981-01-01

    The general status of the materials research and development activities currently under way in support of controlled thermonuclear fusion reactors in the United States is reviewed. In the area of magnetic confinement configurations, attention is given to development programs for first wall materials, which are at various stages for possible austenitic stainless steels, high-strength Fe-Ni-Cr alloys, reactive and refractory metal alloys, specially designed long-range ordered and rapidly solidified alloys, and ferritic/martensitic steels, and for tritium breeding materials, electrical insulators, ceramics, and coolants. The development of materials for inertial confinement reactors is also surveyed in relation to the protection scheme employed for the first wall and the effects of pulsed neutron irradiation. Finally, the materials requirements and selection procedures for the ETF/INTOR and Starfire tokamak reactor designs are examined. Needs for the expansion of research on nonfirst-wall materials and inertial confinement fusion reactor material requirements are pointed out

  13. Recent Accomplishments and Future Directions in US Fusion Safety & Environmental Program

    Energy Technology Data Exchange (ETDEWEB)

    David A. Petti; Brad J. Merrill; Phillip Sharpe; L. C. Cadwallader; L. El-Guebaly; S. Reyes

    2006-07-01

    The US fusion program has long recognized that the safety and environmental (S&E) potential of fusion can be attained by prudent materials selection, judicious design choices, and integration of safety requirements into the design of the facility. To achieve this goal, S&E research is focused on understanding the behavior of the largest sources of radioactive and hazardous materials in a fusion facility, understanding how energy sources in a fusion facility could mobilize those materials, developing integrated state of the art S&E computer codes and risk tools for safety assessment, and evaluating S&E issues associated with current fusion designs. In this paper, recent accomplishments are reviewed and future directions outlined.

  14. Flibe blanket concept for transmuting transuranic elements and long lived fission products

    International Nuclear Information System (INIS)

    Gohar, Y.

    2000-01-01

    A Molten salt (Flibe) fusion blanket concept has been developed to solve the disposition problems of the spent nuclear fuel and the transuranic elements. This blanket concept can achieve the top rated solution, the complete elimination of the transuranic elements and the long-lived fission products. Small driven fusion devices with low neutron wall loading and low neutron fluence can perform this function. A 344-MW integrated fusion power from D-T plasmas for thirty years with an availability factor of 0.75 can dispose of 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. In addition, the utilization of this blanket concept eliminates the need for a geological repository site, which is a major advantage. This application provides an excellent opportunity to develop and to enhance the public acceptance of the fusion energy for the future. The energy from the transmutation process is utilized to produce revenue. Flibe, lithium-lead eutectic, and liquid lead are possible candidates. The liquid blankets have several features, which are suited for W application. It can operate at constant thermal power without interruption for refueling by adjusting the concentration of the transuranic elements and lithium-6. These liquids operate at low-pressure, which reduces the primary stresses in the structure material. Development and fabrication costs of solid transuranic materials are eliminated. Burnup limit of the transuranic elements due to radiation effects is eliminated. Heat is generated within the liquid, which simplifies the heat removal process without producing thermal stresses. These blanket concepts have large negative temperature coefficient with respect to the blanket reactivity, which enhances the safety performance. These liquids are chemically and thermally stable under irradiation conditions, which minimize the radioactive waste volume. The operational record of the Molten Salt Breeder Reactor with Flibe was very successful

  15. Some new ideas for Tandem Mirror blankets

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.

    1981-01-01

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

  16. The state of the art report on the development of manufacturing technology of fusion reactor FW blanket and mock-up in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. Y.; Jeong, Y. H.; Baek, J. H.; Kim, J. H.; Kim, H. G

    2004-08-15

    The joining technology of first wall blanket has been developed by JAERI in collaboration with Kawasaki Heavy Industry, Isuau Motors and University of Tsukuba in Japan. A variety of joining technologies including HIP, brazing, casing and friction welding was applied to the manufacturing of SS/SS and Cu/SS joint. In Be/Cu joining, it was emphasized to find the optimal HIP temperature lower than 650 .deg. C in order to avoid excessive SS sensitization because the joining of Be tile to Cu heat sink is a final processing step in the manufacturing of FW blanket. The selected HIP condition were 620 .deg. C, 150MPa and 2hr with Cu interlayer. Sample tests for joints was completed by 1995. The small scale mockup was manufactured and its performance was qualified by end of 2000. From 2001, the manufacturing and the characterization has been carried out for the larger scale mockup.

  17. The state of the art report on the development of manufacturing technology of fusion reactor FW blanket and mock-up in Japan

    International Nuclear Information System (INIS)

    Park, J. Y.; Jeong, Y. H.; Baek, J. H.; Kim, J. H.; Kim, H. G.

    2004-08-01

    The joining technology of first wall blanket has been developed by JAERI in collaboration with Kawasaki Heavy Industry, Isuau Motors and University of Tsukuba in Japan. A variety of joining technologies including HIP, brazing, casing and friction welding was applied to the manufacturing of SS/SS and Cu/SS joint. In Be/Cu joining, it was emphasized to find the optimal HIP temperature lower than 650 .deg. C in order to avoid excessive SS sensitization because the joining of Be tile to Cu heat sink is a final processing step in the manufacturing of FW blanket. The selected HIP condition were 620 .deg. C, 150MPa and 2hr with Cu interlayer. Sample tests for joints was completed by 1995. The small scale mockup was manufactured and its performance was qualified by end of 2000. From 2001, the manufacturing and the characterization has been carried out for the larger scale mockup

  18. Structural analysis under the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Majumdar, S.

    1985-01-01

    Structural design procedures followed in the Blanket Comparison and Selection Study are briefly reviewed. The American Society of Mechanical Engineers Boilers and Pressure Vessels Code, Section III, Code Case N47 has been used as a design guide. Its relevance to fusion reactor applications, however, is open to question and needs to be evaluated in the future. The primary structural problem encountered in tokamak blanket designs is the high thermal stress due to surface heat flux, with fatigue being an additional concern for pulsed systems. The conflicting requirements of long erosion life and high surface heat flux capability imply that some form of stress relief in the first-wall region will be necessary. Simplified stress and fatigue crack growth analyses are presented to show that the use of orthogonally grooved first wall may be a potential solution for mitigating the thermal stress problem. A comparison of three structural alloys on the basis of both grooved and nongrooved first-wall designs is also presented. Other structural problems encountered in tokamak designs include stresses due to plasma disruptions, and magnetohydrodynamic (MHD) pressure drop in liquid-metal-cooled systems. In particular, it is shown that the maximum stress in the side wall of a uniform duct generated by MHD pressure drop cannot be reduced by increasing the wall thickness or by decreasing the span. In contract to tokamak blankets, tandem mirror blankets are far less severely stressed because of a much lower surface heat flux, coolant pressure, and also because of their axisymmetric geometry. Both blankets, however, will require detailed structural dynamics analysis to verify their ability to withstand seismic loadings if the heavy 17Li-83Pb is used as a coolant

  19. The European fusion nuclear technology effort

    International Nuclear Information System (INIS)

    Darvas, J.

    1989-01-01

    The role of fusion technology in the European fusion development strategy is outlined. The main thrust of the present fusion technology programme is responding to development needs of the Next European Torus. A smaller, but important and growing R and D effort is dealing with problems specific to the Demonstration, or Fusion Power, Reactor. The part of the programme falling under the somewhat arbitrarily defined category of 'fusion nuclear technology' is reviewed and an outlook to future activities is given. The review includes tritium technology, blanket technology and breeder materials development, technology and materials for the protection of the first wall and of other plasma facing components, remote handling technology, and safety and environmental impact studies. A few reflections are offered on the future long-term developments in fusion technology. (orig.)

  20. 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)

  1. Development and trial manufacturing of 1/2-scale partial mock-up of blanket box structure for fusion experimental reactor

    International Nuclear Information System (INIS)

    Hashimoto, Toshiyuki; Takatsu, Hideyuki; Sato, Satoshi

    1994-07-01

    Conceptual design of breeding blanket has been discussed during the CDA (Conceptual Design Activities) of ITER (International Thermonuclear Experimental Reactor). Structural concept of breeding blanket is based on box structure integrated with first wall and shield, which consists of three coolant manifolds for first wall, breeding and shield regions. The first wall must have cooling channels to remove surface heat flux and nuclear heating. The box structure includes plates to form the manifolds and stiffening ribs to withstand enormous electromagnetic load, coolant pressure and blanket internal (purge gas) pressure. A 1/2-scale partial model of the blanket box structure for the outboard side module near midplane is manufactured to estimate the fabrication technology, i.e. diffusion bonding by HIP (Hot Isostatic Pressing) and EBW (Electron Beam Welding) procedure. Fabrication accuracy is a key issue to manufacture first wall panel because bending deformation during HIP may not be small for a large size structure. Data on bending deformation during HIP was obtained by preliminary manufacturing of HIP elements. For the shield structure, it is necessary to reduce the welding strain and residual stress of the weldment to establish the fabrication procedure. Optimal shape of the parts forming the manifolds, welding locations and welding sequence have been investigated. In addition, preliminary EBW tests have been performed in order to select the EBW conditions, and fundamental data on built-up shield have been obtained. Especially, welding deformation by joining the first wall panel to the shield has been measured, and total deformation to build-up shield by EBW has been found to be smaller than 2 mm. Consequently, the feasibility of fabrication technologies has been successfully demonstrated for a 1m-scaled box structure including the first wall with cooling channels by means of HIP, EBW and TIG (Tungsten Inert Gas arc)-welding. (author)

  2. Impact analysis of the time trend of TBR and irradiation damage assessment of HCSB blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Zeng, Qin, E-mail: zengqin@ustc.edu.cn; Chen, Hongli; Lv, Zhongliang; Pan, Lei; Zhang, Haoran; Shi, Wei

    2017-01-15

    Chinese Fusion Engineering Testing Reactor (CFETR) is a test tokamak reactor to bridge the gap between ITER and future fusion power plants and to demonstrate generation of fusion power in China. In fusion power plants, tritium is generated from the reaction of neutron and Lithium. One of the missions of CFETR is the full cycle of tritium self-sufficiency. For the mission, a Helium Cooled Solid Breeder blanket (HCSB) was proposed for CFETR and its conceptual design has been carried out. In order to assess the capacity of the tritium breeding and irradiation damage of first wall of the HCSB blanket during the 8 years’ engineering test stage, this paper presents the time trend of TBR analysis and irradiation damage assessment of HCSB blanket based on the three-dimensional (3D) neutronics model which is created by McCad. In the 3D neutronics model, the outboard blanket on equatorial plane is described based on the detailed 3D engineering model. The calculations were performed by MCNP and FISPACT with FENDL/2.1 data library. The impact analysis of the thickness of coolant plates (CP) and the structural material content in CPs to the TBR is assessment.

  3. The recent progress of laser fusion research and future scope

    International Nuclear Information System (INIS)

    Yamanaka, C.

    1986-01-01

    The plasma compression of spherical fuel pellets is performed by irradiation laser beams on the surface of targets. The short wavelength laser or Xray is effective to get high coupling of laser and plasmas without preheating. The implosion uniformity is essentially important to attain the high compression. As for the direct implosion, the multibeam irradiation is necessary to keep a good uniformity of illumination. Extremely high aspect ratio targets are successfully imploded withy neutron yield 10/sup 12/ or more. The shock wave multiplexing is introduced by tailored laser pulses synchronizing with the compression stagnation. Implosion instability seems to be prevented by this scheme. Energy recovering by nuclear fusion is about 10/sup -3/ of the incident laser beam. The indirect implosion using the Cannonball target is very effective to keep the high absorption and the implosion uniformity. However the suprathermal electrons are increased especially at the region of the beam inlet holes. The larger cavity irradiated by the shorter wavelength laser indicates the better results. The Xray conversion by laser is intensively studied using metal targets. Magnetically Insulated Inetially Confined Fusion (MICF) is tested by using CO/sub 2/ lasers. The basic structure of the MICF target is a double shell structure. The irradiation of laser beams through holes of the outer shell produces a toroidal magnetic field due to the current loop produced by the ejected hot electrons. Self organized magnetic field is expected to confine the plasma energy. Plasmas are preserved by the inertial confinement scheme. The experimental results are very interesting to design a hybrid fusion device

  4. The state of the art report on the fabrication of FW blanket for the fusion reactor and mock-up development in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Whan; Baek, Jong Hyuk; Park, Jeong Yong; Kim, Hyun Gil; Jeong, Yong Hwan

    2004-08-01

    Blanket-shield system in ITER is the component where it directly is faced with high-heat plasma. Function of blanket is to sustain extremely high temperature environment as well as to remove heat flux generated its surface. It mainly consists of plasma facing part, heat sinking part and structural part. Plasma facing part is made of armour materials such as beryllium, tungsten and carbon fiber composite. Heat sinking part is made of copper alloy to maximize heat transfer into flowing coolant inside of blanket. Structural material is used in 316LN stainless steel. As joining such dissimilar materials emerged as an issue, many developed countries have spurred the development of joint technology. This technical report was focused on the activities of EU regarding joining beryllium, copper and stainless steel. EU have adopted to Hot Isostatic Pressing (HIP) to join beryllium, copper and stainless steel. Although brazing process is not actively investigated compared as HIP, it still investigated in some countries to support HIP. Fabrication of mock-up is accomplished by CEA in France to finish small scale mock-up in 1996, medium and large scale mock-up in 1997. In recent, FRAMATOME in EU has focused on manufacturing prototype used for ITER.

  5. The state of the art report on the fabrication of FW blanket for the fusion reactor and mock-up development in Europe

    International Nuclear Information System (INIS)

    Kim, Jun Whan; Baek, Jong Hyuk; Park, Jeong Yong; Kim, Hyun Gil; Jeong, Yong Hwan

    2004-08-01

    Blanket-shield system in ITER is the component where it directly is faced with high-heat plasma. Function of blanket is to sustain extremely high temperature environment as well as to remove heat flux generated its surface. It mainly consists of plasma facing part, heat sinking part and structural part. Plasma facing part is made of armour materials such as beryllium, tungsten and carbon fiber composite. Heat sinking part is made of copper alloy to maximize heat transfer into flowing coolant inside of blanket. Structural material is used in 316LN stainless steel. As joining such dissimilar materials emerged as an issue, many developed countries have spurred the development of joint technology. This technical report was focused on the activities of EU regarding joining beryllium, copper and stainless steel. EU have adopted to Hot Isostatic Pressing (HIP) to join beryllium, copper and stainless steel. Although brazing process is not actively investigated compared as HIP, it still investigated in some countries to support HIP. Fabrication of mock-up is accomplished by CEA in France to finish small scale mock-up in 1996, medium and large scale mock-up in 1997. In recent, FRAMATOME in EU has focused on manufacturing prototype used for ITER

  6. Advanced high performance solid wall blanket concepts

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Malang, S.; Nishio, S.; Raffray, R.; Sagara, A.

    2002-01-01

    First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability

  7. Overview of the TFTB lithium blanket module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

    The Lithium Blanket Module (LBM) is an ∼ 80-cm 3 module, representative of a helium-cooled lithium oxide fusion reactor blanket module. This paper summarizes the design, development, and construction of the LBM, and indicates the present status of the LBM program

  8. General description of preliminary design of an experimental fusion reactor and the future problems

    International Nuclear Information System (INIS)

    Sako, Kiyoshi

    1976-01-01

    Recently, the studies on plasma physics has progressed rapidly, and promising experimental data emerged successively. Especially expectation mounts high that Tokamak will develop into power reactors. In Japan, the construction of large plasma devices such as JT-60 of JAERI is going to start, and after several years, the studies on plasma physics will come to the end of first stage, then the main research and development will be directed to power reactors. The studies on the design of practical fusion reactors have been in progress since 1973 in JAERI, and the preliminary design is being carried out. The purposes of the preliminary design are the clarification of the concept of the experimental reactor and the requirements for the studies on core plasma, the examination of the problems for developing main components and systems of the reactor, and the development of design technology. The experimental reactor is the quasi-steady reactor of 100 MW fusion reaction output, and the conditions set for the design and the basis of their setting are explained. The outline of the design, namely core plasma, blankets, superconductive magnets and the shielding with them, vacuum wall, neutral particle injection heating device, core fuel supply and exhaust system, and others, is described. In case of scale-up the reactor structural material which can withstand neutron damage must be developed. (Kako, I.)

  9. Fusion systems engineering

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    Summaries of research are included for each of the following topics: (1) fusion reactor systems studies, (2) development of blanket processing technology for fusion reactors, (3) safety studies of fusion concepts, (4) the MACK/MACKLIB system for nuclear response functions, and (5) energy storage and power supply systems for fusion reactors

  10. Tertiary proton diagnostics in future inertial confinement fusion experiments

    International Nuclear Information System (INIS)

    Cremer, S.; Verdon, C.P.; Petrasso, R.D.

    1998-01-01

    Recently, it was proposed to use energetic (up to 31 MeV) tertiary protons produced during the final stage of inertial confinement fusion implosions to measure the fuel areal density of compressed deuterium endash tritium (DT). The method is based on seeding the fuel with 3 He. The reaction of 3 He ions with the energetic knock-on deuterons, produced via the elastic scattering of 14.1 MeV neutrons, is a source of very energetic protons capable of escaping from very large areal density targets. This work presents results of detailed time-dependent Monte Carlo simulations of the nuclear processes involved in producing and transporting these protons through imploding targets proposed for direct-drive experiments on OMEGA [D. K. Bradley et al., Phys. Plasmas 5, 1870 (1998)] and the National Ignition Facility [S. W. Haan et al., Phys. Plasmas 2, 2480 (1995)]. copyright 1998 American Institute of Physics

  11. Plasma facing materials and components for future fusion devices - development, characterization and performance under fusion specific loading conditions

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J. [Forschungszentrum Juelich (Germany). Inst. fuer Plasmaphysik

    2006-04-15

    The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive RandD. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation.

  12. Plasma facing materials and components for future fusion devices - development, characterization and performance under fusion specific loading conditions

    International Nuclear Information System (INIS)

    Linke, J.

    2006-01-01

    The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive RandD. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation

  13. Plasma facing materials and components for future fusion devices—development, characterization and performance under fusion specific loading conditions

    Science.gov (United States)

    Linke, J.

    2006-04-01

    The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive R&D. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation.

  14. Superconductivity Engineering and Its Application for Fusion 3.Superconducting Technology as a Gateway to Future Technology

    Science.gov (United States)

    Asano, Katsuhiko

    Hopes for achieving a new source of energy through nuclear fusion rest on the development of superconducting technology that is needed to make future equipments more energy efficient as well as increase their performance. Superconducting technology has made progress in a wide variety of fields, such as energy, life science, electronics, industrial use and environmental improvement. It enables the actualization of equipment that was unachievable with conventional technology, and will sustain future “IT-Based Quality Life Style”, “Sustainable Environmental” and “Advanced Healthcare” society. Besides coil technology with high magnetic field performance, superconducting electoronics or device technology, such as SQUID and SFQ-circuit, high temperature superconducting material and advanced cryogenics technology might be great significance in the history of nuclear fusion which requires so many wide, high and ultra technology. Superconducting technology seems to be the catalyst for a changing future society with nuclear fusion. As society changes, so will superconducting technology.

  15. 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.)

  16. 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.

  17. 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

  18. 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)

  19. Fusion--fission hybrid concepts for laser-induced fusion

    International Nuclear Information System (INIS)

    Maniscalco, J.

    1976-01-01

    Fusion-fission hybrid concepts are viewed as subcritical fission reactors driven and controlled by high-energy neutrons from a laser-induced fusion reactor. Blanket designs encompassing a substantial portion of the spectrum of different fission reactor technologies are analyzed and compared by calculating their fissile-breeding and fusion-energy-multiplying characteristics. With a large number of different fission technologies to choose from, it is essential to identify more promising hybrid concepts that can then be subjected to in-depth studies that treat the engineering safety, and economic requirements as well as the neutronic aspects. In the course of neutronically analyzing and comparing several fission blanket concepts, this work has demonstrated that fusion-fission hybrids can be designed to meet a broad spectrum of fissile-breeding and fusion-energy-multiplying requirements. The neutronic results should prove to be extremely useful in formulating the technical scope of future studies concerned with evaluating the technical and economic feasibility of hybrid concepts for laser-induced fusion

  20. 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

  1. 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)

  2. Discussion on Safety Analysis and Regulatory Framework for the Future Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Myoung-suk; Oh, Kyemin; Heo, Gyunyoung [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    This study aims to secure the core original technologies and expand the base of domestic specialist at a fusion area by pursuing and developing nonprocurement technologies for ITER. From this project, the latest technical data and experiences have been recorded for the development of the safety regulation and safety-related design criteria of the future fusion reactors in Korea. In this context, this paper discusses on the progress of surveying the ITER licensing process and regulatory issues revealed. The regulation and licensing process for a fusion power plant has been expected to be quite different due to unique and unforeseen properties differently from the conventional nuclear facilities. To overcome this, not only various safety issues should be analyzed, but safety objectives, regulatory requirements, and design variables should also be established in detailed design phase. We expect our survey will contribute on the discussion to establish general and technical safety principles for national fusion power plant technology plans.

  3. 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

  4. Fusion

    CERN Document Server

    Mahaffey, James A

    2012-01-01

    As energy problems of the world grow, work toward fusion power continues at a greater pace than ever before. The topic of fusion is one that is often met with the most recognition and interest in the nuclear power arena. Written in clear and jargon-free prose, Fusion explores the big bang of creation to the blackout death of worn-out stars. A brief history of fusion research, beginning with the first tentative theories in the early 20th century, is also discussed, as well as the race for fusion power. This brand-new, full-color resource examines the various programs currently being funded or p

  5. US/Japan collaborative program on fusion reactor materials: Summary of the tenth DOE/JAERI Annex I technical progress meeting on neutron irradiation effects in first wall and blanket structural materials

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1989-01-01

    This meeting was held at Oak Ridge National Laboratory on March 17, 1989, to review the technical progress on the collaborative DOE/JAERI program on fusion reactor materials. The purpose of the program is to determine the effects of neutron irradiation on the mechanical behavior and dimensional stability of US and Japanese austenitic stainless steels. Phase I of the program focused on the effects of high concentrations of helium on the tensile, fatigue, and swelling properties of both US and Japanese alloys. In Phase II of the program, spectral and isotropic tailoring techniques are fully utilized to reproduce the helium:dpa ratio typical of the fusion environment. The Phase II program hinges on a restart of the High Flux Isotope Reactor by mid-1989. Eight target position capsules and two RB* position capsules have been assembled. The target capsule experiments will address issues relating to the performance of austenitic steels at high damage levels including an assessment of the performance of a variety of weld materials. The RB* capsules will provide a unique and important set of data on the behavior of austenitic steels irradiated under conditions which reproduce the damage rate, dose, temperature, and helium generation rate expected in the first wall and blanket structure of the International Thermonuclear Experimental Reactor

  6. 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

  7. 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.)

  8. DEMO relevance of the test blanket modules in ITER-Application to the European test blanket modules

    International Nuclear Information System (INIS)

    Magnani, E.; Gabriel, F.; Boccaccini, L.V.; Li-Puma, A.

    2010-01-01

    Test blanket module (TBM) testing programme in ITER as a support to DEMO design is a very important step on the road map to commercial fusion reactors although it is an ambitious task. Finding as much as possible DEMO relevant tests in view of the future DEMO blanket design is therefore a major goal since ITER and DEMO environment and loading conditions are different. To clarify and quantify the meaning of the DEMO relevance, criteria using a structural, functional and behavioural representation of the breeding blanket acting as a system are investigated. Then, a three-step strategy is proposed to carry out TBM DEMO relevant tests associated with a TBM design modification strategy. Key parameters should intensively be used as target for TBM characterization and numerical code validation. When assessing the relevance, on the other hand, not only the actual difference between DEMO and ITER values should be considered, but also whether the analyzed phenomena have a threshold and a range of applicability, as numerical simulations are usually permitted within these limits. The proposed methodology is at the end applied to the design of the HCLL TBM breeding unit configuration.

  9. An Assessment of the Economics of Future Electric Power Generation Options and the Implications for Fusion

    International Nuclear Information System (INIS)

    Delene, J.G.; Hadley, S.; Reid, R.L.; Sheffield, J.; Williams, K.A.

    1999-01-01

    This study examines the potential range of electric power costs for some major alternatives to fusion electric power generation when it is ultimately deployed in the middle of the 21st century and, thus, offers a perspective on the cost levels that fusion must achieve to be competitive. The alternative technologies include coal burning, coal gasification, natural gas, nuclear fission, and renewable energy. The cost of electricity (COE) from the alternatives to fusion should remain in the 30-50 mils/kWh (1999 dollars) range of today in carbon sequestration is not needed, 30-60 mils/kWh if sequestration is required, or as high as 75 mils/kWh for the worst-case scenario for cost uncertainty. The reference COE range for fusion was estimated at 70-100 nmils/kWh for 1- to 1.3-GW(e) scale power plants. Fusion costs will have to be reduced and/or alternative concepts derived before fusion will be competitive with the alternatives for the future production of electricity. Fortunately, there are routes to achieve this goal

  10. 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.

  11. 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.

  12. The international thermonuclear experimental reactor and the future of nuclear fusion energy

    International Nuclear Information System (INIS)

    Pan Chuanhong

    2010-01-01

    Energy shortage and environmental problems are now the two largest challenges for human beings. Magnetic confinement nuclear fusion, which has achieved great progress since the 1990's, is anticipated to be a way to realize an ideal source of energy in the future because of its abundance, environmental compatibility, and zero carbon release. Exemplified by the construction of the International Thermonuclear Experimental Reactor (ITER), the development of nuclear fusion energy is now in its engineering phase, and should be realized by the middle of this century if all objectives of the ITER project are met. (author)

  13. Inertial fusion research at Lawrence Livermore National Laboratory: program status and future applications

    International Nuclear Information System (INIS)

    Meier, W.R.; Hogan, W.J.

    1986-01-01

    The objectives of the Lawrence Livermore National Laboratory (LLNL) Laser Fusion Program are to understand and develop the science and technology required to utilize inertial confinement fusion (ICF) for both military and commercial applications. The results of recent experiments are described. We point out the progress in our laser studies, where we continue to develop and test the concepts, components, and materials for present and future laser systems. While there are many potential commercial applications of ICF, we limit our discussions to electric power production

  14. Workshop on cold-blanket research

    International Nuclear Information System (INIS)

    1977-05-01

    The objective of the workshop was to identify and discuss cold-plasma blanket systems. In order to minimize the bombardment of the walls by hot neutrals the plasma should be impermeable. This requires a density edge-thickness product of nΔ > 10 15 cm -2 . An impermeable cold plasma-gas blanket surrounding a hot plasma core reduces the plasma wall/limiter interaction. Accumulation of impurities in this blanket can be expected. Fuelling from a blanket may be possible as shown by experimental results, though not fully explained by classical transport of neutrals. Refuelling of a reacting plasma had to be ensured by inward diffusion. Experimental studies of a cold impermeable plasma have been done on the tokamak-like Ringboog device. Simulation calculations for the next generation of large tokamaks using a particular transport model, indicate that the plasma edge profile can be controlled to reduce the production of sputtered impurities to an acceptable level. Impurity control requires a small fraction of the radial space to accomodate the cold-plasma layer. The problem of exhaust is, however, more complicated. If the cold-blanket scheme works as predicted in the model calculations, then α-particles generated by fusion will be transported to the cold outside layer. The Communities' experimental programme of research has been discussed in terms of the tokamaks which are available and planned. Two options present themselves for the continuation of cold-blanket research

  15. Nuclear fusion and its large potential for the future world energy supply

    Directory of Open Access Journals (Sweden)

    Ongena Jef

    2016-12-01

    Full Text Available An overview of the energy problem in the world is presented. The colossal task of ‘decarbonizing’ the current energy system, with ~85% of the primary energy produced from fossil sources is discussed. There are at the moment only two options that can contribute to a solution: renewable energy (sun, wind, hydro, etc. or nuclear fission. Their contributions, ~2% for sun and wind, ~6% for hydro and ~5% for fission, will need to be enormously increased in a relatively short time, to meet the targets set by policy makers. The possible role and large potential for fusion to contribute to a solution in the future as a safe, nearly inexhaustible and environmentally compatible energy source is discussed. The principles of magnetic and inertial confinement are outlined, and the two main options for magnetic confinement, tokamak and stellarator, are explained. The status of magnetic fusion is summarized and the next steps in fusion research, ITER and DEMO, briefly presented.

  16. Introducing the book 'Cold fusion and the future'

    Energy Technology Data Exchange (ETDEWEB)

    Rothwell, Jed

    2006-07-01

    Cold fusion will be the ideal source of energy, provided its introduction can be handled properly. A few cells have shown power density and temperatures suitable for real-world applications. Once these cells can be replicated on demand, commercial development will be straightforward. Manufacturing should not be too demanding, so thousands of companies will compete, and costs will fall quickly. The transition from fossil fuel to cold fusion will be rapid. Many extraordinary new applications will become possible, and seemingly intractable problems such as global warming may be fixed. Some examples will be presented. Public support is essential to funding research, and commercialization. 'Cold Fusion and the Future' is the title of a new book by this author. This paper discusses a few of the topics in the book.

  17. Recent developments in IFE safety and tritium research and considerations for future nuclear fusion facilities

    International Nuclear Information System (INIS)

    Reyes, Susana; Anklam, Tom; Meier, Wayne; Campbell, Patrick; Babineau, Dave; Becnel, James; Taylor, Craig; Coons, Jim

    2016-01-01

    licensing activities, and summarize our most recent thoughts on safety and tritium considerations for future nuclear fusion facilities.

  18. Recent developments in IFE safety and tritium research and considerations for future nuclear fusion facilities

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, Susana, E-mail: reyes20@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA (United States); Anklam, Tom; Meier, Wayne; Campbell, Patrick [Lawrence Livermore National Laboratory, Livermore, CA (United States); Babineau, Dave; Becnel, James [Savannah River National Laboratory, Aiken, SC (United States); Taylor, Craig; Coons, Jim [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-11-01

    licensing activities, and summarize our most recent thoughts on safety and tritium considerations for future nuclear fusion facilities.

  19. The challenge to keep nuclear fusion alive as a future energy source

    International Nuclear Information System (INIS)

    D'haeseleer, W.D.

    1999-01-01

    Few people are preoccupied with the energy issue. Indeed, inflation-corrected energy prices (in euros) are currently lower than before the first oil crisis of 1973; the annual growth rate of primary-energy use in the industrialized world has diminished considerably compared to before 1970, and oil and gas production is characterized by increased exploration activity and a wider geographical spread. Nevertheless, there is a real energy issue. If the greenhouse effect turns out to be real, then mankind should at least slow down the consumption of fossil fuels. Given the fact that world energy consumption (especially by the developing countries) will rise in the future, and that nuclear fission power has become unpopular in the western world, the idea reigning in some circles to cope with this situation by total reliance on energy savings and renewable energy sources comes close to wishful thinking. A realistic analysis makes it clear that there will be a need for large workhorses for electricity generation to keep the overall electricity grid sufficiently robust. From a global and long-term perspective, the logical conclusion is the following: because mankind cannot count on the continued use of fossil fuels (due to the finiteness of the resources combined with the possible climate change effects), our generation has the responsibility to develop alternative energy sources for the distant future. Many parallel lines of research and development therefore need be pursued; because of the uncertainties with other alternative sources, it would be irresponsible to kill some of these development lines. This holds for renewable sources, the nuclear fission breeder, and for nuclear fusion. A major hurdle for the survival of long term energy research and development is the liberalization of the electricity market. Because of the revolutionary changes taking place, utilities concentrate on cost cutting and short-term survival. In addition, they are no longer supposed to take

  20. Conceptual design of the blanket and power conversion system for a mirror hybrid fusion-fission reactor. Addendum 1. Alternate concepts. 12-month progress report addendum, July 1, 1975--June 30, 1976

    International Nuclear Information System (INIS)

    Schultz, K.R.; Dee, J.B.; Backus, G.A.; Culver, D.W.

    1976-01-01

    During the course of the Mirror Hybrid Fusion-Fission Reactor study several alternate concepts were considered for various reactor components. Several of the alternate concepts do appear to exhibit features with potential advantage for use in the mirror hybrid reactor. These are described and should possibly be investigated further in the future

  1. Fusion energy

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The efforts of the Chemical Technology Division in fusion energy include the areas of fuel handling, processing, and containment. Current studies are concerned largely with the development of vacuum pumps for fusion reactors and experiments and with development and evaluation of techniques for recovering tritium from solid or liquid breeding blankets. In addition, a small effort is devoted to support of the ORNL design of a major Tokamak experiment, The Next Step (TNS)

  2. The state of art of the manufacturing technology of FW blanket and the development of mock-up for fusion reactor in Russia

    International Nuclear Information System (INIS)

    Baek, Jong Hyuk; Jeong, Y. H.; Park, J. Y.; Kim, J. H.; Kim, H. G.

    2004-08-01

    In early 1990s, Russia had carried out the performance tests to verify the optimization of Be tile geometry and the bonding integrity of small mock-up using a HHF (High Heat Flux) test and an in-pile test in a research reactor. They had obtained the reliability of the brazing technologies for the Be/Cu bonding. And they had manufactured the near real-size large mock-up (about 0.8 mm in length) to find the bonding integrity by a fast brazing technique. They had a satisfied results from the HHF test for the large mock-up. Additionally, an alternative FW mock-ups, which were manufactured by both casting and fast brazing techniques to reduce the joining parts, showed a good joining performance from the HHF test. Therefore, it was concluded that the fast brazing techniques could be strongly recommended as a one of the preferable joining techniques and be possible to apply to joining for the Be/Cu joining of FW blanket

  3. Conceptual design of a fusion-fission hybrid reactor for transmutation of high level nuclear waste

    International Nuclear Information System (INIS)

    Qiu, L.J.; Wu, Y.C.; Yang, Y.W.; Wu, Y.; Luan, G.S.; Xu, Q.; Guo, Z.J.; Xiao, B.J.

    1994-01-01

    To assess the feasibility of the transmutation of long-lived radioactive waste using fusion-fission hybrid reactors, we are studying all the possible types of blanket, including a comparison of the thermal and fast neutron spectrum blankets. Conceptual designs of a small tokamak hybrid blanket with small inventory of actinides and fission products are presented. The small inventory of wastes makes the system safer. The small hybrid reactor system based on a fusion core with experimental parameters to be realized in the near future can effectively transmute actinides and fission products at a neutron wall loading of 1MWm -2 . An innovative energy system is also presented, including a fusion driver, fuel breeder, high level waste transmuter, fission reactor and so on. An optimal combination of all types of reactor is proposed in the system. ((orig.))

  4. Resonance self-shielding effect analysis of neutron data libraries applied for the dual-cooled waste transmutation blanket of the fusion-driven subcritical system

    International Nuclear Information System (INIS)

    Liu Haibo; Wu Yican; Zheng Shanliang; Zhang Chunzao

    2004-01-01

    Based on the Fusion-Driven Subcritical System (FDS-I), the 25 groups, 175 groups and 620 groups neutron nuclear data libraries with/without resonance self-shielding correction are made with the Njoy and Transx codes, and the K eff and reaction rates are calculated with the Anisn code. The conclusion indicates that the resonance self-shielding effect affects the reaction rates strongly. (authors)

  5. Heat transfer problems in gas-cooled solid blankets

    International Nuclear Information System (INIS)

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

    1976-01-01

    In all fusion reactors using the deuterium-tritium fuel cycle, a large fraction approximately 80 percent of the fusion energy will be released as approximately 14 MeV neutrons which must be slowed down in a relatively thick blanket surrounding the plasma, thereby, converting their kinetic energy to high temperature heat which can be continuously removed by a coolant stream and converted in part to electricity in a conventional power turbine. Because of the primary goal of achieving minimum radioactivity, to date Brookhaven blanket concepts have been restricted to the use of some form of solid lithium, with inert gas-cooling and in some design cases, water-cooling of the shell structure. Aluminum and graphite have been identified as very promising structural materials for fusion blankets, and conceptual designs based on these materials have been made. Depending on the thermal loading on the ''first'' wall which surrounds the plasma as well as blanket design, heat transfer problems may be noticeably different in gas-cooled solid blankets. Approaches to solution of heat removal problems as well as explanation of: (a) the after-heat problems in blankets; (b) tritium breeding in solids; and (c) materials selection for radiation shields relative to the minimum activity blanket efforts at Brookhaven are discussed

  6. Nuclear fusion - Inexhaustible source of energy for tomorrow

    International Nuclear Information System (INIS)

    Leiser, M.; Demchenko, V.

    1989-09-01

    The purpose of this paper is to provide a general description of nuclear fusion as an energy option for the future and to clarify to some extent the various issues - scientific, technological, economic and environmental - which are likely to be relevant to controlled thermonuclear fusion. Section 1 describes the world energy problem and some advantages of nuclear fusion compared to other energy options. Sections 2 and 3 describe the fundamentals of fusion energy, plasma confinement, heating and technological aspects of fusion researches. Some plasma confinement schemes (tokamak, stellarator, inertial confinement fusion) are described. The main experimental results and parameter devices are cited to illustrate the state of the art as of 1989. Various engineering problems associated with reactor design, magnetic systems, materials, plasma purity, fueling, blankets, environment, economics and safety are discussed. A description of both bilateral and multilateral efforts in fusion research under the auspices of the IAEA is presented in Section 4. (author). 11 refs, 4 figs, 1 tab

  7. Modelling the power conversion unit of a generic nuclear fusion plant, with a dual coolant blanket and a supercritical CO2 power cycle, by means of RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Batet, L.

    2015-07-01

    In the framework of the Spanish fusion program TECNO-FUS, a dual coolant blanket design was proposed for DEMO. A generic power conversion system (supercritical recompression CO2 cycle) based on this proposal has been simulated using RELAP5-3D, a multipurpose system thermal-hydraulic code developed by the Idaho National Laboratory (USA). The code allows the dynamic simulation of thermal-hydraulic systems, including the control features. A model has been set up by assembling the available RELAP5-3D components: pipe, branch, pump, compressor, turbine, etc. Thermal fluxes between fluids in heat exchangers are simulated by means of heat structures, which are used as well to simulate the heating from plasma. A number of control features have been designed for the simulated plant, and their parameters have been adjusted. The code is then able to simulate robustly the dynamics of the system with a few boundary conditions. This paper exemplifies the usefulness of the code and model to understand the behavior of the plant and to perform sensitivity analyses of the control parameters or other design features. (Author)

  8. Mirror reactor blankets

    International Nuclear Information System (INIS)

    Lee, J.D.; Barmore, W.L.; Bender, D.J.; Doggett, J.N.; Galloway, T.R.

    1976-01-01

    The general requirements of a breeding blanket for a mirror reactor are described. The following areas are discussed: (1) facility layout and blanket maintenance, (2) heat transfer and thermal conversion system, (3) materials, (4) tritium containment and removal, and (5) nuclear performance

  9. Fusion energy

    International Nuclear Information System (INIS)

    Gross, R.A.

    1984-01-01

    This textbook covers the physics and technology upon which future fusion power reactors will be based. It reviews the history of fusion, reaction physics, plasma physics, heating, and confinement. Descriptions of commercial plants and design concepts are included. Topics covered include: fusion reactions and fuel resources; reaction rates; ignition, and confinement; basic plasma directory; Tokamak confinement physics; fusion technology; STARFIRE: A commercial Tokamak fusion power plant. MARS: A tandem-mirror fusion power plant; and other fusion reactor concepts

  10. Fusion reactor problems

    International Nuclear Information System (INIS)

    Carruthers, R.

    It is pointed out that plasma parameters for a fusion reactor have been fairly accurately defined for many years, and the real plasma physics objective must be to find the means of achieving and maintaining these specifiable parameters. There is good understanding of the generic technological problems: breading blankets and shields, radiation damage, heat transfer and methods of magnet design. The required plasma parameters for fusion self-heated reactors are established at ntausub(E) approximately 2.10 14 cm -3 sec, plasma radius 1.5 to 3 m, wall loading 5 to 10 MW cm -2 , temperature 15 keV. Within this model plasma control by quasi-steady burn as a key problem is studied. It is emphasized that the future programme must interact more closely with engineering studies and should concentrate upon research which is relevant to reactor plasmas. (V.P.)

  11. Blanket testing in NET

    International Nuclear Information System (INIS)

    Chazalon, M.; Daenner, W.; Libin, B.

    1989-01-01

    The testing stages in NET for the performance assessment of the various breeding blanket concepts developed at the present time in Europe for DEMO (LiPb and ceramic blankets) and the requirements upon NET to perform these tests are reviewed. Typical locations available in NET for blanket testing are the central outboard segments and the horizontal ports of in-vessel sectors. These test positions will be connectable with external test loops. The number of test loops (helium, water, liquid metal) will be such that each major class of blankets can be tested in NET. The test positions, the boundary conditions and the external test loops are identified and the requirements for test blankets are summarized (author). 6

  12. Test Blanket Working Group's recent activities

    International Nuclear Information System (INIS)

    Vetter, J.E.

    2001-01-01

    The ITER Test Blanket Working Group (TBWG) has continued its activities during the period of extension of the EDA with a revised charter on the co-ordination of the development work performed by the Parties and by the JCT leading to a co-ordinated test programme on ITER for a DEMO-relevant tritium breeding blanket. This follows earlier work carried out until July 1998, which formed part of the ITER Final Design Report (FDR), completed in 1998. Whilst the machine parameters for ITER-FEAT have been significantly revised compared to the FDR, testing of breeding blanket modules remains a main objective of the test programme and the development of a reactor-relevant breeding blanket to ensure tritium fuel self-sufficiency is recognized a key issue for fusion. Design work and R and D on breeding blanket concepts, including co-operation with the other Contacting Parties of the ITER-EDA for testing these concepts in ITER, are included in the work plans of the Parties

  13. 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.)

  14. Fusion option to dispose of spent nuclear fuel and transuranic elements

    International Nuclear Information System (INIS)

    Gohar, Y.

    2000-01-01

    The fusion option is examined to solve the disposition problems of the spent nuclear fuel and the transuranic elements. The analysis of this report shows that the top rated solution, the elimination of the transuranic elements and the long-lived fission products, can be achieved in a fusion reactor. A 167 MW of fusion power from a D-T plasma for sixty years with an availability factor of 0.75 can transmute all the transuranic elements and the long-lived fission products of the 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. The operating time can be reduced to thirty years with use of 334 MW of fusion power, a system study is needed to define the optimum time. In addition, the fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future. Fusion blankets with a liquid carrier for the transuranic elements can achieve a transmutation rate for the transuranic elements up to 80 kg/MW.y of fusion power with k eff of 0.98. In addition, the liquid blankets have several advantages relative to the other blanket options. The energy from this transmutation is utilized to produce revenue for the system. Molten salt (Flibe) and lithium-lead eutectic are identified as the most promising liquids for this application, both materials are under development for future fusion blanket concepts. The Flibe molten salt with transuranic elements was developed and used successfully as nuclear fuel for the molten salt breeder reactor in the 1960's

  15. 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

  16. 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.

  17. Telescope-based cavity for negative ion beam neutralization in future fusion reactors.

    Science.gov (United States)

    Fiorucci, Donatella; Hreibi, Ali; Chaibi, Walid

    2018-03-01

    In future fusion reactors, heating system efficiency is of the utmost importance. Photo-neutralization substantially increases the neutral beam injector (NBI) efficiency with respect to the foreseen system in the International Thermonuclear Experimental Reactor (ITER) based on a gaseous target. In this paper, we propose a telescope-based configuration to be used in the NBI photo-neutralizer cavity of the demonstration power plant (DEMO) project. This configuration greatly reduces the total length of the cavity, which likely solves overcrowding issues in a fusion reactor environment. Brought to a tabletop experiment, this cavity configuration is tested: a 4 mm beam width is obtained within a ≃1.5  m length cavity. The equivalent cavity g factor is measured to be 0.038(3), thus confirming the cavity stability.

  18. 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

  19. Main features and potentialities of gas-blanket systems

    International Nuclear Information System (INIS)

    Lehnert, B.

    1977-02-01

    A review is given of the features and potentialities of cold-blanket systems, with respect to plasma equilibrium, stability, and reactor technology. The treatment is concentrated on quasi-steady magnetized plasmas confined at moderately high beta values. The cold-blanket concept has specific potentialities as a fusion reactor, e.g. in connection with the desired densities and dimensions of full-scale systems, refuelling, as well as ash and impurity removal, and stability. (author)

  20. Applications of the Aqueous Self-Cooled Blanket concept

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.J.; Varsamis, G.; Wrisley, K.; Deutch, L.; Gierszewski, P.

    1986-01-01

    In this paper a novel water-cooled blanket concept is examined. This concept, designated the Aqueous Self-Cooled Blanket (ASCB), employs water with small amounts of dissolved fertile compounds as both the coolant and the breeding medium. The ASCB concept is reviewed and its application in three different contexts is examined: (1) power reactors; (2) near-term devices such as NET; and (3) fusion-fission hybrids

  1. Fusion technology programme

    International Nuclear Information System (INIS)

    Finken, D.

    1986-05-01

    In 1982, KfK joined the fusion programme of EURATOM as a further association introducing its experience in nuclear technology. KfK closely cooperates with IPP Garching, the two institutions forming a research unit aiming at planning and realization of future development steps of fusion. KfK has combined its forces in the Nuclear Fusion Project (PKF) with participation of several KfK departments to the project tasks. Previous work of KfK in magnetic fusion has addressed mainly superconducting magnets, plasma heating by cluster ions and studies on structural materials. At present, emphasis of our work has concentrated increasingly on the nuclear part, i.e. the first wall and blanket structures and the elements of the tritium extraction and purification system. Associated to this component development are studies of remote maintenance and safety. Most of the actual work addresses NET, the next step to a demonstration of fusion feasibility. NET is supposed to follow JET, the operating plasma physics experiment of Euratom, on the 1990's. Detailed progress of the work in the past half year is described in this report. (orig./GG)

  2. Molten salt cooling/17Li-83Pb breeding blanket concept

    International Nuclear Information System (INIS)

    Sze, D.K.; Cheng, E.T.

    1985-02-01

    A description of a fusion breeding blanket concept using draw salt coolant and static 17 Li- 83 Pb is presented. 17 Li- 83 Pb has high breeding capability and low tritium solubility. Draw salt operates at low pressure and is inert to water. Corrosion, MHD, and tritium containment problems associated with the MARS design are alleviated because of the use of a static LiPb blanket. Blanket tritium recovery is by permeation toward the plasma. A direct contact steam generator is proposed to eliminate some generic problems associated with a tube shell steam generator

  3. 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.

  4. Status of fusion technology development in JAERI stressing steady-state operation for future reactors

    International Nuclear Information System (INIS)

    Matsuda, Shinzaburo

    2000-01-01

    This paper reports on the progress of the fusion reactor technologies developed at the Japan Atomic Energy Research Institute (JAERI) and expected to lead to a future steady state operation reactor. In particular, superconducting coil technology for plasma confinement, NBI and RF systems technology for plasma control and current drive, fueling and pumping systems technology for particle control, heat removal technology, and development of long life materials are highlighted as the important key elements for the future steady state operation. It will be discussed how these key technologies have already been developed by the ITER (International Thermonuclear Experimental Reactor) technology R and D as well as by the Japanese domestic program, and which technologies are planned for the near future

  5. The lithium blanket program at the LOTUS facility

    International Nuclear Information System (INIS)

    File, J.; Haldy, P.A.; Quanci, J.

    1987-01-01

    An experimental program of neutron transport studies of the lithium Blanket Module (LBM) carried out with the LOTUS point-neutron source at the Ecole Polytechnique Federale de Lausanne (EPTL), Switzerland has been concluded. The major objectives of this program are to perform a series of neutron transport and tritium breeding experiments to qualify the LBM for future experiments on toroidal fusion devices such as TFTR to perform neutron multiplier experiments on the LBM employing various materials in a removable slab geometry; and, to compare the experimental results of radiation dosimetry and tritium breeding with the calculations of two and three dimensional neutron transport codes. An overview of the results from the radiation dosimetry and tritium assay are presented and compared to the two and three dimensional neutron transport codes

  6. 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

  7. 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.

  8. The evolution of US helium-cooled blankets

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.

    1991-01-01

    This paper reviews and compares four helium-cooled fusion reactor blanket designs. These designs represent generic configurations of using helium to cool fusion reactor blankets that were studied over the past 20 years in the United States of America (US). These configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. Among these four designs, the nested shell design, which was invented for the ARIES study, is the simplest in configuration and has the least number of critical issues. Both metallic and ceramic-composite structural materials can be used for this design. It is believed that the nested shell design can be the most suitable blanket configuration for helium-cooled fusion power and experimental reactors. (orig.)

  9. Target technologies for laser inertial confinement fusion: state-of-the-art and future perspective

    International Nuclear Information System (INIS)

    Zhang Lin; Du Kai

    2013-01-01

    Targets are physical base of the laser inertial confinement fusion (ICF) researches. The quality of the targets has extremely important influences on the reliabilities and degree of precision of the ICF experimental results. The characteristics of the ICF targets, such as complexity and microscale, high precision, determine that the target fabrication process must be a system engineering. This paper presents progresses on the fabrication technologies of ICF targets. The existing problem and the future needs of ICF target fabrication technologies are also discussed. (authors)

  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. 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

  12. Blankets for thermonuclear device

    International Nuclear Information System (INIS)

    Maki, Koichi; Fukumoto, Hideshi.

    1986-01-01

    Purpose: To produce tritium more than consumed, through thermonuclear reaction. Constitution: The energy spectrum of neutron generated by neutron multiplying reaction in a neutron multiplying blanket and moderated neutrons has a large ratio in a low energy section. In the low-energy absorption region of stainless steel which is a material of cooling pipes constituting a neutron multiplying blanket cooling channel, the neutrons are absorbed, lessening the neutron multiplying effect. To prevent this, the neutron multiplying blanket cooling channel is covered with tritium breeding blankets, thereby enabling the production of a substantially great amount of tritium more than the amount of tritium to be consumed by the thermonuclear reaction by preventing neutron absorption by the component materials of the cooling channel, improving the tritium breeding ratio by 20 to 25 %, and increasing the efficiency of use of neutrons for tritium generation. (Horiuchi, T.)

  13. Methodology and analysis for effects of energy and angular distributions of secondary neutrons in fusion blankets and application to integral beryllium experiments

    International Nuclear Information System (INIS)

    Song, P.M.

    1990-01-01

    The main objective of the US/JAERI (Japan Atomic Energy Research Institute) collaborative experiment program on Fusion Breeder Neutronics is to estimate the uncertainties involved in predicting the TBR in Li 2 O. Beryllium has been used as a neutron multiplier in several experiments performed in that program. The shape of the C/E values (calculation/experiment) for the tritium production rate (TPR) from 6 Li, T 6 observed in these experiments indicate that there is underestimation in T 6 just behind the Be layer. This feature could be related to the Be cross-sections, especially in secondary energy (SED) and angular distribution (SAD) of emitted neutrons from reaction. These SED and SAD of the 9 Be(n,2n) cross-sections are subject to large uncertainties because of inadequate representation of the energy/angle distribution of the emitted neutrons. To access the uncertainty in predicting TPR that results from the current uncertainties in the cross-section data and SED/SAD of emitted neutrons, an extensive two-dimensional sensitivity/uncertainty analysis was performed with the current FORSS module, and new codes, JULIX, VARIX, and UNGSS which were specifically developed to incorporate the new methodology of the present work in treating SED/SAD sensitivity analyses. When analyses were performed, it was found that the local standard deviation in T 6 is 2.1 to 9.3% in the integrated cross-section, 5.2 to 11.2% in the SED (direct variation with the ENDF/BLANL), and 0.14% to 1.37% in the SAD, respectively, with largest uncertainties occurring inside the beryllium layer. The uncertainty in T 6 is mainly attributed to the current uncertainties in 9 Be(n,elastic), 16 O and 7 Li data

  14. Assessment of alkali metal coolants for the ITER blanket

    International Nuclear Information System (INIS)

    Natesan, K.; Reed, C.B.; Mattas, R.F.

    1994-01-01

    The blanket system is one of the most important components of a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The Blanket Comparison and Selection Study, conducted earlier, described the overall comparative performance of different blanket concepts, including liquid metal, molten salt, water, and helium. This paper will discuss the ITER requirements for a self-cooled blanket concept with liquid lithium and for indirectly cooled concepts that use other alkali metals such as NaK. The paper will address the thermodynamics of interactions between the liquid metals (i.e., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data will be used to assess the long-term performance of the first wall in a liquid metal environment

  15. Status of the EU test blanket systems safety studies

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

    2015-01-01

    Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

  16. Status of the EU test blanket systems safety studies

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

    2015-10-15

    Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

  17. Limitation of fusion power plant installation on future power grids under the effect of renewable and nuclear power sources

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Shutaro, E-mail: takeda.shutarou.55r@st.kyoto-u.ac.jp [Graduate School of Advanced Integrated Studies in Human Survivability, Kyoto University, Kyoto, Kyoto (Japan); Sakurai, Shigeki [Graduate School of Advanced Integrated Studies in Human Survivability, Kyoto University, Kyoto, Kyoto (Japan); Yamamoto, Yasushi [Faculty of Engineering Science, Kansai University, Suita, Osaka (Japan); Kasada, Ryuta; Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Uji, Kyoto (Japan)

    2016-11-01

    Graphical abstract: - Highlights: • Future power grids would be unstable due to renewable and nuclear power sources. • Output interruptions of fusion plant would cause disturbances to future grids. • Simulation results suggested they would create limitations in fusion installation. • A novel diagram was presented to illustrate this suggested limitation. - Abstract: Future power grids would be unstable because of the larger share of renewable and nuclear power sources. This instability might bring some additional difficulties to fusion plant installation. Therefore, the authors carried out a quantitative feasibility study from the aspect of grid stability through simulation. Results showed that the more renewable and nuclear sources are linked to a grid, the greater disturbance the grid experiences upon a sudden output interruption of a fusion power plant, e.g. plasma disruption. The frequency deviations surpassed 0.2 Hz on some grids, suggesting potential limitations of fusion plant installation on future grids. To clearly show the suggested limitations of fusion plant installations, a novel diagram was presented.

  18. Overview of the TFTR Lithium Blanket Module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

    The LBM (Lithium Blanket Module) is an approximately cubic module, about 80 cm on each side, with construction representative of a helium-cooled lithium oxide fusion reactor blanket module. Measurements of neutron transport and tritium breeding in the LBM will be made in irradiation programs first with a point-neutron source, and subsequently with the D-D and D-T fusion-neutron sources of the TFTR. This paper summarizes the objectives of the LBM program, the design, development and construction of the LBM, and progress in the experimental tests

  19. Conceptual design of a fission-based integrated test facility for fusion reactor components

    International Nuclear Information System (INIS)

    Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

    1982-01-01

    The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

  20. Identification of future engineering-development needs of alternative concepts for magnetic-fusion energy

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1982-01-01

    A qualitative identification of future engineering needs of alternative fusion concepts (AFCs) is presented. These needs are assessed relative to the similar needs of the tokamak in order to emphasize differences in required technology with respect to the well documented mainline approach. Although nearly thirty AFCs can be identified as being associated with some level of reactor projection, redirection, refocusing, and general similarities can be used to generate a reduced AFC list that includes only the bumpy tori, stellarators, reversed-field pinches, and compact toroids. Furthermore, each AFC has the potential of operating as a conventional (low power density, superconducting magnets) or a compact, high-power-density (HPD) system. Hence, in order to make tractable an otherwise difficult task, the future engineering needs for the AFCs are addressed here for conventional versus compact approaches, with the latter being treated as a generic class and the former being composed of bumpy tori, stellarators, reversed-field pinches, and compact toroids

  1. 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.

  2. Controlled thermonuclear fusion power apparatus and method

    International Nuclear Information System (INIS)

    Bussard, R.W.; Coppi, B.

    1982-01-01

    This invention provides a modular fusion reactor system containing several fusion power cores, each of relatively small size and low cost. Energy from the cores is absorbed in the core structure and within a surrounding blanket, and the cores themselves may be individually removed from the blanket and replaced as they deteriorate from high radiation flux damage

  3. 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

  4. 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

  5. Fusion technology 1998

    International Nuclear Information System (INIS)

    Beaumont, B.; Libeyre, P.; Gentile, B. de; Tonon, G.

    1998-01-01

    The Symposium On Fusion Technology (SOFT) is held every two years with the objective to set the stage for the exchange of information on the design, construction and operation of fusion experiments and on the technology which is being developed for the next step devices and fusion reactors. By decision of the International Organizing Committee, the 20. SOFT includes invited talks, and oral and poster contributions in the following topics: plasma facing components, plasma heating and current drive, plasma engineering and control, experimental systems and diagnostics, magnets and power supplies, fuel technologies, remote operation, blanket and shield technologies, safety and environment, and system engineering and future devices. This symposium differs from the previous ones of this series by the way the present proceedings are produced. In order to have the written material available to the participants and the community at the nearest to the conference event, the papers have been collected 2 months in advance and printed in the present books. The goal was to deliver them to each participant upon arrival to the conference centre. These books contain all the papers corresponding to poster presentation, and the abstracts of the oral contributions and invited papers. The papers corresponding to these presentations, both oral and invited, will be published in 1999, after a standard review process, in a supplement of Fusion Engineering and Design. (author)

  6. Review of fusion synfuels

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1980-01-01

    Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Depending on design, electric generation efficiencies of approx. 40 to 60% and hydrogen production efficiencies by high-temperature electrolysis of approx. 50 to 65% are projected for fusion reactors using high-temperatures blankets. Fusion/coal symbiotic systems appear economically promising for the first generation of commercial fusion synfuels plants. Coal production requirements and the environmental effects of large-scale coal usage would be greatly reduced by a fusion/coal system. In the long term, there could be a gradual transition to an inexhaustible energy system based solely on fusion

  7. 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)

  8. Heat-pipe liquid-pool-blanket concept for the Tandem Mirror Reactor

    International Nuclear Information System (INIS)

    Hoffman, M.A.; Werner, R.W.; Johnson, G.L.

    1981-01-01

    The blanket concept for the tandem mirror reactor described in this paper was developed to produce the medium temperature heat (approx. 850 to 950 K) for the General Atomic sulfur-iodine thermochemical process for producing hydrogen. This medium temperature heat from the blanket constitutes about 81% of the total power output of the fusion reactor

  9. The Sustainable Nuclear Future: Fission and Fusion E.M. Campbell Logos Technologies

    Science.gov (United States)

    Campbell, E. Michael

    2010-02-01

    Global industrialization, the concern over rising CO2 levels in the atmosphere and other negative environmental effects due to the burning of hydrocarbon fuels and the need to insulate the cost of energy from fuel price volatility have led to a renewed interest in nuclear power. Many of the plants under construction are similar to the existing light water reactors but incorporate modern engineering and enhanced safety features. These reactors, while mature, safe and reliable sources of electrical power have limited efficiency in converting fission power to useful work, require significant amounts of water, and must deal with the issues of nuclear waste (spent fuel), safety, and weapons proliferation. If nuclear power is to sustain its present share of the world's growing energy needs let alone displace carbon based fuels, more than 1000 reactors will be needed by mid century. For this to occur new reactors that are more efficient, versatile in their energy markets, require minimal or no water, produce less waste and more robust waste forms, are inherently safe and minimize proliferation concerns will be necessary. Graphite moderated, ceramic coated fuel, and He cooled designs are reactors that can satisfy these requirements. Along with other generation IV fast reactors that can further reduce the amounts of spent fuel and extend fuel resources, such a nuclear expansion is possible. Furthermore, facilities either in early operations or under construction should demonstrate the next step in fusion energy development in which energy gain is produced. This demonstration will catalyze fusion energy development and lead to the ultimate development of the next generation of nuclear reactors. In this presentation the role of advanced fission reactors and future fusion reactors in the expansion of nuclear power will be discussed including synergies with the existing worldwide nuclear fleet. )

  10. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Naruse, Yuji; Yamaoka, Mitsuaki; Ohara, Atsushi; Ono, Kiyoshi; Kobayashi, Shigetada.

    1992-06-01

    Aqueous solution blanket using lithium salts such as LiNO 3 and LiOH have been studied in the US-TIBER program and ITER conceptual design activity. In the JAERI/LANL collaboration program for the joint operation of TSTA (Tritium Systems Test Assembly), preliminary design work of blanket tritium system for lithium ceramic blanket, aqueous solution blanket and liquid metal blanket, have been performed to investigate technical feasibility of tritium demonstration tests using the TSTA. Detail study of the aqueous solution blanket concept have not been performed in the Japanese fusion program, so that this study was carried out to investigate features of its concept and to evaluated its technical problems. The following are the major items studied in the present work: (i) Neutronics of tritium breeding ratio and shielding performance Lithium concentration, Li-60 enrichment, beryllium or lead, composition of structural material/beryllium/solution, heavy water, different lithium-salts (ii) Physicochemical properties of salts Solubility, corrosion characteristics and compatibility with structural materials, radiolysis (iii) Estimation of radiolysis in ITER aqueous solution blanket. (author)

  11. Perspective on the fusion-fission energy concept

    International Nuclear Information System (INIS)

    Liikala, R.C.; Perry, R.T.; Teofilo, V.L.

    1978-01-01

    A concept which has potential for near-term application in the electric power sector of our energy economy is combining fusion and fission technology. The fusion-fission system, called a hybrid, is distinguished from its pure fusion counterpart by incorporation of fertile materials (uranium or thorium) in the blanket region of a fusion machine. The neutrons produced by the fusion process can be used to generate energy through fission events in the blanket or produce fuel for fission reactors through capture events in the fertile material. The performance requirements of the fusion component of hybrids is perceived as being less stringent than those for pure fusion electric power plants. The performance requirements for the fission component of hybrids is perceived as having been demonstrated or could be demonstrated with a modest investment of research and development funds. This paper presents our insights and observations of this concept in the context of why and where it might fit into the picture of meeting our future energy needs. A bibliography of hybrid research is given

  12. Blast venting through blanket material in the HYLIFE ICF reactor

    International Nuclear Information System (INIS)

    Liu, J.C.; Peterson, P.F.; Schrock, V.E.

    1992-01-01

    This work presents a numerical study of blast venting through various blanket configurations in the HYLIFE ICF reactor design. The study uses TSUNAMI -- a multi-dimensional, high-resolution, shock capturing code -- to predict the momentum exchange and gas dynamics for blast venting in complex geometries. In addition, the study presents conservative predictions of wall loading by gas shock and impulse delivered to the protective liquid blanket. Configurations used in the study include both 2700 MJ and 350 MJ fusion yields per pulse for 5 meter and 3 meter radius reactor chambers. For the former, an annular jet array is used for the blanket geometry, while in the latter, both annular jet array as well as slab geometries are used. Results of the study indicate that blast venting and wall loading may be manageable in the HYLIFE-II design by a judicious choice of blanket configuration

  13. Imploding-liner reactor nucleonic studies: the LINUS blanket

    International Nuclear Information System (INIS)

    Dudziak, D.J.

    1977-09-01

    Scoping nucleonic studies have been performed for a small imploding-liner fusion reactor concept. Tritium breeding ratio and time-dependent energy deposition rates were the primary parameters of interest in the study. Alloys of Pb and LiPb were considered for the liquid liner (blanket), and tritium breeding was found to be more than adequate with blankets less than 1 m thick. However, neutron leakages into the solid cylinder block surrounding the liquid liner are generally quite high, so considerable effort was concentrated on minimizing these values. Time-dependent calculations reveal that 89% of the energy is deposited in the blanket within 2 μs. Thus, LINUS's blanket should remain intact for the requisite neutron and gamma-ray lifetimes

  14. Future on the ITER program. On a branch of research on nuclear fusion

    International Nuclear Information System (INIS)

    Masaike, Akira

    2000-01-01

    As a huge cost for research and development of nuclear fusion is required, some international cooperative research such as ITER program have been intended to promote, to which Japanese response is required. As the program can be understood on its meaning at a viewpoint of promotion of basic science, concept on a key of energy problem is not insufficient yet And, its effect on technical problems and environment cannot be neglected Here was shown some proposals necessity for discussion on how the program had to be promoted under consideration of these problems. When a large scale program consuming national budget will be carried out, it is natural that agreement of national peoples must be obtained. Regretfully, in Japan discussion on science program above all nuclear policy has scarcely been experienced at citizens' levels, and some bitter experiences, where the concerned have promoted it in one side under a concept without any change once decided, have been pressured without any response to scientific advancements and social changes. Therefore, future plan on the nuclear fusion must be carried out a number of thorough discussion at a wide range from various viewpoints such as its realizing feasibility, safety, economics, and so forth, to promote careful adaptabilities. And, the concerned under promotion of the program and the relatives in the academic community seem to have a responsibility to easily explain present condition and scope of the plan to not only scientists but also citizens to awake them to promote its discussion with them. (G.K.)

  15. Spatiotemporal Fusion of Multisource Remote Sensing Data: Literature Survey, Taxonomy, Principles, Applications, and Future Directions

    Directory of Open Access Journals (Sweden)

    Xiaolin Zhu

    2018-03-01

    Full Text Available Satellite time series with high spatial resolution is critical for monitoring land surface dynamics in heterogeneous landscapes. Although remote sensing technologies have experienced rapid development in recent years, data acquired from a single satellite sensor are often unable to satisfy our demand. As a result, integrated use of data from different sensors has become increasingly popular in the past decade. Many spatiotemporal data fusion methods have been developed to produce synthesized images with both high spatial and temporal resolutions from two types of satellite images, frequent coarse-resolution images, and sparse fine-resolution images. These methods were designed based on different principles and strategies, and therefore show different strengths and limitations. This diversity brings difficulties for users to choose an appropriate method for their specific applications and data sets. To this end, this review paper investigates literature on current spatiotemporal data fusion methods, categorizes existing methods, discusses the principal laws underlying these methods, summarizes their potential applications, and proposes possible directions for future studies in this field.

  16. Overview of the CTR blanket engineering research program at the University of Tokyo

    International Nuclear Information System (INIS)

    Nakazawa, Masaharu; Madarame, Haruki; Takahashi, Yoichi; Takagi, Toshiyuki

    1989-01-01

    A small overview has been given on the fusion reactor blanket engineering research program at the University of Tokyo as an introduction to the following articles, especially in its history, organization, experimental facilities and ten years research activity. (orig.)

  17. 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

  18. Major achievements of the European shield blanket R and D during the ITER EDA, and their relevance for future next step machines

    Energy Technology Data Exchange (ETDEWEB)

    Daenner, W. E-mail: daenner@ipp.mpg.de; Cardella, A.; Jones, L.; Lorenzetto, P.; Maisonnier, D.; Malavasi, G.; Peacock, A.; Rodgers, E.; Tavassoli, F

    2000-11-01

    In the frame of the international thermonuclear experimental reactors (ITER) collaboration, the European home team (EU HT) has committed significant efforts on the R and D for the Shield Blanket. This paper summarises the main achievements of this programme, which have been obtained over the last 7 years. The depth of R and D extends from generic activities up to the manufacture of prototypes, but has, in accordance with the design progress, reached different stages of maturity for the various components. New ITER options being considered since early 1998 have not made these activities irrelevant. With few exceptions, the results are still applicable for less ambitious next step machines, or transferable to components with similar functions or requirements.

  19. 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

  20. Need for research and development in fusion: Economical energy for a sustainable future with low environmental impact

    International Nuclear Information System (INIS)

    Logan, B.G.; Perkins, L.J.; Moir, R.W.; Ryutov, D.D.

    1995-01-01

    Fusion, advanced fission, and solar-electric plants are the only unlimited nonfossil options for a sustainable energy future for the world. Fusion poses the only indigenous fuel reserve that will last as long as the earth itself lasts. However, continued innovation and diversity in fusion R ampersand D will be required to meet its economic goal. The long-term nature of fusion research means that the required R ampersand D investment will not come from the private sector. However, once fusion is realized commercially, the dividend for humanity will be profound in terms of the welfare of the global community. We should also not underestimate the huge potential export opportunities that would then open up for industry. Federal energy R ampersand D at nearly 1% of U.S. energy costs is prudent and justified to allow pursuit of all three primary energy options for a sustainable energy future. Multiple parallel paths are essential to ensure success. The projected timescale for significant shortfalls in world energy supply to become apparent is nearly 30 to 40 yr depending on assumptions. The time to develop fusion from near-term R ampersand D through significant commercial market penetration is at least of the same order, so its development must not be delayed. 6 refs., 2 figs

  1. 'Maintain FENDL library for fusion applications'. FENDL-2 library for fusion applications - Status and future developments. Summary report

    International Nuclear Information System (INIS)

    Forrest, R.; Trkov, A.

    2003-11-01

    The discussions and conclusions of the meeting to 'Maintain FENDL library for Fusion Applications' are summarized in this report. A presentation was made by each of the participants, followed by a review of FENDL-2: evaluations and recommendations, and discussions on the special purpose libraries and processed files, with relevant further action thereon being determined. (author)

  2. High temperature blankets for the production of synthetic fuels

    International Nuclear Information System (INIS)

    Powell, J.R.; Steinberg, M.; Fillo, J.; Makowitz, H.

    1977-01-01

    The application of very high temperature blankets to improved efficiency of electric power generation and production of H 2 and H 2 based synthetic fuels is described. The blanket modules have a low temperature (300 to 400 0 C) structure (SS, V, Al, etc.) which serves as the vacuum/coolant pressure boundary, and a hot (>1000 0 C) thermally insulated interior. Approximately 50 to 70% of the fusion energy is deposited in the hot interior because of deep penetration by high energy neutrons. Separate coolant circuits are used for the two temperature zones: water for the low temperature structure, and steam or He for the hot interior. Electric generation efficiencies of approximately 60% and H 2 production efficiencies of approximately 50 to 70%, depending on design, are projected for fusion reactors using these high temperature blankets

  3. Integrity of the first wall in fusion reactors

    International Nuclear Information System (INIS)

    Kurihara, Ryoichi

    2004-07-01

    Future fusion power reactors DREAM and A-SSTR2, which have been conceptually designed in the Japan Atomic Energy Research Institute, use the SiC/SiC composite material as the first wall of the blanket because of its characteristics of high heat-resistance and low radiation material. DEMO reactor, which was conceptually designed in 2001, uses the low activation ferritic steel as the first-wall material of the blanket. The problems in the thermal structural design of the plasma facing component such as the blanket first wall and the divertor plate which receives very high heat flux were examined in the design of the fusion power reactors. Compact high fusion power reactor must give high heat flux and high-speed neutron flux from the plasma to the first wall and the divertor plate. In this environmental situation, the micro cracks should be generated in material of the first wall. Structural integrity of the first wall would be very low during the operation of the reactor, if those micro-cracks grow in a crack having significant size by the fatigue or the creep. The crack penetration in the first wall can be a factor which threatens the safety of the fusion power reactor. This paper summarizes the problems on the structural integrity in the first wall made of the SiC/SiC composite material or the ferritic steel. (author)

  4. Fusion power in the E.E.C. - some considerations concerning the future programme

    International Nuclear Information System (INIS)

    Carruthers, R.

    1976-01-01

    The problems of fusion reactor technology, the assessment of potential reactor systems and an estimate of the overall investment of manpower likely to be needed to reach a practical fusion power reactor are presented. (U.K.)

  5. Fusion research principles

    CERN Document Server

    Dolan, Thomas James

    2013-01-01

    Fusion Research, Volume I: Principles provides a general description of the methods and problems of fusion research. The book contains three main parts: Principles, Experiments, and Technology. The Principles part describes the conditions necessary for a fusion reaction, as well as the fundamentals of plasma confinement, heating, and diagnostics. The Experiments part details about forty plasma confinement schemes and experiments. The last part explores various engineering problems associated with reactor design, vacuum and magnet systems, materials, plasma purity, fueling, blankets, neutronics

  6. Neutron multiplier alternative for fusion reactor blankets

    International Nuclear Information System (INIS)

    Taczanowski, S.

    1980-01-01

    A proposal is given to replace neutron multiplier needed to enable low lithium and tritium inventories simultaneously assuring sufficient production of tritium, by an efficient moderator ( 7 LiH or 7 LiD). The advantageous effect of the intensified neutron energy degradation is due to the 1/v character of the main tritium producing reaction. The slowing-down medium is designed to be the source of moderated neutrons for the surrounding Li ( 6 Li enriched) region where the most of tritium is to be produced. The surplus tritium production remains stored in the moderator zone. Some preliminary calculations illustrating the above concept were carried out and the neutron flux and tritium production distributions are presented. The indications regarding further studies are also suggested. (author)

  7. Fusion reactors and the environment

    International Nuclear Information System (INIS)

    Wrixon, A.D.

    1976-01-01

    A summary is given of the report of a study group set up in 1971 by the Director of the UKAEA Culham Laboratory to investigate environmental and safety aspects of future commercial fusion reactors (1975, Carruthers, R., Dunster, H.J., Smith, R.D., Watson, C.J.H., and Mitchell, J.T.D., Culham Study Group Report on Fusion Reactors and the Environment, CLM-R148, HMSO, London). This report was originally issued in 1973 under limited distribution, but has only recently been made available for open circulation. Deuterium/tritium fusion is thought to be the most likely reaction to be used in the first generation of reactors. Estimates were made of the local and world-wide population hazards from the release of tritium, both under normal operating conditions and in the event of an accident. One serious type of accident would be a lithium metal fire in the blanket region of the reactor. The use of a fusible lithium salt (FLIBE), eliminating the lithium fire risk, is considered but the report concentrates on lithium metal in the blanket region. The main hazards to operating staff arise both from tritium and from neutron activation of the construction materials. Remote servicing of the reactor structure will be essential, but radioactive waste management seems less onerous than for fission reactors. Meaningful comparison of the overall hazards associated with fusion and fission power programmes is not yet possible. The study group emphasized the need for more data to aid the safety assessments, and the need for such assessments to keep pace with fusion power station design. (U.K.)

  8. Blanket/first wall challenges and required R&D on the pathway to DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, Mohamed, E-mail: abdou@fusion.ucla.edu; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

    2015-11-15

    The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

  9. Blanket/first wall challenges and required R&D on the pathway to DEMO

    International Nuclear Information System (INIS)

    Abdou, Mohamed; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

    2015-01-01

    The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

  10. Fusion power

    International Nuclear Information System (INIS)

    Hancox, R.

    1981-01-01

    The principles of fusion power, and its advantages and disadvantages, are outlined. Present research programmes and future plans directed towards the development of a fusion power reactor, are summarized. (U.K.)

  11. First preliminary design of an experimental fusion reactor

    International Nuclear Information System (INIS)

    1977-09-01

    A preliminary design of a tokamak experimental fusion reactor to be built in the near future is under way. The goals of the reactor are to achieve reactor-level plasma conditions for a sufficiently long operation period and to obtain design, construction and operational experience for the main components of full-scale power reactors. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics, shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel circulating system, reactor cooling system, tritium recovery system and maintenance scheme. The main design parameters are as follows: the reactor fusion power 100 MW, torus radius 6.75 m, plasma radius 1.5 m, first wall radius 1.75 m, toroidal magnet field on axis 6 T, blanket fertile material Li 2 O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

  12. Advanced spheromak fusion reactor

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1996-01-01

    The spheromak has no toroidal magnetic field coils or other structure along its geometric axis, and is thus more attractive than the leading magnetic fusion reactor concept, the tokamak. As a consequence of this and other attributes, the spheromak reactor may be compact and produce a power density sufficiently high to warrant consideration of a liquid 'blanket' that breeds tritium, converts neutron kinetic energy to heat, and protects the reactor vessel from severe neutron damage. However, the physics is more complex, so that considerable research is required to learn how to achieve the reactor potential. Critical physics problems and possible ways of solving them are described. The opportunities and issues associated with a possible liquid wall are considered to direct future research

  13. The TFTR lithium blanket module program

    International Nuclear Information System (INIS)

    Jassby, D.L.; Bertone, P.C.; Creedon, R.L.; File, J.; Graumann, D.W.

    1985-01-01

    The Lithium Blanket Module (LBM) is an approximately 80X80X80 cm cubic module, representative of a helium-cooled lithium oxide fusion reactor blanket module, that will be installed on the TFTR (Tokamak Fusion Test Reactor) in late 1986. The principal objective of the LBM Program is to perform a series of neutron transport and tritium-breeding measurements throughout the LBM when it is exposed to the TFTR toroidal fusion neutron source, and to compare these data with the predictions of Monte Carlo (MCNP) neutronics codes. The LBM consists of 920 2.5-cm diameter breeder rods constructed of lithium oxide (Li 2 O) pellets housed in thin-walled stainless steel tubes. Procedures for mass-producing 25,000 Li 2 O pellets with satisfactory reproducibility were developed using purified Li 2 O powder, and fabrication of all the breeder rods was completed in early 1985. Tritium assay methods were investigated experimentally using both small lithium metal samples and LBM-type pellets. This work demonstrated that the thermal extraction method will be satisfactory for accurate evaluation of the minute concentrations of tritium expected in the LBM pellets (0.1-1nCi/g)

  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. The FELIX experimental program and future upgrades

    International Nuclear Information System (INIS)

    Turner, L.R.; Praeg, W.F.; Lari, R.J.; Wehrle, R.B.

    1981-01-01

    As part of the DOE First Wall/Blanket/Shield (FW/B/S) Engineering Test Program, Argonne National Laboratory FELIX (Fusion ELectromagnetic Induction EXperiment to study electromagnetic effects. T.he earliest test will select and verify appropriate eddy current simulation computer program (codes), followed by component concept tests, component model tests, and finally tests with prototypes. This paper describes the experimental and computer code plans and future upgrades for the FELIX facility

  16. Development of laser-based techniques for in situ characterization of the first wall in ITER and future fusion devices

    NARCIS (Netherlands)

    Philipps, V.; Malaquias, A.; Hakola, A.; Karhunen, J.; Maddaluno, G.; Almaviva, S.; Caneve, L.; Colao, F.; Fortuna, E.; Gasior, P.; Kubkowska, M.; Czarnecka, A.; Laan, M.; Lissovski, A.; Paris, P.; van der Meiden, H. J.; Petersson, P.; Rubel, M.; Huber, A.; Zlobinski, M.; Schweer, B.; Gierse, N.; Xiao, Q.; Sergienko, G.

    2013-01-01

    Analysis and understanding of wall erosion, material transport and fuel retention are among the most important tasks for ITER and future devices, since these questions determine largely the lifetime and availability of the fusion reactor. These data are also of extreme value to improve the

  17. 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

  18. Present status of fusion reactor materials, 4

    International Nuclear Information System (INIS)

    Nagasaki, Ryukichi; Shiraishi, Kensuke; Watanabe, Hitoshi; Murakami, Yoshio; Takamura, Saburo

    1982-01-01

    Recently, the design of fusion reactors such as Intor has been carried out, and various properties that fusion reactor materials should have been clarified. In the Japan Atomic Energy Research Institute, the research and development of materials aiming at a tokamak type experimental fusion reactor are in progress. In this paper, the problems, the present status of research and development and the future plan about the surface materials and structural materials for the first wall, blanket materials and magnet materials are explained. The construction of the critical plasma testing facility JT-60 developed by JAERI has progressed smoothly, and the operation is expected in 1985. The research changes from that of plasma physics to that of reactor technology. In tokamak type fusion reactors, high temperature D-T plasma is contained with strong magnetic field in vacuum vessels, and the neutrons produced by nuclear reaction, charged particles diffusing from plasma and neutral particles by charge exchange strike the first wall. The PCA by improving 316 stainless steel is used as the structural material, and TiC coating techniques are developed. As the blanket material, Li 2 O is studied, and superconducting magnets are developed. (Koko, I.)

  19. The ORNL fusion power demonstration study

    International Nuclear Information System (INIS)

    Shannon, T.E.; Steiner, D.

    1978-01-01

    In this paper, we review the design approach developed in the ORNL Fusion Power Demonstration Study [1]. The major emphasis of this study is in the application of current and near-term technology as the most logical path to near-term demonstration of tokamak fusion power. In addition we are pursuing a number of concepts to simplify the tokamak reactor to be more acceptable to the utility industry as a future source of energy. The discussion will focus on the areas having the greatest overall impact on reactor feasibility: 1) overall size and power output, 2) remote maintenance considerations, 3) electrical power supplies, 4) blanket design; and 5) economics. The tokamak device, by nature of its configuration and pulsed operation, is an exceptionally complex engineering design problem. We have concluded that innovative design concepts are essential to cope with this basic complexity. We feel that the feasibility of tokamak fusion power has been significantly improved by these design approaches. (author)

  20. The 22nd symposium on fusion technology

    International Nuclear Information System (INIS)

    Taehtinen, S.; Rintamaa, R.; Asikainen, M.; Tuomisto, H.

    2002-01-01

    The Symposium on Fusion Technology (SOFT) was held at the Marina Congress Center, Helsinki, Finland, from 9th to 13th September 2002. It was organized by the Association Euratom-Tekes and hosted by the VTT Technical Research Centre of Finland, Fortum Nuclear Services Ltd. and PrizzTech Oy. The sympoisum included invited and contributed papers as well as poster presentations and an industrial and R and D exhibition. The main topics included all aspects of fusion technology: current and future devices, plasma facing components, plasma heating and current drive, plasma engineering and control, diagnostics, data acquisition and remote participation, magnets and power supplies, fuel cycle, remote handling, vessel, blanket and shield, safety and environment, power plant and socio-economic studies, inertial fusion energy, and transfer of technology. The number of invited speakers was 15, selected presentations 22 and poster presentations 404. The abstracts of the presentations and posters are included in this book. (orig.)

  1. Fusion Power: A Strategic Choice for the Future Energy Provision. Why is So Much Time Wasted for Decision Making?

    International Nuclear Information System (INIS)

    D'haeseleer, William D.

    2005-01-01

    From a general analysis of the world energy issue, it is argued that an affordable, clean and reliable energy supply will have to consist of a portfolio of primary energy sources, a large fraction of which will be converted to a secondary carrier in large baseload plants. Because of all future uncertainties, it would be irresponsible not to include thermonuclear fusion as one of the future possibilities for electricity generation.The author tries to understand why nuclear-fusion research is not considered of strategic importance by the major world powers. The fusion programs of the USA and Europe are taken as prime examples to illustrate the 'hesitation'. Europe is now advocating a socalled 'fast-track' approach, thereby seemingly abandoning the 'classic' time frame towards fusion that it has projected for many years. The US 'oscillatory' attitude towards ITER in relation to its domestic program is a second case study that is looked at.From the real history of the ITER design and the 'siting' issue, one can try to understand how important fusion is considered by these world powers. Not words are important, but deeds. Fast tracks are nice to talk about, but timely decisions need to be taken and sufficient money is to be provided. More fundamental understanding of fusion plasma physics is important, but in the end, real hardware devices must be constructed to move along the path of power plant implementation.The author tries to make a balance of where fusion power research is at this moment, and where, according to his views, it should be going

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

    International Nuclear Information System (INIS)

    Tanaka, Satoru

    1987-01-01

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

  3. Preliminary electromagnetic analysis of Helium Cooled Solid Blanket for CFETR by MAXWELL

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Cheng; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn

    2016-11-15

    Highlights: • A FEM model of the blanket and magnetic system was built. • Electromagnetic forces and moments of the typical blanket for ferromagnetic and non-ferromagnetic materials were computed and analyzed. • Maxwell forces and Lorentz forces were computed and compared. • Eddy current in the blanket was analyzed under MD condition. - Abstract: A Helium Cooled Solid Blanket (HCSB) for CFETR (Chinese Fusion Engineering Test Reactor) was designed by USTC. The structural and thermal-hydraulic analysis has been carried out, while electromagnetic analysis was not carefully researched. In this paper, a FEM (finite element method) model of the HCSB was developed and electromagnetic forces as well as moments was computed by a FEM software called MAXWELL integrated in ANSYS Workbench. In the geometrical model, flow channels and small connecting parts were neglected because of the extreme complication and the reasonable conservative assumption by neglecting these circumstantial details. As for electromagnetic (EM) analysis, Lorentz forces due to eddy currents caused by main disruption and Maxwell forces due to the magnetization of RAFM steel (i.e. EUROFER97) were computed. Since the unavailability of the details of the plasma in CFETR, when disruptions happen, the condition where a linear current quench of main disruption occurs was assumed. The maximum magnitude of the electromagnetic forces was 356.45 kN and the maximum value of the coupled electromagnetic moments was 1899.40 N m around the radial direction. It is feasible to couple electromagnetic analysis, structural analysis and thermal-hydraulic analysis in the future since MAXWELL has good channels to exchange data between different analytic parts.

  4. Liquid lithium blanket processing studies

    International Nuclear Information System (INIS)

    Talbot, J.B.; Clinton, S.D.

    1979-01-01

    The sorption of tritium on yttrium from flowing molten lithium and the subsequent release of tritium from yttrium for regeneration of the metal sorbent were investigated to evaluate the feasibility of such a tritium-recovery process for a fusion reactor blanket of liquid lithium. In initial experiments with the forced convection loop, yttrium samples were contacted with lithium at 300 0 C. A mass transfer coefficient of 2.5 x 10 - cm/sec, which is more than an order of magnitude less than the value measured in earlier static experiments, was determined for the flowing lithium system. Rates of tritium release from yttrium samples were measured to evaluate possible thermal regeneration of the sorbent. Values for diffusion coefficients at 505, 800, and 900 0 C were estimated to be 1.1 x 10 -13 , 4.9 x 10 -12 , and 9.3 x 10 -10 cm 2 /sec, respectively. Tritium release from yttrium was investigated at higher temperatures and with hydrogen added to the argon sweep gas to provide a reducing atmosphere

  5. Particle and energy transport studies on TFTR and implications for helium ash in future fusion devices

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Efthimion, P.C.; Rewoldt, G.; Stratton, B.C.; Tang, W.M.; Bell, R.E.; Grek, B.; Hulse, R.A.; Johnson, D.W.; Hill, K.W.; Mansfield, D.K.; McCune, D.; Mikkelsen, D.R.; Park, H.K.; Ramsey, A.T.; Scott, S.D.; Taylor, G.; Timberlake, J.; Zarnstorff, M.C.

    1992-01-01

    Particle and energy transport in tokamak plasmas have long been subjects of vigorous investigation. Present-day measurement techniques permit radially resolved studies of the transport of electron perturbations, low- and high-Z impurities, and energy. In addition, developments in transport theory provide tools that can be brought to bear on transport issues. Here, we examine local particle transport measurements of electrons, fully-stripped thermal helium, and helium-like iron in balanced-injection L-mode and enhanced confinement deuterium plasmas on TFTR of the same plasma current, toroidal field, and auxiliary heating power. He 2+ and Fe 24+ transport has been studied with charge exchange recombination spectroscopy, while electron transport has been studied by analyzing the perturbed electron flux following the same helium puff used for the He 2+ studies. By examining the electron and He 2+ responses following the same gas puff in the same plasmas, an unambiguous comparison of the transport of the two species has been made. The local energy transport has been examined with power balance analysis, allowing for comparisons to the local thermal fluxes. Some particle and energy transport results from the Supershot have been compared to a transport model based on a quasilinear picture of electrostatic toroidal drift-type microinstabilities. Finally, implications for future fusion reactors of the observed correlation between thermal transport and helium particle transport is discussed

  6. Development and evaluation of plasma facing materials for future thermonuclear fusion reactors

    International Nuclear Information System (INIS)

    Linke, J.; Pintsuk, G.; Roedig, M.; Schmidt, A.; Thomser, C.

    2010-01-01

    More and more attention is directed towards thermonuclear fusion as a possible future energy source. Major advantages of this energy conversion technology are the almost inexhaustible resources and the option to produce energy without CO 2 -emissions. However, in the most advanced field of magnetic plasma confinement a number of technological challenges have to be met. In particular high-temperature resistant and plasma compatible meterials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PEMs) and components (PFCs) in such fusion devices, i.e. the first wall (FW), the limiters and the divertor, are strongly affected by the plasma wall interaction processes and the applied intense thermal loads during plasma operation. On the one hand, these mechanisms have a strong influence on the plasma performance; on the other hand, they have major impact on the lifetime of the plasma facing armour. Materials for plasma facing components have to fulfill a number of requirements. First of all the materials have to be plasma compatible, i.e. they should exhibit a low atomic number to avoid radiative losses whenever atoms from the wall material will be ionized in the plasma. In addition, the materials must have a high melting point, a high thermal conductivity, and adequate mechanical properties. To select the most suitable material candidates, a comprehensive data base is required which includes all thermo-physical and mechanical properties. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm -2 , meanwhile the limiters and the divertor are expected to be exposed to power densities being at least one order of magnitude above the FW-level, i.e. up to 20 MWm -2 for next step tokamaks such as ITER or DEMO. These requirements are responsible for high demands on the selection of qualified PFMs and heat

  7. Development and evaluation of plasma facing materials for future thermonuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J.; Pintsuk, G.; Roedig, M.; Schmidt, A.; Thomser, C. [Forschungszentrum Juelich GmbH, EURATOM Association, Juelich (Germany)

    2010-07-01

    More and more attention is directed towards thermonuclear fusion as a possible future energy source. Major advantages of this energy conversion technology are the almost inexhaustible resources and the option to produce energy without CO{sub 2}-emissions. However, in the most advanced field of magnetic plasma confinement a number of technological challenges have to be met. In particular high-temperature resistant and plasma compatible meterials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PEMs) and components (PFCs) in such fusion devices, i.e. the first wall (FW), the limiters and the divertor, are strongly affected by the plasma wall interaction processes and the applied intense thermal loads during plasma operation. On the one hand, these mechanisms have a strong influence on the plasma performance; on the other hand, they have major impact on the lifetime of the plasma facing armour. Materials for plasma facing components have to fulfill a number of requirements. First of all the materials have to be plasma compatible, i.e. they should exhibit a low atomic number to avoid radiative losses whenever atoms from the wall material will be ionized in the plasma. In addition, the materials must have a high melting point, a high thermal conductivity, and adequate mechanical properties. To select the most suitable material candidates, a comprehensive data base is required which includes all thermo-physical and mechanical properties. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm{sup -2}, meanwhile the limiters and the divertor are expected to be exposed to power densities being at least one order of magnitude above the FW-level, i.e. up to 20 MWm{sup -2} for next step tokamaks such as ITER or DEMO. These requirements are responsible for high demands on the selection of qualified PFMs

  8. 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)

  9. 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

  10. 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

  11. Optimization design study of an innovative divertor concept for future experimental tokamak-type fusion reactors

    International Nuclear Information System (INIS)

    Willem Janssens, Ir.; Crutzen, Y.; Farfaletti-Casali, F.; Matera, R.

    1991-01-01

    The design optimization study of an innovative divertor concept for future experimental tokamak-type fusion devices is both an answer to the actual problems encountered in the multilayer divertor proposals and an illustration of a rational modelling philosophy and optimization strategy for the development of a new divertor structure. Instead of using mechanical attachment or metallurgical bonding of the protective material to the heat sink as in most actual divertor concepts, the so-called brush divertor in this study uses an array of unidirectional fibers penetrating in both the protective armor and the underling composite heat sink. Although the approach is fully concentrated on the divertor performance, including both a description of its function from the theoretical point of view and an overview of the problems related to the materials choice and evaluation, both the approach followed in the numerical modelling and the judgment of the results are thought to be valid also for other applications. Therefore the spin-off of the study must be situated in both the technological progress towards a feasible divertor solution, which introduces no additional physical uncertainties, and in the general area of the thermo-mechanical finite-element modelling on both macro-and microscale. The brush divertor itself embodies the use, and thus the modelling, of advanced materials such as tailor-made metal matrix composites and dispersion strengthened metals, and is shown to offer large potential advantages, demanding however and experimental validation under working conditions. It is clearly indicated where the need originates for an integrated experimental program which must allow to verify the basic modelling assumptions in order to arrive at the use of numerical computation as a powerful and realistic tool of structural testing and life-time prediction

  12. 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.

  13. Results from levels 2/3 fusion implementations: issues, challenges, retrospectives, and perspectives for the future an annotated perspective

    Science.gov (United States)

    Kadar, Ivan; Bosse, Eloi; Salerno, John; Lambert, Dale A.; Das, Subrata; Ruspini, Enrique H.; Rhodes, Bradley J.; Biermann, Joachim

    2008-04-01

    Even though the definition of the Joint Director of Laboratories (JDL) "fusion levels" were established in 1987, published 1991, revised in 1999 and 2004, the meaning, effects, control and optimization of interactions among the fusion levels have not as yet been fully explored and understood. Specifically, this is apparent from the abstract JDL definitions of "Levels 2/3 Fusion" - situation and threat assessment (SA/TA), which involve deriving relations among entities, e.g., the aggregation of object states (i.e., classification and location) in SA, while TA uses SA products to estimate/predict the impact of actions/interactions effects on situations taken by the participant entities involved. Given all the existing knowledge in the information fusion and human factors literature, (both prior to and after the introduction of "fusion levels" in 1987) there are still open questions remaining in regard to implementation of knowledge representation and reasoning methods under uncertainty to afford SA/TA. Therefore, to promote exchange of ideas and to illuminate the historical, current and future issues associated with Levels 2/3 implementations, leading experts were invited to present their respective views on various facets of this complex problem. This paper is a retrospective annotated view of the invited panel discussion organized by Ivan Kadar (first author), supported by John Salerno, in order to provide both a historical perspective of the evolution of the state-of-the-art (SOA) in higher-level "Levels 2/3" information fusion implementations by looking back over the past ten or more years (before JDL), and based upon the lessons learned to forecast where focus should be placed to further enhance and advance the SOA by addressing key issues and challenges. In order to convey the panel discussion to audiences not present at the panel, annotated position papers summarizing the panel presentation are included.

  14. Magnetic fusion energy

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    The efforts of the Chemical Technology Division in the area of fusion energy include fuel handling, processing, and containment. These studies are closely coordinated with the ORNL Fusion Energy Division. Current experimental studies are concerned with the development of vacuum pumps for fusion reactors, the evaluation and development of techniques for recovering tritium (fuel) from either solid or liquid lithium containing blankets, and the use of deep beds of sorbents as roughing pumps and/or transfer operations. In addition, a small effort is devoted to the support of the ORNL design of The Next Step (TNS) in tokamak reactor development. The more applied studies--vacuum pump development and TNS design--are funded by the DOE/Magnetic Fusion Energy, and the more fundamental studies--blanket recovery and sorption in deep beds--are funded by the DOE/Basic Energy Sciences

  15. 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)

  16. Panel discussion: Future directions in magnetic fusion--comments of John Sheffield, Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Sheffield, J.

    1992-01-01

    I will discuss two important issues for the US magnetic fusion program: the role of alternate magnetic configurations to the tokamak, and factors which need to be considered in planning the evolution of the US program

  17. Low cost, high yield IFE reactors: Revisiting Velikhov's vaporizing blankets

    International Nuclear Information System (INIS)

    Logan, B.G.

    1992-01-01

    The performance (efficiency and cost) of IFE reactors using MHD conversion is explored for target blanket shells of various materials vaporized and ionized by high fusion yields (5 to 500 GJ). A magnetized, prestressed reactor chamber concept is modeled together with previously developed models for the Compact Fusion Advanced Rankine II (CFARII) MHD Balance-of-Plant (BoP). Using conservative 1-D neutronics models, high fusion yields (20 to 80 GJ) are found necessary to heat Flibe, lithium, and lead-lithium blankets to MHD plasma temperatures, at initial solid thicknesses sufficient to capture most of the fusion yield. Advanced drivers/targets would need to be developed to achieve a ''Bang per Buck'' figure-of-merit approx-gt 20 to 40 joules yield per driver $ for this scheme to be competitive with these blanket materials. Alternatively, more realistic neutronics models and better materials such as lithium hydride may lower the minimum required yields substantially. The very low CFARII BoP costs (contributing only 3 mills/kWehr to CoE) allows this type of reactor, given sufficient advances that non-driver costs dominate, to ultimately produce electricity at a much lower cost than any current nuclear plant

  18. Thermochemical hydrogen production based on magnetic fusion

    International Nuclear Information System (INIS)

    Krikorian, O.H.; Brown, L.C.

    1982-01-01

    Conceptual design studies have been carried out on an integrated fusion/chemical plant system using a Tandem Mirror Reactor fusion energy source to drive the General Atomic Sulfur-Iodine Water-Splitting Cycle and produce hydrogen as a future feedstock for synthetic fuels. Blanket design studies for the Tandem Mirror Reactor show that several design alternatives are available for providing heat at sufficiently high temperatures to drive the General Atomic Cycle. The concept of a Joule-boosted decomposer is introduced in one of the systems investigated to provide heat electrically for the highest temperature step in the cycle (the SO 3 decomposition step), and thus lower blanket design requirements and costs. Flowsheeting and conceptual process designs have been developed for a complete fusion-driven hydrogen plant, and the information has been used to develop a plot plan for the plant and to estimate hydrogen production costs. Both public and private utility financing approaches have been used to obtain hydrogen production costs of $12-14/GJ based on July 1980 dollars

  19. Fusion for Energy: The European joint undertaking for ITER and the development of fusion energy

    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

  20. 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.)