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Sample records for lithium test blanket

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

    International Nuclear Information System (INIS)

    Wu, Y.

    2007-01-01

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

  2. Safety Analysis of the US Dual Coolant Liquid Lead-Lithium ITER Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2006-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium (DCLL) DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER Test Blanket Module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER International Team (IT) to address specific reactor safety concerns, such as VV pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system, and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  3. A low-risk aqueous lithium salt blanket for engineering test reactors

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1986-09-01

    A simple blanket concept is proposed based on 1-3 wt.% lithium dissolved as a salt in low temperature (80 degrees C) and low pressure (0.1 MPa) water. This concept can provide, for example, a 0.5 tritium breeding ratio with 60% steel structure and 70% coverage. The use of neutron multipliers, other structural materials (especially zirconium alloys), higher coverage and higher lithium salt concentrations allows tritium breeding ratios over unity if necessary. Other advantages of this concept include the simple shield-like geometry, substantial structural volume for mechanical strength, excellent heat transfer ability of water coolant, efficient neutron and gamma shielding through the combination of high-Z structure and low-Z water, and conventional tritium recovery and control technology. This concept could initially provide the shielding needs for an engineering test reactor and later, by the addition of lithium salt and tritium recovery systems, also provide tritium breeding. This staged operation and liquid breeder/coolant allows control over the tritium inventory in the device without machine disassembly. 14 refs

  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. 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)

  6. 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)

  7. 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

  8. Initial three-dimensional neutronics calculations for the EU water cooled lithium-lead test blanket module for ITER-FEAT

    International Nuclear Information System (INIS)

    Jordanova, J.; Poitevin, Y.; Li Puma, A.; Kirov, N.

    2003-01-01

    The paper summarizes the main results of the initial three-dimensional radiation transport analysis of the EU water-cooled lithium-lead test blanket module performed using the Monte Carlo code MCNP. Estimates of tritium production rate, nuclear energy deposition and cumulative fluence effects such as radiation damage through atomic displacement and production of He and H are presented. (author)

  9. 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

  10. 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

  11. 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

  12. Thermomechanical analysis of the DFLL test blanket module for ITER

    International Nuclear Information System (INIS)

    Chen Hongli; Wu Yican; Bai Yunqing

    2006-01-01

    The finite element code is used to simulate two kinds of blanket design structure, which are SLL (Quasi-Static Lithium Lead) and DLL (Dual-cooled Lithium Lead) blanket concepts for the Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM) submitted to the ITER test blanket working group. The temperature and stress distributions have been presented for the two kinds of blanket structure on the basis of the structural design, thermal-hydraulic design and neutronics analysis. Also the mechanical performance is presented for the high temperature component of blanket structure according to the ITER Structural Design Criteria (ISDC). The rationality and feasibility of the two kinds of blanket structure design of DFLL-TBM have been analyzed based on the above results which also acted as the theoretical base for further optimized analysis. (authors)

  13. 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.).

  14. The requirements for processing tritium recovered from liquid lithium blankets: The blanket interface

    International Nuclear Information System (INIS)

    Clemmer, R.G.; Finn, P.A.; Greenwood, L.R.; Grimm, T.L.; Sze, D.K.; Bartlit, J.R.; Anderson, J.L.; Yoshida, H.; Naruse.

    1988-03-01

    We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. 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. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  16. 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

  17. 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

  18. 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.

  19. NET test blanket design and remote maintenance

    International Nuclear Information System (INIS)

    Holloway, C.; Hubert, P.

    1991-01-01

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

  20. 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.

  1. 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

  2. Status on DEMO Helium Cooled Lithium Lead breeding blanket thermo-mechanical analyses

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, G.; Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Kiss, B. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Morin, A. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France)

    2016-11-01

    Highlights: • CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. The DEMO HCLL breeding blanket design capitalizes on the experience acquired on the HCLL Test Blanket Module designed for ITER. Design improvements are being implemented to adapt the design to DEMO specifications and performance objectives. • Thermal and mechanical analyses have been carried out in order to justify the design of the HCLL breeding blanket showing promising results for tie rods modules’ attachments system and relatively good behavior of the box in case of LOCA when comparing to RCC-MRx criteria. • CFD thermal analyses on generic breeding unit have enabled the consolidation of the results obtained with previous FEM design analyses. - Abstract: The EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the breeding blanket surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which is investigated for DEMO. It is made of a Eurofer structure and uses the eutectic liquid lithium–lead as tritium breeder and neutron multiplier, and helium gas as coolant. Within the EUROfusion organization, CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. This paper presents the status of the thermal and mechanical analyses carried out on the HCLL breeding blanket in order to justify the design. CFD thermal analyses on generic breeding unit including stiffening plates and cooling plates have been performed with ANSYS in order to consolidate results obtained with previous FEM design analyses. Moreover in order to expand the justification of the HCLL Breeding blanket design, the most loaded area of

  3. Lithium-vanadium advanced blanket development. ITER final report on U.S. contribution: Task T219/T220

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Mattas, R.F. [comps.

    1997-07-01

    The objective of this task is to develop the required data base and demonstrate the performance of a liquid lithium-vanadium advanced blanket design. The task has two main activities related to vanadium structural material and liquid lithium system developments. The vanadium alloy development activity included four subtasks: (1.1) baseline mechanical properties of non irradiated base metal and weld metal joints; (1.2) compatibility with liquid lithium; (1.3) material irradiation tests; and (1.4) development of material manufacturing and joining methods. The lithium blanket technology activity included four subtasks: (2.1) electrical insulation development and testing for liquid metal systems; (2.2) MHD pressure drop and heat transfer study for self-cooled liquid metal systems; (2.3) chemistry of liquid lithium; and (2.4) design, fabrication and testing of ITER relevant size blanket mockups. A summary of the progress and results obtained during the period 1995 and 1996 in each of the subtask areas is presented in this report.

  4. Lithium-vanadium advanced blanket development. ITER final report on U.S. contribution: Task T219/T220

    International Nuclear Information System (INIS)

    Smith, D.L.; Mattas, R.F.

    1997-07-01

    The objective of this task is to develop the required data base and demonstrate the performance of a liquid lithium-vanadium advanced blanket design. The task has two main activities related to vanadium structural material and liquid lithium system developments. The vanadium alloy development activity included four subtasks: (1.1) baseline mechanical properties of non irradiated base metal and weld metal joints; (1.2) compatibility with liquid lithium; (1.3) material irradiation tests; and (1.4) development of material manufacturing and joining methods. The lithium blanket technology activity included four subtasks: (2.1) electrical insulation development and testing for liquid metal systems; (2.2) MHD pressure drop and heat transfer study for self-cooled liquid metal systems; (2.3) chemistry of liquid lithium; and (2.4) design, fabrication and testing of ITER relevant size blanket mockups. A summary of the progress and results obtained during the period 1995 and 1996 in each of the subtask areas is presented in this report

  5. 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

  6. 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

  7. 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

  8. Tritium transport modeling at system level for the EUROfusion dual coolant lithium-lead breeding blanket

    Science.gov (United States)

    Urgorri, F. R.; Moreno, C.; Carella, E.; Rapisarda, D.; Fernández-Berceruelo, I.; Palermo, I.; Ibarra, A.

    2017-11-01

    The dual coolant lithium lead (DCLL) breeding blanket is one of the four breeder blanket concepts under consideration within the framework of EUROfusion consortium activities. The aim of this work is to develop a model that can dynamically track tritium concentrations and fluxes along each part of the DCLL blanket and the ancillary systems associated to it at any time. Because of tritium nature, the phenomena of diffusion, dissociation, recombination and solubilisation have been modeled in order to describe the interaction between the lead-lithium channels, the structural material, the flow channel inserts and the helium channels that are present in the breeding blanket. Results have been obtained for a pulsed generation scenario for DEMO. The tritium inventory in different parts of the blanket, the permeation rates from the breeder to the secondary coolant and the amount of tritium extracted from the lead-lithium loop have been computed. Results present an oscillating behavior around mean values. The obtained average permeation rate from the liquid metal to the helium is 1.66 mg h-1 while the mean tritium inventory in the whole system is 417 mg. Besides the reference case results, parametric studies of the lead-lithium mass flow rate, the tritium extraction efficiency and the tritium solubility in lead-lithium have been performed showing the reaction of the system to the variation of these parameters.

  9. Potential and problems of an aqueous lithium salt solution blanket for NET

    International Nuclear Information System (INIS)

    Kuechle, M.; Bojarsky, E.; Dorner, S.; Fischer, U.; Reimann, J.; Reiser, H.

    1987-07-01

    The report describes design studies on a water cooled in-vessel shield blanket for NET and its modification into an aqueous lithium salt blanket. The shield blankets are exchangable against breeding blankets and fulfill their shielding and heat removal functions. Emphasis is on simplicity and reliability. The water cooled shield is a large steel container in the shape of the blanket segment which is filled by water and containes a grid structure of poloidally arranged steel plates. The water flows several times in poloidal direction through the channels formed by the steel plates and is thereby heated up from 40degC to 70degC. When the water is replaced by an aqueous lithium salt solution the shield can be converted into a tritium breeding blanket without any design modification or invessel component replacement. When compared with other concepts this blanket has the advantage that the solution can replace water cooling also in the divertor and in segments dedicated to plasma heating and diagnostics, what increases the coverage considerably. Extensive three-dimensional neutronics calculations were done which, together with literature studies on candidate materials, corrosion, and tritium recovery led to a first assessment of the concept. There is an indication that no major corrosion problems are to be expected in the low temperature region envisaged. Tritium recovery capital costs were estimated to be in the 20 MECU to 50 MECU range and tritium breeding ratio is comparable to the best breeding blanket. (orig./GG) [de

  10. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Sato, Satoshi; Hatano, Toshihisa; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hara, Shigemitsu

    1999-03-01

    A design and a fabrication of heating test facility for a mock-up of the blanket module to be installed in International Thermonuclear Experimental Reactor (ITER) have been conducted to evaluate/demonstrate its heat removal performance and structural soundness under cyclic heat loads. To simulate surface heat flux to the blanket module, infrared heating method is adopted so as to heat large surface area uniformly. The infrared heater is used in vacuum environment (10{sup -4} Torr{approx}), and the lamps are cooled by air flowing through an annulus between the lamp and a cover tube made of quartz glass. Elastomer O rings (available to be used up to {approx}300degC) and used for vacuum seal at outer surface of the cover tube. To prevent excessive heating of the O ring, the end part of the cover tube is specially designed including the tube shape, flow path of air and gold coating on the surface of the cover tube to protect the O ring against thermal radiation from glowing tungsten filament. To examine the performance of the facility, steady state and cyclic operation of the infrared heater were conducted using a small-scaled shielding blanket mock-up as a test specimen. The important results are as follows: (1) Heat flux at the surface of the small-scaled mock-up measured by a calorimeter was {approx}0.2 MW/m{sup 2}. (2) A comparison of thermal analysis results and measured temperature responses showed that the small-scaled mock-up had good heat removal performance. (3) Steady state operation and cyclic operation with step response between the rated and zero powers of the infrared heater were successfully performed, and it was confirmed that this heating facility was well-prepared and available for the thermal cyclic test of a blanket module. (author)

  11. Liquid metal magnetohydrodynamic flows in manifolds of dual coolant lead lithium blankets

    Energy Technology Data Exchange (ETDEWEB)

    Mistrangelo, C., E-mail: chiara.mistrangelo@kit.edu; Bühler, L.

    2014-10-15

    Highlights: • MHD flows in model geometries of DCLL blanket manifolds. • Study of velocity, pressure distributions and flow partitioning in parallel ducts. • Flow partitioning affected by 3D MHD pressure drop and velocity distribution in the expanding zone. • Reduced pressure drop in a continuous expansion compared to a sudden expansion. - Abstract: An attractive blanket concept for a fusion reactor is the dual coolant lead lithium (DCLL) blanket where reduced activation steel is used as structural material and a lead lithium alloy serves both to produce tritium and to remove the heat in the breeder zone. Helium is employed to cool the first wall and the blanket structure. Some critical issues for the feasibility of this blanket concept are related to complex induced electric currents and 3D magnetohydrodynamic (MHD) phenomena that occur in distributing and collecting liquid metal manifolds. They can result in large pressure drop and undesirable flow imbalance in parallel poloidal ducts forming blanket modules. In the present paper liquid metal MHD flows are studied for different design options of a DCLL blanket manifold with the aim of identifying possible sources of flow imbalance and to predict velocity and pressure distributions.

  12. 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.)

  13. 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

  14. 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

  15. 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

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

    International Nuclear Information System (INIS)

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

    2006-07-01

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

  17. Neutronics and activation analysis of lithium-based ternary alloys in IFE blankets

    Energy Technology Data Exchange (ETDEWEB)

    Jolodosky, Alejandra, E-mail: aleja311@berkeley.edu [University of California Berkeley, Berkeley, CA 94706 (United States); Kramer, Kevin [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA (United States); Meier, Wayne; DeMuth, James; Reyes, Susana [TerraPower, Bellevue, WA 98005 (United States); Fratoni, Massimiliano [University of California Berkeley, Berkeley, CA 94706 (United States)

    2016-06-15

    Highlights: • Monte Carlo calculations were performed on numerous lithium ternary alloys. • Elements with high neutron multiplication performed well with low absorbers. • Enriching lithium decreases minimum lithium concentration of alloys by 60% or more. • Alloys that performed well neutronically were selected for activation calculations. • Alloys activated, except LiBaBi, do not pose major environmental or safety concerns. - Abstract: An attractive feature of using liquid lithium as the breeder and coolant in fusion blankets is that it has very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and presents plant safety concerns. The Lawrence Livermore National Laboratory is carrying an effort to develop a lithium-based ternary alloy that maintains the beneficial properties of lithium (e.g. high tritium breeding and solubility) and at the same time reduces overall flammability concerns. This study evaluates the neutronics performance of lithium-based alloys in the blanket of an inertial fusion energy chamber in order to inform such development. 3-D Monte Carlo calculations were performed to evaluate two main neutronics performance parameters for the blanket: tritium breeding ratio (TBR), and the fusion energy multiplication factor (EMF). It was found that elements that exhibit low absorption cross sections and higher q-values such as Pb, Sn, and Sr, perform well with those that have high neutron multiplication such as Pb and Bi. These elements meet TBR constrains ranging from 1.02 to 1.1. However, most alloys do not reach EMFs greater than 1.15. Additionally, it was found that enriching lithium with {sup 6}Li significantly increases the TBR and decreases the minimum lithium concentration by more than 60%. The amount of enrichment depends on how much total lithium is in the alloy to begin with. Alloys that performed well in the TBR

  18. Enhanced fuel production in thorium/lithium hybrid blankets utilizing uranium multipliers

    Energy Technology Data Exchange (ETDEWEB)

    Pitulski, R.H.

    1979-10-01

    A consistent neutronics analysis is performed to determine the effectiveness of uranium bearing neutron multiplier zones on increasing the production of U/sup 233/ in thorium/lithium blankets for use in a tokamak fusion-fission hybrid reactor. The nuclear performance of these blankets is evaluated as a function of zone thicknesses and exposure by using the coupled transport burnup code ANISN-CINDER-HIC. Various parameters such as U/sup 233/, Pu/sup 239/, and H/sup 3/ production rates, the blanket energy multiplication, isotopic composition of the fuels, and neutron leakages into the various zones are evaluated during a 5 year (6 MW.y.m/sup -2/) exposure period. Although the results of this study were obtained for a tokomak magnetic fusion device, the qualitative behavior associated with the use of the uranium bearing neutron multiplier should be applicable to all fusion-fission hybrids.

  19. Enhanced fuel production in thorium/lithium hybrid blankets utilizing uranium multipliers

    International Nuclear Information System (INIS)

    Pitulski, R.H.

    1979-10-01

    A consistent neutronics analysis is performed to determine the effectiveness of uranium bearing neutron multiplier zones on increasing the production of U 233 in thorium/lithium blankets for use in a tokamak fusion-fission hybrid reactor. The nuclear performance of these blankets is evaluated as a function of zone thicknesses and exposure by using the coupled transport burnup code ANISN-CINDER-HIC. Various parameters such as U 233 , Pu 239 , and H 3 production rates, the blanket energy multiplication, isotopic composition of the fuels, and neutron leakages into the various zones are evaluated during a 5 year (6 MW.y.m -2 ) exposure period. Although the results of this study were obtained for a tokomak magnetic fusion device, the qualitative behavior associated with the use of the uranium bearing neutron multiplier should be applicable to all fusion-fission hybrids

  20. Tritium isolation from lithium inorganic compounds applicable to thermonuclear reactor breeding blanket

    International Nuclear Information System (INIS)

    Vasil'ev, V.G.; Ershova, Z.V.; Nikiforov, A.S.

    1982-01-01

    Tritium separation from inorganic lithium compounds: Li 2 O, LiAlO 2 , Li 2 SiO 3 , Li 4 SiO 4 , LiF, LiBeF 3 , Li 2 BeF 4 irradiated with a beam of a gamma facility and a nuclear reactor, has been studied. In the first case the gas phase is absent. In the latter one- the tritium amount in the gas does not exceed 1-2% of its total amount in the salt. Based on the EPR spectra of irradiated salts the concentrations of paramagnetic centres are calculated. It is shown that during thermal annealing the main portion of tritium in the gas phase is in the form of oxide (HTO, T 2 O). Tritium is separated from lithium fluoroberyllates in the form of hydrogen (HT, T 2 ). The kinetics of tritium oxide isolation from irradiated lithium oxide aluminate, metha- and orthosilicates, lithium sulphate has been studied. The activation energies of tritium oxide separation process are presented. A supposition is made that chemical reaction of the HTO (T 2 O) or HT(T 2 ) or HF(TF) formation is a limiting stage. Clarification of the process stage limiting the rate of tritium recovery will permit to evaluate conditions for the optimum work of lithium material in the blanket, lithium zone to select the lithium element structure and temperature regime of irradiation

  1. 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

  2. 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)

  3. 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.

  4. 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.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  6. Current Status on the Korean Test Blanket Module Development for testing in the ITER

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Bae, Young Dug; Yoon, Jae Sung; Jung, Ki Sok

    2010-01-01

    Korea has proposed and designed a Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) to be tested in the International Thermonuclear Experimental Reactor (ITER). Ferrite Martensitic (FM) steel is used as the structural material and helium (He) is used as a coolant to cool the first wall (FW) and breeding zone. Liquid lithium (Li) is circulated for a tritium breeding, not for a cooling purpose. Main purpose for developing the TBM is to develop the design technology for DEMO and fusion reactor and it should be proved through the experiment in the ITER with TBM. Therefore, we have developed the design scheme and related codes including the safety analysis for obtain the license to be tested in the ITER. In order to develop and install at the ITER, several technologies were developed in parallel; fabrication, breeder, He cooling, tritium extraction and so on. Figure 1 shows the overall TBM development scheme. In Korea, official strategy for developing the TBM is to participate to other parties' concept such as US and EU ones, in which PbLi (lead lithium eutectic), He, and FM steel were used for liquid breeder, coolant, and structural material, respectively

  7. Lithium Blanket Module dosimetry measurements at the LOTUS 14-MeV neutron source facility

    International Nuclear Information System (INIS)

    Tsang, F.Y.; Leo, W.R.; Sahraoui, C.; Wuthrich, S.; Harker, Y.D.

    1986-01-01

    This paper describes the measurements and results of the dosimeter material reaction rates inside the Lithium Blanket Module (LBM) after irradiation by the LOTUS 14-MeV neutron source at the Ecole Polytechnique Federale de Lausanne. The measurement program has been designed to utilize sets of passive dosimeter materials in the form of foils and wires. The dosimetry materials reaction thresholds and interaction response ranges chosen for this series of measurements encompass the entire neutron spectra along the full length of the LBM fuel rods

  8. 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

  9. 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

  10. 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

  11. Irradiation of lithium-based ceramics for fusion blanket application

    International Nuclear Information System (INIS)

    Hastings, I.J.; Miller, J.M.; Verrall, R.A.; Bokwa, S.R.; Rose, D.H.

    1986-06-01

    Unvented CREATE (Chalk River Experiment to Assess Tritium Emission) tests have shown that under reducing conditions, most of the tritium (greater than 70%) is released from LiAlO 2 and Li 2 O as HT or T 2 ; the balance as HTO or T 2 O. Residual tritium is very small, less than 0.02%. Varying the sweep gas composition has a dramatic effect on the form of tritium released. With a quartz extraction tube during post- irradiation heating, a He sweep gas results in 10-30% release as HT or T 2 ; with a He-1%H 2 sweep gas, greater than 60% release as HT is achieved. The effect of extraction tube material is also significant. Using pure He sweep gas, a quartz extraction tube results in 10-30% release as HT or T 2 ; stainless steel produces 80-95% as HT or T 2 . Chalk River and CEA (Saclay) - fabricated LiAlO 2 behaved similarly to that from ANL in these tests. The first vented test at Chalk River, CRITIC-I, planned for 1986/87, will examine ANL-fabricated Li 2 O, 0.3 wt% 6 Li, 30 mm ID, 40 mm OD annular pellets, in a six-month irradiation at 700-1200 K, varying the sweep gas, with on-line HT/HTO measurement

  12. The conversion of a room temperature NaK loop to a high temperature MHD facility for Li/V blanket testing

    International Nuclear Information System (INIS)

    Reed, C.B.; Haglund, R.C.; Miller, M.E.; Nasiatka, J.R.; Kirillov, I.R.; Ogorodnikov, A.P.; Preslitski, G.V.; Goloubovitch, G.P.; Xu, Zeng Yu

    1996-01-01

    The Vanadium/Lithium system has been the recent focus of ANL's Blanket Technology Pro-ram, and for the last several years, ANL's Liquid Metal Blanket activities have been carried out in direct support of the ITER (International Thermonuclear Experimental Reactor) breeding blanket task area. A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the Near the development of insulator coatings. Design calculations, Hua and Gohar, show that an electrically insulating layer is necessary to maintain an acceptably low magneto-hydrodynamic (MHD) pressure drop in the current ITER design. Consequently, the decision was made to convert Argonne's Liquid Metal EXperiment (ALEX) from a 200 degrees C NaK facility to a 350 degrees C lithium facility. The upgraded facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups at Hartmann numbers (M) and interaction parameters (N) in the range of 10 3 to 10 5 in lithium at 350 degrees C. Following completion of the upgrade work, a short performance test was conducted, followed by two longer multiple-hour, MHD tests, all at 230 degrees C. The modified ALEX facility performed up to expectations in the testing. MHD pressure drop and test section voltage distributions were collected at Hartmann numbers of 1000

  13. 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.

  14. A conceptual composite blanket design for the Tokamak type of thermonuclear reactor incorporating thermoelectric pumping of liquid lithium

    International Nuclear Information System (INIS)

    Dutta Gupta, P.B.

    1981-01-01

    The conceptual liquid lithium blanket design for the tokamak type of thermonuclear reactor put forward is a modification of the initial simple but novel design concept enunciated earlier that exploits the availability of suitably oriented magnetic fields and temperature gradients within the blanket to pump the liquid as has been shown feasible by laboratory model experiments. The modular construction of the blanket cells is retained but the earlier simple back to back double spiralling channel module is replaced by a composite unit of three radially nested layer-structures to optimise heat and tritium extraction from the blanket. The layer-structure at the first wall generates liquid lithium circulation by thermoelectric magnetohydrodynamic forces and the segregated double spiralling channels serve as inlet-outlet driving devices. The outermost layer-structure is cooled by helium. Liquid lithium in the intermediate layer-structure is pumped at a very slow rate. The choice of the relative dimensional proportions of the three layer-structure and the channel cross-section, material property and the spiralling contour is of critical importance for the design. This paper presents the design data for a conceptual design of such a blanket with a 5000 MW (th) rating. (author)

  15. 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

  16. Progress on the Fabrication Methods Development for the Korean Test Blanket Module First Wall in the ITER

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Bae, Young Dug; Yoon, Jae Sung; Cho, Seung Yon

    2010-01-01

    A Korean helium cooled molten lithium (HCML) test blanket module (TBM) has been designed to be tested in the International Thermonuclear Experimental Reactor (ITER) TBM and related fabrication methods have been developed especially for the purpose of joining. Since the first wall (FW) of the HCML TBM is composed of a beryllium (Be) as an armor material and a FMS as a structural one, joining with Be to FMS and FMS to FMS should be developed in order to fabricate it

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, B.J., E-mail: Brad.Merrill@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Wong, C.P.C. [General Atomics, San Diego, CA 92186-5608 (United States); Cadwallader, L.C. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Abdou, M.; Morley, N.B. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States)

    2014-10-15

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

  20. Development and qualification of functional materials for the EU Test Blanket Modules: Strategy and R and D activities

    Energy Technology Data Exchange (ETDEWEB)

    Zmitko, M., E-mail: milan.zmitko@f4e.europa.eu [Fusion for Energy (F4E), 08019 Barcelona (Spain); Poitevin, Y. [Fusion for Energy (F4E), 08019 Barcelona (Spain); Boccaccini, L., E-mail: lorenzo.boccaccini@inr.fzk.de [Institut Fuer Neutronenphysik und Reaktortechnik, FZK, D-76021 Karlsruhe (Germany); Salavy, J.-F., E-mail: jfsalavy@cea.fr [CEA/Saclay, DEN/DM2S, F-91191 Gif-sur-Yvette (France); Knitter, R., E-mail: regina.knitter@imf.fzk.de [Institut Fuer Materialforschung III, FZK, D-76021 Karlsruhe (Germany); Moeslang, A., E-mail: anton.moeslang@imf.fzk.de [Institut Fuer Materialforschung I, FZK, D-76021 Karlsruhe (Germany); Magielsen, A.J., E-mail: magielsen@nrg.eu [NRG Petten, 1755 ZG Petten (Netherlands); Hegeman, J.B.J. [NRG Petten, 1755 ZG Petten (Netherlands); Laesser, R. [Fusion for Energy (F4E), 08019 Barcelona (Spain)

    2011-10-01

    Europe has developed two reference tritium breeder blankets concepts for a DEMO fusion reactor: the Helium-Cooled Lithium-Lead and the Helium-Cooled Pebble-Bed. Both will be tested in ITER under the form of Test Blanket Modules (TBMs). The paper reviews the current status of development and qualification of the EU TBMs functional materials; i.e. ceramic solid breeder materials, beryllium/beryllides multiplier materials and Lithium-Lead liquid metal breeder material Pb-15.7Li. For each functional material the main functional/performance requirements with key qualification issues, current status of the R and D activities and the EU development strategy are presented. In the development strategy major steps considered are listed pointing out importance of the 'Development/qualification/procurement plan', currently under elaboration, for definition of a roadmap of further activities aiming at delivery of qualified functional materials to be used in the European TBMs in ITER.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  2. New Monte Carlo results for the TFTR/Lithium Blanket Module system

    International Nuclear Information System (INIS)

    Engholm, B.A.

    1985-01-01

    Neutronics analysis results from Phase II of the TFTR Lithium Blanket Module (LBM) program are reported. Principal activities were analyses of new coverplate and protective plate designs; updating of the MCNP Monte Carlo model of TFTR/LBM; and performing new reference calculations for D-D and D-T plasmas. The new protective plate was found to reduce LBM responses by 20%. Updating the model included a new tally structure in which the LBM is divided into 92 volume elements corresponding to foil locations. A new version of the MCNP surface-source routine was used, along with the latest pointwise cross sections. All flux, tritium and foil responses are stored at NMFECC and are available for comparison with measurements, when the experimental program gets underway

  3. Upgrading the data acquisition and control systems of the European Breeding Blanket Test Facility

    International Nuclear Information System (INIS)

    Mannori, Simone; Sermenghi, Valerio; Utili, Marco; Malavasi, Andrea; Gianotti, Daniel

    2013-01-01

    Highlights: • Data Acquisition and Control Systems (DACS) upgrading of experimental plant for full size thermo hydraulic testing of nuclear subsystems. • DACS development using integrated hardware/software platform with graphical programming (LabVIEW). • Development of simplified models for real-time simulation. • Rapid prototyping with real time simulation of the complete plant. • Using the code developed for the real time simulator for the real plant DACS. -- Abstract: The EBBTF (European Breeding Blanket Test Facility) experimental plant is a key component for the development of the breeding blankets (TBMs test blanket modules, HCLL helium cooled lithium lead and HCPB helium cooled pebble bed types) used by ITER. EBBTF is an experimental plant which provides the double breeding/cooling loops (liquid metal and gas) required for HCLL testing. EBBTF is composed of four subsystems (TBM, IELLLO integrated European lead lithium loop, HE-FUS3 helium fusion loop, version 3 and helium compressor build by ATEKO) with dedicated control systems realized with hardware/software combinations covering 15 years (1995–2010) time span. At the end of 2010 we began to upgrade the HE-FUS3 data acquisition control systems (DACS) replacing the obsolete PLC Siemens S5 with National Instruments Compact FieldPoint and LabVIEW. The control room has been completely reorganized using high resolution monitors and workstations linked with standard Ethernet interfaces. The data acquisition, control, safety and SCADA software has been completely developed in ENEA using LabVIEW. In this paper we are going to discuss the technical difficulties and the solutions that we have used to accomplish the upgrade

  4. Thermal hydraulic and power cycle analysis of liquid lithium blanket designs

    International Nuclear Information System (INIS)

    Misra, B.; Stevens, H.C.; Maroni, V.A.

    1977-01-01

    Thermal hydraulic and power cycle analyses were performed for the first-wall and blanket systems of tokamak-type fusion reactors under a typical set of design and operating conditions. The analytical results for lithium-cooled blanket cells show that with stainless steel as construction material and with no divertor present, the maximum allowable neutron wall loading is approximately 2 MW/m 2 and is limited by thermal stress criteria. With vanadium alloy as construction material and no divertor present, the maximum allowable neutron wall loading is approximately 8 MW/m 2 and is limited by an interplay of constraints imposed on the maximum allowable structural temperature and the minimum allowable coolant inlet temperature. With a divertor these wall loadings can be increased by from 40 to 90 percent. The cost of the vanadium system is found to be competitive with the stainless steel system because of the higher allowable structural temperatures and concomitant higher thermal efficiencies afforded by the vanadium alloys

  5. Fluidized-bed design for ICF reactor blankets using solid-lithium compounds

    International Nuclear Information System (INIS)

    Sucov, E.W.; Malick, F.S.; Green, L.; Hall, B.O.

    1983-01-01

    A fluidized-bed concept for blankets of dry or wetted first-wall ICF reactors using solid-lithium compounds is described. The reaction chamber is a right cylinder, 32 m high and 20 m in diameter; the blanket is composed of 36 steel tanks, 32 m high, which carry the sintered Li 2 O particles in the fluidizing helium gas. Each tank has a radial thickness of 2 m which generates a tritium breeding ration (TBR) of 1.27 and absorbs over 98% of the neutron energy; reducing the thickness to 1.2 m produces a TBR of 1.2 and energy absorption of 97% which satisfy the design goals. Calculations of tritium diffusion through the grains and heat removal from the grains showed that neither could be removed by the carrier gas; tritium and heat are therefore removed by removing the grains themselves by varying the helium flow rate. The particles are continuously fed into the bottom of the tanks at 300 0 C and removed at the top at 475 0 C. Tritium and heat extraction are easily and conveniently done outside the reactor

  6. Thermal-hydraulic and neutronic considerations for designing a lithium-cooled tokamak blanket

    International Nuclear Information System (INIS)

    Chao, J.; Mikic, B.; Todreas, N.

    1978-12-01

    A methodology for the design of lithium cooled blankets is developed. The thermal-hydraulics, neutronics and interactions between them are extensively investigated. In thermal hydraulics, two models illustrate the methodology used to obtain the acceptable ranges for a set of design parameters. The methodology can be used to identify the limiting constraints for a particular design. A complete neutronic scheme is set up for the calculations of the volumetric heating rate as a function of the distance from the first wall, the breeding ratio as a function of the amount of structural material in the blanket, and the radiation damage in terms of atom displacements and gas production rate. Different values of the volume percent of Type-316 stainless steel are assigned in four breeding zones to represent a nonuniformly distributed structural material which satisfies various thermal-hydraulic requirements. The role that the radiation damage plays in the overall design methodology is described. The product of the first wall lifetime and neutron loading is limited by the radiation damage which degrades the mechanical properties of the material

  7. 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.).

  8. 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

  9. Engineering test station for TFTR blanket module experiments

    International Nuclear Information System (INIS)

    Jassby, D.L.; Leinoff, S.

    1979-12-01

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

  10. Direct tritium measurement in lithium titanate for breeding blanket mock-up experiments with D-T neutrons

    International Nuclear Information System (INIS)

    Klix, A.; Ochiai, K.; Nishitani, T.; Takahashi, A.

    2004-01-01

    At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different 6 Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation

  11. Dynamic tritium inventory of a NET/ITER fuel cycle with lithium salt solution blanket

    International Nuclear Information System (INIS)

    Spannagel, G.; Gierszewski, P.

    1991-01-01

    At the Karlsruhe Nuclear Research Center (KfK) a flexible tool is being developed to simulate the dynamics of tritium inventories. This tool can be applied to any tritium handling system, especially to the fuel cycle components of future nuclear fusion devices. This instrument of simulation will be validated in equipment to be operated at the Karlsruhe Tritium Laboratory. In this study tritium inventories in a NET/ITER type fuel cycle involving a lithium salt solution blanket are investigated. The salt solution blanket serves as an example because it offers technological properties which are attractive in modeling the process; the example does not impair the general validity of the tool. Usually, the operation strategy of complex structures will deteriorate due to failures of the subsystems involved. These failures together with the reduced availability ensuing from them will be simulated. The example of this study is restricted to reduced availabilities of two subsystems, namely the reactor and the tritium recovery system. For these subsystems the influence of statistically varying intervals of operation is considered. Strategies we selected which are representative of expected modes of operation. In the design of a fuel cycle, care will be taken that prescribed availabilities of the subsystems can be achieved; however, the description of reactor operation is a complex task since operation breaks down into several campaigns for which rules have been specified which enable determination of whether a campaign has been successful and can be stopped. Thus, it is difficult to predict the overall behavior prior to a simulation which includes stochastic elements. Using the example mentioned above the capabilities of the tool will be illustrated; besides the presentation of results of inventory simulation, the applicability of these data will be discussed. (orig.)

  12. Stress analysis of the tokamak engineering test breeder blanket

    International Nuclear Information System (INIS)

    Huang Zhongqi

    1992-01-01

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

  13. 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.

  14. 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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  16. Optimization of the first wall for the DEMO water cooled lithium lead blanket

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, Julien, E-mail: julien.aubert@cea.fr [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Aiello, Giacomo [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Bachmann, Christian [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Di Maio, Pietro Alessandro [Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Giammusso, Rosario [ENEA C.R. Brasimone, 40032 Camugnano, Bologna (Italy); Li Puma, Antonella; Morin, Alexandre [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Tincani, Amelia [ENEA C.R. Brasimone, 40032 Camugnano, Bologna (Italy)

    2015-10-15

    Highlights: • This paper presents the optimization of the first wall of the water cooled lithium lead DEMO blanket with pressurized water reactor condition and circular channels in order to find the best geometry that can allow the maximum heat flux considering design criteria since an estimate of the engineering limit of the first wall heat load capacity is an essential input for the decision to implement limiters in DEMO. • An optimization study was carried out for the flat first wall design of the DEMO Water-Cooled Lithium Lead considering thermal and mechanical constraint functions, assuming T{sub inlet}/T{sub outlet} equal to 285 °C/325 °C, based on geometric design parameters. • It became clear that through the optimization the advantages of a waved First Wall are diminished. • The analysis shows that the maximum heat load could achieve 2.53 MW m{sup −2}, but considering assumptions such as a coolant velocity ≤8 m/s, pipe diameter ≥5 mm and a total first wall thickness ≤22 mm, heat flux is limited to 1.57 MW m{sup −2}. - Abstract: The maximum heat load capacity of a DEMO First Wall (FW) of reasonable cost may impact the decision of the implementation of limiters in DEMO. An estimate of the engineering limit of the FW heat load capacity is an essential input for this decision. This paper describes the work performed to optimize the FW of the Water Cooled Lithium-Lead (WCLL) blanket concept for DEMO fusion reactor in order to increase its maximum heat load capacity. The optimization is based on the use of water at typical Pressurised Water Reactors conditions as coolant. The present WCLL FW with a waved plasma-faced surface and with circular channels was studied and the heat load limit has been predicted with FEM analysis equal to 1.0 MW m{sup −2} with respect to the Eurofer temperature limit. An optimization study was then carried out for a flat FW design considering thermal and mechanical constraints assuming inlet and outlet

  17. Preliminary piping layout and integration of European test blanket modules subsystems in ITER CVCS area

    Energy Technology Data Exchange (ETDEWEB)

    Tarallo, Andrea, E-mail: andrea.tarallo@unina.it [CREATE, University of Naples Federico II, DII, P.le Tecchio, 80, 80125 Naples (Italy); Mozzillo, Rocco; Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, DII, P.le Tecchio, 80, 80125 Naples (Italy); Aiello, Antonio; Utili, Marco [ENEA UTIS, C.R. Brasimone, Bacino del Brasimone, I-40032 Camugnano, BO (Italy); Ricapito, Italo [TBM& MD Project, Fusion for Energy, EU Commission, Carrer J. Pla, 2, Building B3, 08019 Barcelona (Spain)

    2015-04-15

    Highlights: • The use of human modeling tools for piping design in view of maintenance is discussed. • A possible preliminary layout for TBM subsystems in CVCS area has been designed with CATIA. • A DHM-based method to quickly check for maintainability of piping systems is suggested. - Abstract: This paper explores a possible integration of some ancillary systems of helium-cooled lithium lead (HCLL) and helium-cooled pebble-bed (HCPB) test blanket modules in ITER CVCS area. Computer-aided design and ergonomics simulation tools have been fundamental not only to define suitable routes for pipes, but also to quickly check for maintainability of equipment and in-line components. In particular, accessibility of equipment and systems has been investigated from the very first stages of the design using digital human models. In some cases, the digital simulations have resulted in changes in the initial space reservations.

  18. Feasibility study of a neutron activation system for EU test blanket systems

    Energy Technology Data Exchange (ETDEWEB)

    Tian, Kuo, E-mail: kuo.tian@kit.edu [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Calderoni, Pattrick [Fusion for Energy(F4E), Barcelona (Spain); Ghidersa, Bradut-Eugen; Klix, Axel [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany)

    2016-11-01

    Highlights: • This paper summarizes the technical baseline and preliminary design of EU TBM Neutron Activation System, briefly describes the key components, and outlines the major integration challenges. - Abstract: The Neutron Activation System (NAS) for the EU Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebble Bed (HCPB) Test Blanket Systems (TBSs) is an instrument that is proposed to determine the absolute neutron fluence and absolute neutron flux with information on the neutron spectrum in selected positions of the corresponding Test Blanket Modules (TBMs). In the NAS activation probes are exposed to the ITER neutron flux for periods ranging from several tens of seconds up to a full plasma pulse length, and the induced gamma activities are subsequently measured. The NAS is composed of a pneumatic transfer system and a counting station. The pneumatic transfer system includes irradiation ends in TBMs, transfer pipes, return gas pipes, a transfer station with a distributor (carousel), and a pressurized gas driving system, while the counting station consists of gamma ray detectors, signal processing electronic devices, and data analyzing software for neutron source strength evaluation. In this paper, a brief description on the proposed TBM NAS as well as the key components is presented, and the integration challenges of TBM NAS are outlined.

  19. Activation and afterheat analyses for the HCPB test blanket

    International Nuclear Information System (INIS)

    Pereslavtsev, P.; Fischer, U.

    2007-01-01

    The Helium-Cooled Pebble Bed (HCPB) blanket is one of two breeder blanket concepts developed in the framework of the European Fusion Technology Programme for performance tests in ITER. The recent development programme focussed on the detailed engineering design of the Test Blanket Module (TBM) and associated systems including the assessment of safety and licensing related issues with the objective to prepare for a preliminary Safety Report. To provide a sound data basis for the safety analyses of the HCPB TBM system in ITER, the afterheat and activity inventories were assessed making use of a code system that allows performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. A suitable MCNP model of a 20 degree ITER torus sector with an integrated TBM of the HCPB PI (Plant Integration) type in the horizontal test blanket port was developed and adapted to the requirements for coupled 3D neutron transport and activation calculations. Two different irradiation scenarios were considered in the coupled 3D neutron transport and activation calculations. The first one is representative for the TBM irradiation in ITER with a total of 9000 neutron pulses over a three (calendar) years period. It was simulated by a continuous irradiation for 3 years minus the last month and a discontinuous irradiation with 250 pulses (420 s pulse length, 1200 s power-off in between) over the last month. The second (conservative) irradiation scenario assumes an extended irradiation time over the full anticipated lifetime of ITER according to the M-DRG-1 irradiation scenario with a total first wall fluence of 0.3 MWa/m 2 . For both irradiation scenarios the radioactivity inventories, the afterheat and the contact gamma dose were calculated as function of the decay time. Data were processed for the total activity and afterheat of the TBM, its constituting components and materials including their

  20. Strategy for the development of EU Test Blanket Systems instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Calderoni, P., E-mail: Pattrick.Calderoni@f4e.europa.eu; Ricapito, I.; Poitevin, Y.

    2013-10-15

    Highlights: • We developed a strategy for the development of instrumentation for EU ITER TBSs. • TBSs instrumentation functions: safety, operation and scientific mission. • Described activities are in support of ITER design review process. -- Abstract: The instrumentation of the HCLL and HCPB Test Blanket System is fundamental in ensuring that ITER safety and operational requirements are satisfied as well as in enabling the scientific mission of the TBM program. It carries out three essential functions: (i) safety, intended as compliance with ITER requirements toward public and workers protection; (ii) system control, intended as compliance with ITER operational requirements and investment protection; and (iii) scientific mission, intended as validating technology and predictive tools for blanket concepts relevant to fusion energy systems. This paper describes the strategy for instrumentation development by providing details of the following five steps to be implemented in procured activities in the short to mid-term (3–4 years): (i) provide mapping of sensors requirements based on critical review of preliminary design data; (ii) develop functional specifications for TBS sensors based on the analysis of operative conditions in the various ITER buildings in which they are located; (iii) assess availability of commercial sensors against developed specifications; (iv) develop prototypes when no available solution is identified; and (v) perform single effect tests for the most critical solicitations and post-test examination of commercial products and prototypes. Examples of technology assessment in two technical areas are included to reinforce and complement the strategy description.

  1. Radwaste management aspects of the test blanket systems in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Laan, J.G. van der, E-mail: JaapG.vanderLaan@iter.org [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Canas, D. [CEA, DEN/DADN, centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Chaudhari, V. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India); Iseli, M. [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Kawamura, Y. [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan); Lee, D.W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Petit, P. [European Commission, DG ENER, Brussels (Belgium); Pitcher, C.S.; Torcy, D. [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Ugolini, D. [Fusion for Energy, Barcelona (Spain); Zhang, H. [China Nuclear Energy Industry Corporation, Beijing 100032 (China)

    2016-11-01

    Highlights: • Test Blanket Systems are operated in ITER to test tritium breeding technologies. • The in-vessel parts of TBS become radio-active during the ITER nuclear phase. • For each TBM campaign the TBM, its shield and the Pipe Forests are removed. • High tritium contents and novel materials are specific TBS radwaste features. • A preliminary assessment confirmed RW routing, provided its proper conditioning. - Abstract: Test Blanket Systems (TBS) will be operated in ITER in order to prepare the next steps towards fusion power generation. After the initial operation in H/He plasmas, the introduction of D and T in ITER will mark the transition to nuclear operation. The significant fusion neutron production will give rise to nuclear heating and tritium breeding in the in-vessel part of the TBS. The management of the activated and tritiated structures of the TBS from operation in ITER is described. The TBS specific features like tritium breeding and power conversion at elevated temperatures, and the use of novel materials require a dedicated approach, which could be different to that needed for the other ITER equipment.

  2. European Helium Cooled Pebble Bed (HCPB) test blanket. ITER design description document. Status 1.12.1996

    International Nuclear Information System (INIS)

    Albrecht, H.; Boccaccini, L.V.; Dalle Donne, M.; Fischer, U.; Gordeev, S.; Hutter, E.; Kleefeldt, K.; Norajitra, P.; Reimann, G.; Ruatto, P.; Schleisiek, K.; Schnauder, H.

    1997-04-01

    The Helium Cooled Pebble Bed (HCPB) blanket is based on the use of separate small lithium orthosilicate and beryllium pebble beds placed between radial toroidal cooling plates. The cooling is provided by helium at 8 MPa. The tritium produced in the pebble beds is purged by the flow of helium at 0.1 MPa. The structural material is martensitic steel. It is foreseen, after an extended R and D work, to test in ITER a blanket module based on the HCPB design, which is one of the two European proposals for the ITER Test Blanket Programme. To facilitate the handling operation the Blanket Test Module (BTM) is bolted to a surrounding water cooled frame fixed to the ITER shield blanket back plate. For the design of the test module, three-dimensional Monte Carlo neutronic calculations and thermohydraulic and stress analyses for the operation during the Basic Performance Phase (BPP) and during the Extended Performance Phase (EPP) of ITER have been performed. The behaviour of the test module during LOCA and LOFA has been investigated. Conceptual designs of the required ancillary loops have been performed. The present report is the updated version of the Design Description Document (DDD) for the HCPB Test Module. It has been written in accordance with a scheme given by the ITER Joint Central Team (JCT) and accounts for the comments made by the JCT to the previous version of this report. This work has been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhne and it is supported by the European Union within the European Fusion Technology Program. (orig.) [de

  3. The Test Blanket Modules project in Europe: From the strategy to the technical plan over next ten years

    International Nuclear Information System (INIS)

    Poitevin, Y.; Zmitko, M.; Orco, G. dell; Laesser, R.; Diegele, E.; Sundstroem, J.; Boccaccini, L.; Salavy, J.-F.

    2006-01-01

    The testing of Breeding Blanket concepts in ITER is recognized as an essential milestone in the development of a future reactor ensuring tritium self-sufficiency, extraction of high grade heat and electricity production. Europe is currently developing two reference breeding blankets for DEMO reactor specifications that will be tested in ITER: the Helium-Cooled Lithium-Lead (HCLL) blanket which uses the eutectic Pb-15. 7 Li as both breeder and neutron multiplier, and the Helium-Cooled Pebble-Bed (HCPB) blanket which features lithiated ceramic pebbles (Li 4 SiO 4 or Li 2 TiO 3 ) as breeder and beryllium pebbles as neutron multiplier. Both blankets are using the pressurized He technology for heat extraction (8 MPa, inlet/outlet temperature 300/500 o C) and a 9% CrWVTa Reduced Activation Ferritic Martensitic (RAFM) steel as structural material, the EUROFER. Referring to the so called '' fast-track '' EU scenario, those concepts are intended to be tested in ITER, getting the maximum of information required for launching the DEMO blanket design and construction after the first 10 years of ITER operation. For that, the EU has adopted a blanket testing strategy based on the development of Test Blanket Modules (TBMs) that are expected to use DEMO relevant technologies and are designed for each ITER plasma phase to optimize the feedback and to avoid any impact on ITER availability. Following the decision on ITER construction, the EU has reviewed and detailed the fundamental elements for an implementation of the future EU TBMs Project aimed at delivering TBMs Systems to ITER under suitable schedule and acceptance standards. For that the following items have been analyzed in detail and are reported in the present paper: · Impact of the ITER environment (design, standards, schedule, operational scheme) on the TBM systems design and development plan · Project technical plan with focus on the next ten years up to the installation of the first TBMs in ITER · Project risk

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

    International Nuclear Information System (INIS)

    Petrizzi, L.

    1994-01-01

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

  5. Vibration damage testing of thermal barrier fibrous blanket insulation

    International Nuclear Information System (INIS)

    Black, W.E.; Betts, W.S.

    1984-01-01

    GA Technologies is engaged in a long-term, multiphase program to determine the vibration characteristics of thermal barrier components leading to qualification of assemblies for High Temperature Gas-Cooled Reactor (HTGR) service. The phase of primary emphasis described herein is the third in a series of acoustic tests and uses as background the more elemental tests preceding it. Two sizes of thermal barrier coverplates with one fibrous blanket insulation type were tested in an acoustic environment at sound pressure levels up to 160 dB. Three tests were conducted using sinusoidal and random noise for up to 200 h duration at room temperature. Frequent inspections were made to determine the progression of degradation using definition of stages from a prior test program. Initially the insulation surface adjacent to the metallic seal sheets (noise side) assumed a chafed or polished appearance. This was followed by flattening of the as-received pillowed surface. This stage was followed by a depression being formed in the vicinity of the free edge of the coverplate. Next, loose powder from within the blanket and from fiber erosion accumulated in the depression. Prior experience showed that the next stage of deterioration exhibited a consolidation of the powder to form a local crust. In this test series, this last stage generally failed to materialize. Instead, surface holes generated by solid ceramic particulates (commonly referred to as 'shot') constituted the stage following powdering. With the exception of some manufacturing-induced anomalies, the final stage, namely, gross fiber breakup, did not occur. It is this last stage that must be prevented for the thermal barrier to maintain its integrity. (orig./GL)

  6. Response distributions of 6LiF and 7LiF thermoluminescence dosimeters in lithium blanket assemblies

    International Nuclear Information System (INIS)

    Maekawa, Hiroshi; Kusano, Jyo-ichi; Seki, Yasushi

    1976-11-01

    Measurement of the radiation-heating rate distribution in the fusion blanket is as important as measurement of the fission-rate distribution in a fission reactor. To obtain the information of radiation heating, the response (integral glow value) distributions in pseudo-spherical lithium assemblies with and without a graphite reflector were measured with 6 LiF and 7 LiF TLD's. The measured responses are normalized to values per source neutron. Experimental error is about 35%, and the error in positions of TLD's is about +- 3 mm. The experimental results are compared with those of calculation using RADHEAT code system and ENDF/B-III data file. (auth.)

  7. Welding techniques development of CLAM steel for Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Li Chunjing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China)], E-mail: lcj@ipp.ac.cn; Huang Qunying; Wu Qingsheng; Liu Shaojun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China); Lei Yucheng [Jiangsu University, Zhenjiang, Jiangsu, 212013 (China); Muroga, Takeo; Nagasaka, Takuya [National Institute for Fusion Science, Toki, Jifu, 509-5292 (Japan); Zhang Jianxun [Xi' an Jiaotong University, Xi' an, Shanxi, 710049 (China); Li Jinglong [Northwestern Polytechnical University, Xi' an, Shanxi, 710072 (China)

    2009-06-15

    Fabrication techniques for Test Blanket Module (TBM) with CLAM are being under development. Effect of surface preparation on the HIP diffusion bonding joints was studied and good joints with Charpy impact absorbed energy close to that of base metal have been obtained. The mechanical properties test showed that effect of HIP process on the mechanical properties of base metal was little. Uniaxial diffusion bonding experiments were carried out to study the effect of temperature on microstructure and mechanical properties. And preliminary experiments on Electron Beam Welding (EBW), Tungsten Inert Gas (TIG) Welding and Laser Beam Welding (LBW) were performed to find proper welding techniques to assemble the TBM. In addition, the thermal processes assessed with a Gleeble thermal-mechanical machine were carried out as well to assist the fusion welding research.

  8. 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.)

  9. Helium Loop for the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Neuberger, H.; Boccaccini, L.V.; Ghidersa, B. E.; Jin, X.; Meyder, R.

    2006-01-01

    In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group, the Helium loop for the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) in ITER has been investigated with regard to the layout definition, selection of components, control, dimensioning and integration. This paper presents the status of development. The loop design for the HCPB-TBM in ITER will mainly base on the experience gained from Helium Loop Karlsruhe (HELOKA) which is currently developed at the FZK for experiments under ITER relevant conditions. The ITER loop will be equipped with similar components like HELOKA and will mainly consist of a circulator with variable speed drive, a recuperator, an electric heater, a cooler, a dust filter and auxilary components e.g. pipework and valves. A Coolant Purification System (CPS) and a Pressure Control System (PCS) are foreseen to meet the requirements on coolant conditioning. To prepare a TBM for a new experimental campaign, a succession of operational states like '' cold maintenance '', '' baking '' and '' cold standby '' is required. Before a pulse operation, a '' hot stand-by '' state should be achieved providing the TBM with inlet coolant at nominal conditions. This operation modus is continued in the dwell time waiting for the successive pulse. A '' tritium out-gassing '' will be also required after several TBM-campaigns to remove the inventory rest of T in the beds for measurement purpose. The dynamic circuit behaviour during pulses, transition between different operational states as well as the behaviour in accident situations are investigated with RELAP. The main components of the loop will be accommodated inside the Tokamak Cooling Water System(TCWS)- vault from where the pipes require connection to the TBM which is attached to port 16 of the vacuum vessel. Therefore pipes across the ITER- building of about 110 m in length (each) are required. Additional equipment is also located in the port cell

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-01

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

  11. The thermo-mechanical design of the water cooled PB-17Li test blanket module for ITER

    International Nuclear Information System (INIS)

    Nardi, C.; Palmieri, A.; Pinna, T.; Porfini, M.T.; Rapisarda, M.; Roccella, M.; Futterer, M.; Lucca, F.

    1998-01-01

    The Water Cooled Lithium Lead (WCLL) blanket is one of the two European concepts to be further developed. A Test Blanket Module (TBM) representative of the DEMO blanket shall be tested in ITER. This paper reports on the activities related to the thermo-mechanical design analysis, taking into account the electromagnetic and neutronic loads in normal and off normal conditions. These loads were applied to a finite elements model of the structure, and the structural response was compared to the allowable value, dependent on the operating conditions. Besides the loads assumed by the design specifications (pressure, temperature, etc), electro-mechanical and thermal loads have been evaluated. A model of the TBM has been performed to compute the loads related to the electromagnetic effects of a centered plasma disruption. The thermal loads have been evaluated considering the heat deposition from the plasma and from the neutrons. The neutronic analysis has been carried out also in order to evaluate the shielding characteristics of the TBM. Taking into account the thermal and mechanical loads a fracture mechanics analysis has been carried out. From this analysis the J Ic parameter was evaluated at the crack tip and compared with the allowable value. The work carried out showed that the TBM present design fulfills ITER normal operation requirements. (authors)

  12. Feasibility analysis of vacuum sieve tray for tritium extraction in the HCLL test blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Okino, Fumito, E-mail: fumito.okino@iae.kyoto-u.ac.jp [Kyoto University Institute of Advanced Energy, 611-0011 Gokasho, Uji, Kyoto (Japan); Calderoni, Pattrick [Fusion For Energy, 08019 Barcelona (Spain); Kasada, Ryuta; Konishi, Satoshi [Kyoto University Institute of Advanced Energy, 611-0011 Gokasho, Uji, Kyoto (Japan)

    2016-11-01

    Highlights: • The authors discovered faster mass transport on a droplet falling in a vacuum. • Primary cause of the hydrogen release from droplet is by the oscillation of a droplet. • The spherical oscillation induces the internal advection and enhances mass transfer. • This assumption agreed with previous experimental results. - Abstract: This paper describes the quantitative analysis for the design of a tritium extraction system that uses liquid PbLi droplets in vacuum (Vacuum Sieve Tray, VST), for application to the ITER helium-cooled lithium lead (HCLL) test blanket system (TBS). The parametric dependences of tritium extraction efficiency from the main geometrical features such as initial droplet velocity, nozzle head height, nozzle diameter, and flow rate are discussed. With nozzle diameters between 0.4 and 0.6 mm, extraction efficiency is estimated from 0.77 to 0.96 at the falling height of 0.5 m, with flow rate between 0.2 and 1.0 kg/s. The device has a height of 1.6 m, within the external dimensions of the HCLL Test Blanket Module (TBM), and no additional pumping power is required. The attained results are considered attractive not only for ITER, but also in view of the application of the VST concept as a candidate tritium extraction system for the European Union's demonstration fusion reactor (DEMO). The extraction efficiency of a single droplet column, which is the basis of the design analysis presented, has been validated experimentally with hydrogen. However, further experiments are required on an integrated system with size relevant to the proposed HCLL-TBS design to validate system-level effects, particularly regarding the desorption process in an array of multiple droplets.

  13. 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

  14. Nuclear maintenance strategy and first steps for preliminary maintenance plan of the EU HCLL & HCPB Test Blanket Systems

    Energy Technology Data Exchange (ETDEWEB)

    Galabert, Jose, E-mail: jose.galabert@f4e.europa.eu [F4E Fusion for Energy, EU Domestic Agency, c/Josep Pla, 2. B3, 08019, Barcelona (Spain); Hopper, Dave [AMEC Foster Wheeler, Faraday Street, Birchwood Park, WA3 6GN (United Kingdom); Neviere, Jean-Cristophe [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, 13067, St. Paul Lez Durance Cedex (France); Nodwell, David [CCFE, Culham Science Centre, Abingdon, OX14 3DB, Oxfordshire (United Kingdom); Pascal, Romain [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, 13067, St. Paul Lez Durance Cedex (France); Poitevin, Yves; Ricapito, Italo [F4E Fusion for Energy, EU Domestic Agency, c/Josep Pla, 2. B3, 08019, Barcelona (Spain); White, Gareth [AMEC Foster Wheeler, Faraday Street, Birchwood Park, WA3 6GN (United Kingdom)

    2017-03-15

    Highlights: • Nuclear maintenance strategy for the two European (EU) Test Blanket Systems (TBS): i/. Helium Cooled Lead Lithium (HCLL) and ii/. Helium Cooled Pebble Bed (HCPB). • Preliminary identification of maintenance tasks for most relevant components of the EU HCLL & HCPB TBS. • Preliminary feasibility analysis for hands-on maintenance tasks of some relevant components of the European Test Blanket Systems. • Design recommendations for enhancement of the European Test Blanket Systems maintainability. - Abstract: This paper gives an overview of nuclear maintenance strategy to be followed for the European HCLL & HCPB Test Blanket Systems (TBS) to be installed in ITER. One of the several core documents to prepare in view of their licensing is their respective ‘Maintenance Plan’. This document is fundamental for ensuring sound performance and safety of the TBS during ITER’s operational phase and shall include, amongst others, relevant information on: maintenance organization, preventive and corrective maintenance task procedures, condition monitoring for key components, maintenance work planning, and a spare parts plan, just to mention some of the key topics. In compliance with the ITER Plant Maintenance policy, first steps have been taken aimed at defining nuclear maintenance strategy for some of the most relevant HCLL & HCPB TBS components, conducted by F4E in collaboration with industry. After a brief recall of maintenance strategy of the TBM Program (PBS-56), this paper analyses main features of EU HCLL & HCPB TBS maintainability and identifies, at their conceptual design phase, a preliminary list of maintenance tasks to be developed for their most representative components. In addition, the paper also presents the first nuclear maintenance studies conducted for replacement of the Q{sub 2} Getter Beds, identifying some design recommendations for their sound maintainability.

  15. Nuclear maintenance strategy and first steps for preliminary maintenance plan of the EU HCLL & HCPB Test Blanket Systems

    International Nuclear Information System (INIS)

    Galabert, Jose; Hopper, Dave; Neviere, Jean-Cristophe; Nodwell, David; Pascal, Romain; Poitevin, Yves; Ricapito, Italo; White, Gareth

    2017-01-01

    Highlights: • Nuclear maintenance strategy for the two European (EU) Test Blanket Systems (TBS): i/. Helium Cooled Lead Lithium (HCLL) and ii/. Helium Cooled Pebble Bed (HCPB). • Preliminary identification of maintenance tasks for most relevant components of the EU HCLL & HCPB TBS. • Preliminary feasibility analysis for hands-on maintenance tasks of some relevant components of the European Test Blanket Systems. • Design recommendations for enhancement of the European Test Blanket Systems maintainability. - Abstract: This paper gives an overview of nuclear maintenance strategy to be followed for the European HCLL & HCPB Test Blanket Systems (TBS) to be installed in ITER. One of the several core documents to prepare in view of their licensing is their respective ‘Maintenance Plan’. This document is fundamental for ensuring sound performance and safety of the TBS during ITER’s operational phase and shall include, amongst others, relevant information on: maintenance organization, preventive and corrective maintenance task procedures, condition monitoring for key components, maintenance work planning, and a spare parts plan, just to mention some of the key topics. In compliance with the ITER Plant Maintenance policy, first steps have been taken aimed at defining nuclear maintenance strategy for some of the most relevant HCLL & HCPB TBS components, conducted by F4E in collaboration with industry. After a brief recall of maintenance strategy of the TBM Program (PBS-56), this paper analyses main features of EU HCLL & HCPB TBS maintainability and identifies, at their conceptual design phase, a preliminary list of maintenance tasks to be developed for their most representative components. In addition, the paper also presents the first nuclear maintenance studies conducted for replacement of the Q_2 Getter Beds, identifying some design recommendations for their sound maintainability.

  16. Preparation of acceptance tests and criteria for the Test Blanket Systems to be operated in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Laan, J.G. van der, E-mail: JaapG.vanderLaan@iter.org [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Cuquel, B. [AIRBUS Defence and Space S.A.S., 13115 Saint Paul Lez Durance (France); Demange, D.; Ghidersa, B.-E. [Karlsruhe Institute of Technology, Karlsruhe (Germany); Giancarli, L.M.; Iseli, M.; Jourdan, T. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Nevière, J.-C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Pascal, R.; Ring, W. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • Initial guideline for acceptance testing and acceptance criteria for Test Blanket Systems in ITER. • These tests complement those required by the applicable codes and standards, and regulations. • Completion of TBS manufacture will be followed by Factory Acceptance Testing, prior to shipment. • Next steps are “Reception Inspection Tests”, and on-site pre-installation and components tests. • This guideline allows the detailing of the TBS specific test plans and their scheduling. - Abstract: This paper describes the main acceptance criteria and required acceptance tests for the components of the six Test Blanket Systems to be installed and operated in ITER. It summarizes the guide-line toward the establishment of detailed test plans for the TBS, starting from the end-product at the ITER Members factories, and to generally define the type of tests that have to be performed on the ITER site after shipment, and/or prior to the systems final commissioning phase.

  17. Materials development for ITER shielding and test blanket in China

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.M., E-mail: Chenjm@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wu, J.H.; Liu, X.; Wang, P.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wang, Z.H.; Li, Z.N. [Ningxia Orient Non-ferrous Metals Group Co. Ltd., P.O. Box 105, Shizuishan (China); Wang, X.S.; Zhang, P.C. [China Academy of Engineering Physics, P.O. Box 919-71, Mianyang 621900 (China); Zhang, N.M.; Fu, H.Y.; Liu, D.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China)

    2011-10-01

    China is a member of the ITER program and is developing her own materials for its shielding and test blanket modules. The materials include vacuum-hot-pressing (VHP) Be, CuCrZr alloy, 316L(N) and China low activation ferritic/martensitic (CLF-1) steels. Joining technologies including Be/Cu hot isostatic pressing (HIP) and electron beam (EB) weldability of 316L(N) were investigated. Chinese VHP-Be showed good properties, with BeO content and ductility that satisfy the ITER requirements. Be/Cu mock-ups were fabricated for Be qualification tests at simulated ITER vertical displacement event (VDE) and heat flux cycling conditions. Fine microstructure and good mechanical strength of the CuCrZr alloy were achieved by a pre-forging treatment, while the weldability of 316L(N) by EB was demonstrated for welding depths varying from 5 to 80 mm. Fine microstructure, high strength, and good ductility were achieved in CLF-1 steel by an optimized normalizing, tempering and aging procedure.

  18. Shutdown dose rate analysis of European test blanket modules shields in ITER Equatorial Port #16

    Energy Technology Data Exchange (ETDEWEB)

    Juárez, Rafael, E-mail: rjuarez@ind.uned.es [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain); Sauvan, Patrick; Perez, Lucia [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain); Panayotov, Dobromir; Vallory, Joelle; Zmitko, Milan; Poitevin, Yves [Fusion for Energy (F4E), Torres Diagonal Litoral B3, Josep Pla 2, Barcelona 08019 (Spain); Sanz, Javier [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain)

    2016-11-01

    Highlights: • Nuclear analysis for European TBMs and shields, in ITER Equatorial Port #16, has been conducted in support of the ‘Concept Design Review’ from ITER. • The objective of the work is the characterization of the Shutdown Dose Rates at Equatorial Port #16 interspace. • The role played by the TBM and TBM shields, the equatorial port gaps and the vacuum vessel permeation, in terms of neutron flux transmission is assessed. • The role played by the TBM, TBM shields, Port Plug Frame, Pipe Forest and the machine in terms of activation is also investigated. - 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). An essential element of the Conceptual Design Review (CDR) of these TBSs is the demonstration of capability of Test Blanket Modules (TBM) and their shields to fulfil their function and comply with the design requirements. One of the TBM shields highly relevant design aspects is the project target for shutdown dose rates (SDDR) in the interspace. We investigated two functions of the TBMs and TBM shields—the neutron flux attenuation along the shields, and the reduction of the activation of the components contributing to SDDR. It is shown that TBMs and TBM shields reduce significantly the neutron flux in the port plug (PP). In terms of neutron flux attenuation, the TBM shield provides sufficient neutron flux reduction, being responsible for 5 × 10{sup 6} n/cm{sup 2} s at port interspace, while the EPP gaps and BSM gaps are responsible for 5 × 10{sup 7} n/cm{sup 2} s each. When considering closed upper, lower and lateral neighbour equatorial ports (thus, excluding the cross-talk between ports), a SDDR of 121 μSv/h averaged near the port closure flange was obtained, out of which, only 4 μSv/h are due to the activation of TBMs and TBM shields. Maximum SDDR in the range

  19. 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

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

    International Nuclear Information System (INIS)

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

    1995-07-01

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

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

    International Nuclear Information System (INIS)

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

    1991-04-01

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

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

    International Nuclear Information System (INIS)

    Malang, S.; Mattas, R.

    1994-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  4. Main maintenance operations for Test Blanket Systems in ITER TBM port cells

    Energy Technology Data Exchange (ETDEWEB)

    Pascal, R., E-mail: romain.pascal@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Cortes, P.; Friconneau, J.-P.; Giancarli, L.M.; Gotewal, K.K.; Iseli, M.; Kim, B.Y.; Levesy, B.; Martins, J.-P.; Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Nevière, J.-C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Patisson, L. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Siarras, A. [Sogetti, Parc de la Duranne, 13857 Aix-en-Provence (France); Tesini, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: • The Test Blanket System components layout in Port Cell room is described. • The maintenance of the two Test Blanket Systems in ITER port cell is addressed. • The overall replacement/maintenance strategy is defined. • The main maintenance tasks of the systems are discussed. • The maintenance strategy and required tools are presented. -- Abstract: Each Test Blanket System in ITER is formed by an in-vessel component, the Test Blanket Module, and several associated ancillary systems (coolant and Tritium systems, instrumentation and control systems). The paper describes the overall replacement/maintenance strategy and the main maintenance tasks that have to be considered in the design of the systems. It shows that there are no critical issues.

  5. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2010-06-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  6. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2007-08-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  7. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    International Nuclear Information System (INIS)

    Lee C. Cadwallader

    2007-01-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with 'generic' component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance

  8. 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)

  9. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    International Nuclear Information System (INIS)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available

  10. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    Energy Technology Data Exchange (ETDEWEB)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available.

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

    International Nuclear Information System (INIS)

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

    1988-05-01

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

  12. 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

  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. Development and testing of a zero stitch MLI blanket using plastic pins for space use

    Science.gov (United States)

    Hatakenaka, Ryuta; Miyakita, Takeshi; Sugita, Hiroyuki; Saitoh, Masanori; Hirai, Tomoyuki

    2014-11-01

    New types of MLI blanket have been developed to achieve high thermal performance while maintaining production and assembly workability equivalent to the conventional type. Tag-pins, which are widely used in commercial applications to hook price tags to products, are used to fix the films in place and the pin material is changed to polyetheretherketone (PEEK) for use in space. Thermal performance is measured by using a boil-off calorimeter, in which a rectangular liquid nitrogen tank is used to evaluate the degradation at the bending corner and joint of the blanket. Zero-stitch- and multi-blanket-type MLIs show significantly improved thermal performance (ɛeff is smaller than 0.0050 at room temperature) despite having the same fastener interface as traditional blankets, while the venting design and number of tag-pins are confirmed as appropriate in a depressurization test.

  15. 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.

  16. 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.

  17. Test module in NET for a self-cooled liquid metal blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Arheidt, K.; Fischer, U.

    1989-01-01

    The application of a self-cooled liquid metal blanket concept to the condition of a DEMO-reactor and its testing in NET is described. The neutronics analysis shows that tritium self-sufficiency can be achieved without beryllium multiplier if breeding blankets are arranged at both outboard and inboard side of the torus or, using beryllium as multiplier, with outboard breeding only. First estimates indicate that it should be possible to test all relevant features of the concept in one of the horizontal plug positions of NET. (author). 6 refs.; 7 figs.; 1 tab

  18. 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

  19. Vanadium—lithium in-pile loop for comprehensive tests of vanadium alloys and multipurpose coatings

    Science.gov (United States)

    Lyublinski, I. E.; Evtikhin, V. A.; Ivanov, V. B.; Kazakov, V. A.; Korjavin, V. M.; Markovchev, V. K.; Melder, R. R.; Revyakin, Y. L.; Shpolyanskiy, V. N.

    1996-10-01

    The reliable information on design and material properties of self-cooled Li sbnd Li blanket and liquid metal divertor under neutron radiation conditions can be obtained using the concept of combined technological and material in-pile tests in a vanadium—lithium loop. The method of in-pile loop tests includes studies of vanadium—base alloys resistance, weld resistance under mechanical stress, multipurpose coating formation processes and coatings' resistance under the following conditions: high temperature (600-700°C), lithium velocities up to 10 m/s, lithium with controlled concentration of impurities and technological additions, a neutron load of 0.4-0.5 MW/m 2 and level of irradiation doses up to 5 dpa. The design of such an in-pile loop is considered. The experimental data on corrosion and compatibility with lithium, mechanical properties and welding technology of the vanadium alloys, methods of coatings formation and its radiation tests in lithium environment in the BOR-60 reactor (fast neutron fluence up to 10 26 m -2, irradiation temperature range of 500-523°C) are presented and analyzed as a basis for such loop development.

  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. On the optimization of the first wall of the DEMO water-cooled lithium lead outboard breeding blanket equatorial module

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A., E-mail: pietroalessandro.dimaio@unipa.it; Arena, P.; Bongiovì, G.; Chiovaro, P.; Forte, R.; Garitta, S.

    2016-11-01

    Highlights: • The geometric optimization of the DEMO WCLL blanket module first wall has been performed, maximizing the heat flux it may safely undergo. • Attention has been focused on the FW flat concept endowed with square cooling channels. • A theoretical-computational approach based on the finite element method (FEM) has been followed, adopting a qualified commercial FEM code. • Four optimized FW configurations have been found to safely withstand a heat flux up to 2 MW/m{sup 2} fulfilling all the rules prescribed by safety codes. - Abstract: Within the framework of EUROfusion R&D activities a research campaign has been carried out at the University of Palermo in order to investigate the thermo-mechanical performances of the DEMO water-cooled lithium lead (WCLL) breeding blanket first wall (FW). The research campaign has been mainly focused on the optimization of the FW geometric configuration in order to maximize the heat flux it may safely withstand fulfilling all the thermal, hydraulic and mechanical requirements foreseen by safety codes. Attention has been focused on the FW flat concept endowed with square cooling channels and the potential influence of its four main geometrical parameters on its thermo-mechanical performances has been assessed performing a parametric analysis by means of a qualified commercial finite element method code. A set of 5929 different FW geometric configurations has been considered and the thermal performances of each one of them have been numerically assessed in case it undergoes 26 different values of heat flux on its plasma-facing surface. The resulting 154154 thermal analyses have allowed to select those cases fulfilling the adopted thermal-hydraulic requirements, whose thermo-mechanical performances have been numerically assessed under both normal operation and over-pressurization steady state loading scenarios to check whether they met the mechanical requirements prescribed by the pertaining SDC-IC safety rules. Four

  2. On the optimization of the first wall of the DEMO water-cooled lithium lead outboard breeding blanket equatorial module

    International Nuclear Information System (INIS)

    Di Maio, P.A.; Arena, P.; Bongiovì, G.; Chiovaro, P.; Forte, R.; Garitta, S.

    2016-01-01

    Highlights: • The geometric optimization of the DEMO WCLL blanket module first wall has been performed, maximizing the heat flux it may safely undergo. • Attention has been focused on the FW flat concept endowed with square cooling channels. • A theoretical-computational approach based on the finite element method (FEM) has been followed, adopting a qualified commercial FEM code. • Four optimized FW configurations have been found to safely withstand a heat flux up to 2 MW/m"2 fulfilling all the rules prescribed by safety codes. - Abstract: Within the framework of EUROfusion R&D activities a research campaign has been carried out at the University of Palermo in order to investigate the thermo-mechanical performances of the DEMO water-cooled lithium lead (WCLL) breeding blanket first wall (FW). The research campaign has been mainly focused on the optimization of the FW geometric configuration in order to maximize the heat flux it may safely withstand fulfilling all the thermal, hydraulic and mechanical requirements foreseen by safety codes. Attention has been focused on the FW flat concept endowed with square cooling channels and the potential influence of its four main geometrical parameters on its thermo-mechanical performances has been assessed performing a parametric analysis by means of a qualified commercial finite element method code. A set of 5929 different FW geometric configurations has been considered and the thermal performances of each one of them have been numerically assessed in case it undergoes 26 different values of heat flux on its plasma-facing surface. The resulting 154154 thermal analyses have allowed to select those cases fulfilling the adopted thermal-hydraulic requirements, whose thermo-mechanical performances have been numerically assessed under both normal operation and over-pressurization steady state loading scenarios to check whether they met the mechanical requirements prescribed by the pertaining SDC-IC safety rules. Four

  3. Analysis of the thermo-mechanical behaviour of the DEMO Water-Cooled Lithium Lead breeding blanket module under normal operation steady state conditions

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A.; Arena, P. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Aubert, J. [CEA Saclay, DEN/DANS/DM2S/SEMT, 91191 Gif sur Yvette Cedex (France); Bongiovì, G. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Chiovaro, P., E-mail: pierluigi.chiovaro@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Giammusso, R. [ENEA – C.R. Brasimone, 40032 Camugnano (Italy); Li Puma, A. [CEA Saclay, DEN/DANS/DM2S/SEMT, 91191 Gif sur Yvette Cedex (France); Tincani, A. [ENEA – C.R. Brasimone, 40032 Camugnano (Italy)

    2015-10-15

    Highlights: • A DEMO WCLL blanket module thermo-mechanical behaviour has been investigated. • Two models of the WCLL blanket module have been set-up adopting a code based on FEM. • The water flow domain in the module has been considered. • A set of uncoupled steady state thermo-mechanical analyses has been carried out. • Critical temperature is not overcome. Safety verifications are generally satisfied. - Abstract: Within the framework of DEMO R&D activities, a research cooperation has been launched between ENEA, the University of Palermo and CEA to investigate the thermo-mechanical behaviour of the outboard equatorial module of the DEMO1 Water-Cooled Lithium Lead (WCLL) blanket under normal operation steady state scenario. The research campaign has been carried out following a theoretical–computational approach based on the Finite Element Method (FEM) and adopting a qualified commercial FEM code. In particular, two different 3D FEM models (Model 1 and Model 2), reproducing respectively the central and the lateral poloidal–radial slices of the WCLL blanket module, have been set up. A particular attention has been paid to the modelling of water flow domain, within both the segment box channels and the breeder zone tubes, to simulate realistically the coolant-box thermal coupling. Results obtained are herewith reported and critically discussed.

  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. Tritium breeding measurements in a lithium blanket module with Pb/Be multipliers at the LOTUS facility

    International Nuclear Information System (INIS)

    Azam, S.; Kumar, A.

    1987-01-01

    The lithium blanket module (LBM) was lent for a fixed duration in 1985 to Ecole Polytechnique Federale de Lausanne under an agreement with the Electric Power Research Institute and Princeton Plasma Physics Laboratory. The first tritium breeding measurements in the central rod of the LBM and their analysis have been reported previously. Some time ago, we carried out additional experiments wherein the Li 2 O sample disk, each having a theoretical density of ∼85% and dimensions of 17.8-mm diam x 0.9-mm thickness, were placed in four removable rods. In addition to the central rod, the other rods were at ∼6-, 18-, and 39-cm radial distances from the axis of the central one. The sample disks wee kept at every 3 cm inside each of these rods up to a length of 30 cm in the Li 2 O part of the LBM. The choice of the off-axis rods resulted from our interest in investigating the effect of room return on tritium breeding in the LBM. We chose two of the leading neutron multipliers: (a) a 5-cm-thick (∼100- x 110-cm) lead slab and (b) a 6-cm-thick (∼66- x 66-cm) beryllium slab. The experimental assembly, consisting of the multiplier followed by the LBM, was kept at 10 cm from the generator. A packet of three foils, zirconium, indium, and aluminum, was placed at the center of the flat face of the generator to monitor the source intensity during the 10-h operation for the experiments with each multiplier. The source intensity is deduced to be ∼1.9 x 10 12 n/s for both the experiments. 5 refs., 3 figs

  6. The TBM-CA configuration management approach for the ITER test blanket module - application to the HCLL TBS

    International Nuclear Information System (INIS)

    Jourd'Heuil, L.; Panayotov, D.; Salavy, J.-F.; Storto, C.; Colombo, M.; Sardain, P.

    2011-01-01

    The European Test Blanket Modules (EU-TBM) are first prototypes of a fusion reactor breeding blanket. They will be tested in dedicated equatorial ports n o 16 of ITER. Technical developments are performed by a Consortium of European Associates (TBM-CA) and supported within the framework of F4E agency. Designing a complex nuclear system like TBM for ITER necessitates an organizational structure inside the consortium to manage in permanence the coherence between requirements (F4E technical and management specifications) and the TBM development through their life time. At the present stage, evolutionary nature of the design from the different teams is important. Highest priority is assigned to the Management support and Design Integration Team (MDIT) to perform an efficient control of the Configuration Management (CM). The TBM-CA CM comprises 4 main processes: a) identifying configuration of a product characteristics, including its interfaces (Configuration identification), b) controlling the evolution from agreed baseline (Configuration Control), c) creating the knowledge database in order to manage the information all along the lifecycle of the items (Configuration status accounting) and d) verifying the current configuration status of the items (Audits). CM is then a powerful tool to link the requirements for engineering, safety, quality assurance and test and acceptance activities. The application of the CM approach is illustrated through the case of TBM-HCLL (Helium Cooled Lithium Lead). The result shows that the proposed methodology and tools are suitable and provide quality solution for the items with a complex configuration such as TBM HCLL.

  7. Water testing of the FMIT lithium target

    International Nuclear Information System (INIS)

    Hassberger, J.A.; Ingham, J.G.

    1981-11-01

    Results of water tests of the lithium target design for the Fusion Materials Irradiation Test (FMIT) Facility demonstrate hydraulic features essential for acceptable target operation and confirm predictions of the target performance. This high speed, free surface, curved-wall jet has been shown to generate a stable surface shape and to provide the high velocities and pressures within the fluid needed to remove the 3.5 MW of power generated within the jet during FMIT operation. Measurements of the jet performance are found to fall within limits bounded by one- and two-dimensional potential flow predictions. This agreement between measured and predicted performance provides for a significant level of confidence in the ability of the FMIT lithium target to meet its design and functional objectives

  8. Lithium test module on ITER: engineering design of the tritium recovery system

    International Nuclear Information System (INIS)

    Finn, P.A.

    1988-01-01

    The design presented is an overview of the tritium recovery system for a lithium module on an ITER type reactor. The design of a tritium recovery system for larger blanket units, sectors, etc. could use the information developed in this report. A goal of this design was to ensure that a reliable, integrated performance of the tritium recovery system could be demonstrated. An equally important goal was to measure and account for the tritium in the liquid lithium blanket module and its recovery system in order to validate the operation of the blanket module

  9. 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

  10. 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)

  11. Water-cooled Pb-17Li test blanket module for ITER: impact of the structural material grade on the neutronic responses

    Energy Technology Data Exchange (ETDEWEB)

    Vella, G.; Aiello, G.; Oliveri, E. [Palermo Univ. (Italy). Dipt. di Ingegneria Nucl.; Fuetterer, M.A.; Giancarli, L. [CEA - Saclay, DRN/DMT/SERMA, Gif-sur-Yvette (France); Tavassoli, F. [CEA - Saclay, CEREM, Gif-sur-Yvette (France)

    1998-10-01

    The water-cooled lithium lead (WCLL) DEMO blanket is one of the two EU lines to be further developed with the aim of manufacturing by 2010 a test blanket module for ITER (TBM). In this paper results of a 3D-Monte Carlo neutronic analysis of the TBM design are reported. A fully 3D heterogeneous model of the WCLL-TBM has been inserted into an existing ITER model accounting for a proper D-T neutron source. The structural material assumed for the calculations was martensitic 9% Cr steel code named Z 10 CDV Nb 9-1. Results have been compared with those obtained using MANET. The main nuclear responses of the TBM have been determined, such as detailed power deposition density, material damage through DPA and He and H gas production rate, radial distribution of tritium production rate and total tritium production in the module. The impact of using natural lithium on the TBM system operation has also been evaluated. (orig.) 13 refs.

  12. Water-cooled, fire boom blanket, test and evaluation for system prototype development

    International Nuclear Information System (INIS)

    Stahovec, J. G.; Urban, R. W.

    1999-01-01

    Initial development of actively cooled fire booms indicated that water-cooled barriers could withstand direct oil fire for several hours with little damage if cooling water were continuously supplied. Despite these early promising developments, it was realized that to build reliable full-scale system for Navy host salvage booms would require several development tests and lengthy evaluations. In this experiment several types of water-cooled fire blankets were tested at the Oil and Hazardous Materials Simulated Test Tank (OHMSETT). After the burn test the blankets were inspected for damage and additional tests were conducted to determine handling characteristics for deployment, recovery, cleaning and maintenance. Test results showed that water-cooled fire boom blankets can be used on conventional offshore oil containment booms to extend their use for controlling large floating-oil marine fires. Results also demonstrated the importance of using thermoset rubber coated fabrics in the host boom to maintain sufficient reserve seam strength at elevated temperatures. The suitability of passively cooled covers should be investigated to protect equipment and boom from indirect fire exposure. 1 ref., 2 tabs., 8 figs

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

    Science.gov (United States)

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

    2006-02-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  16. Qualification Test for Korean Mockups of ITER Blanket First Wall

    International Nuclear Information System (INIS)

    Kim, S. K.; Lee, D. W.; Bae, Y. D.; Hong, B. G.; Jung, H. K.; Jung, Y. I.; Park, J. Y.; Jeong, Y. H.; Choi, B. K.; Kim, B. Y.

    2009-01-01

    ITER First Wall (FW) includes the beryllium armor tiles joined to CuCrZr heat sink with stainless steel cooling tubes. This first wall panels are one of the critical components in the ITER machine with the surface heat flux of 0.5 MW/m 2 or above. So qualification program needs to be performed with the goal to qualify the joining technologies required for the ITER First Wall. Based on the results of tests, the acceptance of the developed joining technologies will be established. The results of this qualification test will affect the final selection of the manufacturers for the ITER First Wall

  17. 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)

  18. Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

    International Nuclear Information System (INIS)

    Tanigawa, H.; Hirose, T.; Shiba, K.; Kasada, R.; Wakai, E.; Serizawa, H.; Kawahito, Y.; Jitsukawa, S.; Kimura, A.; Kohno, Y.; Kohyama, A.; Katayama, S.; Mori, H.; Nishimoto, K.; Klueh, R.L.; Sokolov, M.A.; Stoller, R.E.; Zinkle, S.J.

    2008-01-01

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on high-temperature properties and weldability. Extensive irradiation studies have conducted on F82H, and it has the most extensive available database of irradiated and unirradiated properties of all RAFMs. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of an ITER test blanket module (TBM) suggested from the recent research achievements in Japan. This work clarified that the primary issues with F82H involve welding techniques and the mechanical properties of weld joints. This is the result of the distinctive nature of the joint caused by the phase transformation that occurs in the weld joint during cooling, and its impact on the design of a TBM will be discussed

  19. Measurement and analysis of neutron flux spectra in a neutronics mock-up of the HCLL test blanket module

    International Nuclear Information System (INIS)

    Klix, A.; Batistoni, P.; Boettger, R.; Lebrun-Grandie, D.; Fischer, U.; Henniger, J.; Leichtle, D.; Villari, R.

    2010-01-01

    Fast neutron and gamma-ray flux spectra and time-of-arrival spectra of slow neutrons have been measured in a neutronics mock-up of the European Helium-Cooled Lithium-Lead Test Blanket Module with the aim to validate nuclear cross-section data. The mock-up was irradiated with fusion peak neutrons from the DT neutron generator of the Technical University of Dresden. A well characterized cylindrical NE-213 scintillator was inserted into two positions in the LiPb/EUROFER assembly. Pulse height spectra from neutrons and gamma-rays were recorded from the NE-213 output. The spectra were then unfolded with experimentally obtained response matrices of the NE-213 detector. Time-of-arrival spectra of slow neutrons were measured with a 3 He counter placed in the mock-up, and the neutron generator was operated in pulsed mode. Monte Carlo calculations using the MCNP code and nuclear cross-section data from the JEFF-3.1.1 and FENDL-2.1 libraries were performed and the results are compared with the experimental results. A good agreement of measurement and calculation was found with some deviations in certain energy intervals.

  20. Mock-up test on key components of ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Matsumoto, Yasuhiro; Taguchi, Koh; Kozaka, Hiroshi; Shibanuma, Kiyoshi; Tesini, Alessandro

    2009-01-01

    The maintenance operation of the ITER in-vessel component, such as a blanket and divertor, must be executed by the remote equipment because of the high gamma-ray environment. During the Engineering Design Activity (EDA), the Japan Atomic Energy Agency (then called as Japan Atomic Energy Research Institute) had been fabricated the prototype of the vehicle manipulator system for the blanket remote handling and confirmed feasibility of this system including automatic positioning of the blanket and rail deployment procedure of the articulated rail. The ITER agreement, which entered into force in the last year, formally decided that Japan will procure the blanket remote handling system and the JAEA, as the Japanese Domestic Agency, is continuing several R and Ds so that the system can be procured smoothly. The residual key issues after the EDA are rail connection and cable handling. The mock-ups of the rail connection mechanism and the cable handling system were fabricated from the last year and installed at the JAEA Naka Site in this March. The former was composed of the rail connecting mechanism, two rail segments and their handling systems. The latter one utilized a slip ring, which implemented 80 lines for power and 208 lines for signal, because there is an electrical contact between the rotating spool and the fixed base. The basic function of these systems was confirmed through the mock-up test. The rail connection mechanism, for example, could accept misalignment of 1.5-2 mm at least. The future test plan is also mentioned in the paper.

  1. Applications of the aqueous self-cooled blanket (ASCB) concept to the Next European Torus (NET)

    International Nuclear Information System (INIS)

    Embrechts, M.J.; Bogaerts, W.; Cardella, A.; Chazalon, M.; Danner, W.; Dinner, P.; Libin, B.

    1987-01-01

    The Aqueous Self-Cooled Blanket Concept (ASCB) leads to a low-technology blanket design that relies on just structural material and coolant with small amounts of lithium compound dissolved in the coolant to provide for tritium production. The application of the ASCB concept in NET is being considered as a driver blanket that would operate at low temperature and low pressure and provide a reliable environment for machine operation during the technology phase. Shielding and tritium production are the primary objectives for such a low-technology blanket. Net tritium breeding is not a design requirement per se for a driver blanket for NET. A DEMO relevant ASCB based blanket test module with (local) tritium self-sufficiency and energy recovery as primary objectives might also be tested in NET if future developments confirm their viability

  2. Dynamic test of the ITER blanket key and ceramic insulated pad

    International Nuclear Information System (INIS)

    Khomyakov, S.; Sysoev, G.; Strebkov, Yu.; Kucherov, A.; Ioki, K.

    2010-01-01

    The dynamic testing of the blanket module's key integrated into ITER vacuum vessel portion has been performed in 2008 to investigate its capability to react the electro-magnetic (EM) loads. The preliminary analysis showed the large dynamic amplification factor (DAF) of the reactions because of technological gaps between the blanket module and key. Shock load may yield the bronze pads, which protect the blanket electrical insulation from damage. However the dynamic analysis of such particularly non-linear system needs an experimental ground and confirmation. Toward this end, as well as demonstration of the key reliability, the special test facility has been made, and the full-scale mock-up of the inboard intermodular key was tested. So as not to scale non-linear dynamic parameters, 1-ton mass was built on the single flexible support. The key was welded in a 60-mm thick steel plate modeled with a fragment of the VV. The different gaps were set in between the bronze pad of the key and the mass shock worker. This system (supplemented with some additional constraints) has natural oscillations like as the 4-ton module built on four flexible supports. Thus the most critical radial torque might be modeled with a straight force. The objectives of the test were as follows: dynamic response, DAF and damping factor determination; measurement of the strain oscillations in the key's base and in the weld seam; comparison of the measured data with computation results. The paper will present the analytical grounds of the testing conditions, test facility description, analytical adaptation of the facility, experimental results, its comparison with analysis and discussion, and guidelines for the next experimental phase.

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

    International Nuclear Information System (INIS)

    Conrad, R.; Debarberis, L.

    1991-01-01

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

  4. 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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  6. Numeric implementation of a nucleation, growth and transport model for helium bubbles in lead-lithium HCLL breeding blanket channels: Theory and code development

    Energy Technology Data Exchange (ETDEWEB)

    Batet, L., E-mail: lluis.batet@upc.edu [Technical University of Catalonia (UPC), Energy and Radiation Studies Research Group (GREENER), Technology for Fusion T4F, Barcelona (Spain); UPC, Department of Physics and Nuclear Engineering (DFEN), ETSEIB, Av. Diagonal 647, 08028 Barcelona (Spain); Fradera, J. [Technical University of Catalonia (UPC), Energy and Radiation Studies Research Group (GREENER), Technology for Fusion T4F, Barcelona (Spain); UPC, Department of Physics and Nuclear Engineering (DFEN), ETSEIB, Av. Diagonal 647, 08028 Barcelona (Spain); Valls, E. Mas de les [Technical University of Catalonia (UPC), Energy and Radiation Studies Research Group (GREENER), Technology for Fusion T4F, Barcelona (Spain); UPC, Department of Heat Engines (DMMT), ETSEIB, Av. Diagonal 647, 08028 Barcelona (Spain); Sedano, L.A. [EURATOM-CIEMAT Association, Fusion Technology Division, Av. Complutense 22, 28040 Madrid (Spain)

    2011-06-15

    Large helium (He) production rates in liquid metal breeding blankets of a DT fusion reactor might have a significant influence in the system design. Low He solubility together with high local concentrations may create the conditions for He cavitation, which would have an impact in the components performance. The paper states that such a possibility is not remote in a helium cooled lithium-lead breeding blanket design. A model based on the Classical Nucleation Theory (CNT) has been developed and implemented in order to have a specific tool able to simulate HCLL systems and identify the key parameters and sensitivities. The nucleation and growth model has been implemented in the open source CFD code OpenFOAM so that transport of dissolved atomic He and nucleated He bubbles can be simulated. At the current level of development it is assumed that void fraction is small enough not to affect either the hydrodynamics or the properties of the liquid metal; thus, bubbles can be represented by means of a passive scalar. He growth and transport has been implemented using the mean radius approach in order to save computational time. Limitations and capabilities of the model are shown by means of zero-dimensional simulation and sensitivity analysis under HCLL breeding unit conditions.

  7. Electrical, thermal and abusive tests on lithium thionyl chloride cells

    Science.gov (United States)

    Frank, H. A.

    1980-04-01

    Electrical characterizations, thermal characterizations, and outer limits tests of lithium thionyl chloride cells are discussed. Graphs of energy density vs power density and heat rate vs time are presented along with results of forced reversal and high rate discharge tests.

  8. 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

  9. 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

  10. Electromagnetic analysis of the Korean helium cooled ceramic reflector test blanket module set

    International Nuclear Information System (INIS)

    Lee, Youngmin; Ku, Duck Young; Lee, Dong Won; Ahn, Mu-Young; Park, Yi-Hyun; Cho, Seungyon

    2016-01-01

    Korean helium cooled ceramic reflector (HCCR) test blanket module set (TBM-set) will be installed at equatorial port #18 of Vacuum Vessel in ITER in order to test the breeding blanket performance for forthcoming fusion power plant. Since ITER tokamak has a set of electromagnetic coils (Central Solenoid, Poloidal Field and Toroidal Field coil set) around Vacuum Vessel, the HCCR TBM-set, the TBM and associated shield, is greatly influenced by magnetic field generated by these coils. In the case of fast transient electromagnetic events such as major disruption, vertical displacement event or magnet fast discharge, magnetic field and induced eddy current results in huge electromagnetic load, known as Lorentz load, on the HCCR TBM-set. In addition, the TBM-set experiences electromagnetic load due to magnetization of the structural material not only during the fast transient events but also during normal operation since the HCCR TBM adopts Reduced Activation Ferritic Martensitic (RAFM) steel as a structural material. This is known as Maxwell load which includes Lorentz load as well as load due to magnetization of structure material. This paper presents electromagnetic analysis results for the HCCR TBM-set. For analysis, a 20° sector finite model was constructed considering ITER configuration such as Vacuum Vessel, ITER shield blankets, Central Solenoid, Poloidal Field, Toroidal Field coil set as well as the HCCR TBM-set. Three major disruptions (operational event, likely event and highly unlikely event) were selected for analysis based on the load specifications. ANSYS-EMAG was used as a calculation tool. The results of EM analysis will be used as input data for the structural analysis.

  11. Electromagnetic analysis of the Korean helium cooled ceramic reflector test blanket module set

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

    Korean helium cooled ceramic reflector (HCCR) test blanket module set (TBM-set) will be installed at equatorial port #18 of Vacuum Vessel in ITER in order to test the breeding blanket performance for forthcoming fusion power plant. Since ITER tokamak has a set of electromagnetic coils (Central Solenoid, Poloidal Field and Toroidal Field coil set) around Vacuum Vessel, the HCCR TBM-set, the TBM and associated shield, is greatly influenced by magnetic field generated by these coils. In the case of fast transient electromagnetic events such as major disruption, vertical displacement event or magnet fast discharge, magnetic field and induced eddy current results in huge electromagnetic load, known as Lorentz load, on the HCCR TBM-set. In addition, the TBM-set experiences electromagnetic load due to magnetization of the structural material not only during the fast transient events but also during normal operation since the HCCR TBM adopts Reduced Activation Ferritic Martensitic (RAFM) steel as a structural material. This is known as Maxwell load which includes Lorentz load as well as load due to magnetization of structure material. This paper presents electromagnetic analysis results for the HCCR TBM-set. For analysis, a 20° sector finite model was constructed considering ITER configuration such as Vacuum Vessel, ITER shield blankets, Central Solenoid, Poloidal Field, Toroidal Field coil set as well as the HCCR TBM-set. Three major disruptions (operational event, likely event and highly unlikely event) were selected for analysis based on the load specifications. ANSYS-EMAG was used as a calculation tool. The results of EM analysis will be used as input data for the structural analysis.

  12. Remote handling of the blanket segments: Testing of 1/3 scale mock-ups on the ROBERTINO facility

    International Nuclear Information System (INIS)

    Maisonnier, D.; Amelotti, F.; Chiasera, A.

    1994-01-01

    The remotized replacement of the blanket segments inside the Vacuum Vessel of a fusion reactor is one of the critical tasks for reactor components design, operational procedures, and safety. This open-quotes hostile environmentclose quotes task must be accomplished by a specific Blanket Handling Device, with a grasping device acting as open-quotes end-effectorclose quotes, because of intervention complexity, of components dimensions and weights, and of consequences of possible accidents during the blanket segments handling operations. Therefore, specific support experimental studies in this field appear to be necessary in order to: select appropriate blanket handling devices and procedures; assess the design of all components involved in the handling operations; perform checks in all field related to the robotized handling control (kinematics and dynamics of the grasping device trajectory planning and motion control, sensing and intelligence of the blanket handling devices, etc.); improve reliability and safety for the replacement sequences; give a realistic estimation of the time duration of the replacement duration. During the test phase, handling operations were carried out on the blanket mock-ups by means of different gripping devices. The operations were driven in the control room by means of the Motion command computer and the real time sensing data display allowed operations' control. The results were analyzed by charting the sensors' data

  13. EBR-II blanket fuel leaching test using simulated J-13 well water

    International Nuclear Information System (INIS)

    Fonnesbeck, J. E.

    1999-01-01

    This paper discusses the results of a pulsed-flow leaching test using simulated J-13 well water leachant. This test was performed on three blanket fuel segments from the ANL-W EBR-II nuclear reactor which were originally made up of depleted uranium (DU). This experiment was designed to mimic conditions which would exist if, upon disposal of this material in a geological repository, it came in direct contact with groundwater. These segments were contained in pressure vessels and maintained at a constant temperature of 90 C. Weekly aliquots of leachate were taken from the three vessels and replaced with an equal volume of fresh leachant. These weekly aliquots were analyzed for both 90 Sr and 137 Cs. The results of the pulsed-flow leach test showed the formation of uranium oxide (UO 2 ) and uranium hydride (UH 3 ) particulate with rapid release of the 137 Cs and 90 Sr to the leachant. On the fifth week of sampling, one of the vessels became over pressurized and vented gas when opened. The most reasonable explanation for the presence of gas in this vessel is that the unoxidized uranium metal in the blanket segment could have reacted with the surrounding water leachant to form hydrogen. However, an investigation is currently being undertaken to both qualify and quantify H 2 formation during uranium spent nuclear fuel corrosion in water

  14. 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.

  15. Progress in the integration of Test Blanket Systems in ITER equatorial port cells and in the interfaces definition

    Energy Technology Data Exchange (ETDEWEB)

    Pascal, R., E-mail: romain.pascal@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Beloglazov, S.; Bonagiri, S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Commin, L. [CEA, IRFM, Cadarache (France); Cortes, P.; Giancarli, L.M.; Gliss, C.; Iseli, M.; Lanza, R.; Levesy, B.; Martins, J.-P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Neviere, J.-C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Patisson, L.; Plutino, D.; Shu, W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Swami, H.L. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer The design integration of two test blanket systems in ITER port cell is addressed. Black-Right-Pointing-Pointer Definition of interfaces of TBSs with building and other ITER systems is done. Black-Right-Pointing-Pointer Designs of pipe forest, bioshield plug and ancillary equipment unit are described. Black-Right-Pointing-Pointer The maintenance of the two test blanket systems in ITER port cell is considered. Black-Right-Pointing-Pointer The management of the heat and tritium releases in the TBM port cell is described. - Abstract: In the framework of the TBM Program, three ITER vacuum vessel equatorial ports (no. 16, no. 18 and no. 02) have been allocated for the testing of up to six mock-ups of six different DEMO tritium breeding blankets. Each one is called a Test Blanket System (TBS). A TBS consists mainly of the Test Blanket Module (TBM), the in-vessel component facing the plasma, and several ancillary systems, in particular the cooling system and the tritium extraction system. Each port accommodates two TBMs and therefore the two TBSs have to share the corresponding port cell. This paper deals with the design integration aspects of the two TBSs in each port cell performed at ITER Organization (IO) with the corresponding definition of interfaces with other ITER systems. The performed activities have raised several issues that are discussed in the paper and for which design solutions are proposed.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  20. 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).

  1. 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).

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

    International Nuclear Information System (INIS)

    Lux, M.

    2007-01-01

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

  3. 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

  4. Analysis of ER string test thermally instrumented interconnect 80-K MLI blanket

    International Nuclear Information System (INIS)

    Daly, E.; Pletzer, R.

    1992-04-01

    An 80-K Multi Layer Insulation (MLI) blanket in the interconnect region between magnets DD0019 and DD0027 in the Fermi National Accelerator Laboratory (FNAL) ER string was instrumented with temperature sensors to obtain the steady-state temperature gradient through the blanket after string cooldown. A thermal model of the 80-K blanket assembly was constructed to analyze the steady-state temperature gradient data. Estimates of the heat flux through the 80-K MLI blanket assembly and predicted temperature gradients were calculated. The thermal behavior of the heavy polyethylene terapthalate (PET) cover layers separating the shield and inner blanket and inner and outer blankets was derived empirically from the data. The results of the analysis predict a heat flux of 0.363--0.453 W/m 2 based on the 11 sets of data. These flux values are 33--46% below the 80-K MLI blanket heat leak budget of 0.676 W/m 2 . The effective thermal resistance of the two heavy PET cover layers between the shield and inner blanket was found to be 2.1 times that of a single PET spacer layer, and the effective resistance of the two heavy PET cover layers between the inner blanket and outer blanket was found to be 7 times that of a single PET spacer layer. Based on these results, the 80-K MLI blanket assembly appears to be performing more than adequately to meet the 80-K static IR heat leak budget. However, these results should not be construed as a verification of the 80-K static IR heat leak, since no actual heat leak was measured. The results have been used to improve the empirically based model data in the 80-K MLI blanket thermal model, which has previously not included the effects of heavy PET cover layers on 80-K MLI blanket thermal performance

  5. Tritium and heat management in ITER Test Blanket Systems port cell for maintenance operations

    Energy Technology Data Exchange (ETDEWEB)

    Giancarli, L.M., E-mail: luciano.giancarli@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Cortes, P.; Iseli, M.; Lepetit, L.; Levesy, B. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Livingston, D. [Frazer-Nash Consultancy Ltd., Stonebridge House, Dorking Business Park, Dorking, Surrey RH4 1HJ (United Kingdom); Nevière, J.C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Pascal, R. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ricapito, I. [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Shu, W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Wyse, S. [Frazer-Nash Consultancy Ltd., Stonebridge House, Dorking Business Park, Dorking, Surrey RH4 1HJ (United Kingdom)

    2014-10-15

    Highlights: •The ITER TBM Program is one of the ITER missions. •We model a TBM port cell with CFD to optimize the design choices. •The heat and tritium releases management in TBM port cells has been optimized. •It is possible to reduce the T-concentration below one DAC in TBM port cells. •The TBM port cells can have human access within 12 h after shutdown. -- Abstract: Three ITER equatorial port cells are dedicated to the assessment of six different designs of breeding blankets, known as Test Blanket Modules (TBMs). Several high temperature components and pipework will be present in each TBM port cell and will release a significant quantity of heat that has to be extracted in order to avoid the ambient air and concrete wall temperatures to exceed allowable limits. Moreover, from these components and pipes, a fraction of the contained tritium permeates and/or leaks into the port cell. This paper describes the optimization of the heat extraction management during operation, and the tritium concentration control required for entry into the port cell to proceed with the required maintenance operations after the plasma shutdown.

  6. 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.

  7. 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.

  8. RELAP/SCDAPSIM/MOD4.0 modification for transient accident scenario of Test Blanket Modules in ITER involving helium flows into heavy liquid metal

    Energy Technology Data Exchange (ETDEWEB)

    Freixa, J.; Pérez, M.; Mas de les Valls, E.; Batet, L.; Sandeep, T.; Chaudhari, V.; Reventós, F.

    2015-07-01

    The Institute for Plasma Research (IPR), India, is currently involved in the design and development of its Test Blanket Module (TBM) for testing in ITER (International Thermo nuclear Experimental Reactor). The Indian TBM concept is a Lead-Lithium cooled Ceramic Breeder (LLCB), which utilizes lead-lithium eutectic alloy (LLE) as tritium breeder, neutron multiplier and coolant. The first wall facing the plasma is cooled by helium gas. In preparation of the regulatory safety files of ITER-TBM, a number of off-normal event sequences have been postulated. Thermal hydraulic safety analyses of the TBM system will be carried out with the system code RELAP/SCDAPSIM/MOD4.0 which was initially designed to predict the behavior of light water reactor systems during normal and accidental conditions. In order to analyze some of the postulated off-normal events, there is the need to simulate the mixing of Helium and Lead-Lithium fluids. The Technical University of Catalonia is cooperating with IPR to implement the necessary changes in the code to allow for the mixing of helium and liquid metal. In the present study, the RELAP/SCDAPSIM/MOD4 two-phase flow 6-equations structure has been modified to allow for the mixture of LLE in the liquid phase with dry Helium in the gas phase. Practically obtaining a two-fluid 6-equation model where each fluid is simulated with a set of energy, mass and momentum balance equations. A preliminary flow regime map for LLE and helium flow has been developed on the basis of numerical simulations with the OpenFOAM CFD toolkit. The new code modifications have been verified for vertical and horizontal configurations. (Author)

  9. 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

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

    International Nuclear Information System (INIS)

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

    1991-12-01

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

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

    International Nuclear Information System (INIS)

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

    1991-12-01

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

  12. EBR-II blanket fuel leaching test using simulated J-13 well water.

    Energy Technology Data Exchange (ETDEWEB)

    Fonnesbeck, J. E.

    1998-05-15

    A pulsed-flow leaching test is being conducted using three EBR-II blanket fuel segments. These samples are immersed in simulated J-13 well water. The samples are kept at a constant temperature of 90 C. Leachate is exchanged weekly and analyzed for various nuclides which are of interest from a mobility and longevity point of view. Our primary interest is in the longer-lived species such as {sup 99}Tc, {sup 237}Np, and {sup 241}Am. In addition, the behavior of U, Pu, {sup 90}Sr, and {sup 137}Cs are being analyzed. During the course of this experiment, an interesting observation has been made involving one of the samples which could indicate the possible rapid ''anoxic'' oxidation of uranium metal to UO{sub 2}.

  13. Development of Reduced Activation Ferritic-Martensitic Steels and fabrication technologies for Indian test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Raj, Baldev [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Jayakumar, T., E-mail: tjk@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2011-10-01

    For the development of Reduced Activation Ferritic-Martensitic Steel (RAFMS), for the Indian Test Blanket Module for ITER, a 3-phase programme has been adopted. The first phase consists of melting and detailed characterization of a laboratory scale heat conforming to Eurofer 97 composition, to demonstrate the capability of the Indian industry for producing fusion grade steel. In the second phase which is currently in progress, the chemical composition will be optimized with respect to tungsten and tantalum for better combination of mechanical properties. Characterization of the optimized commercial scale India-specific RAFM steel will be carried out in the third phase. The first phase of the programme has been successfully completed and the tensile, impact and creep properties are comparable with Eurofer 97. Laser and electron beam welding parameters have been optimized and welding consumables were developed for Narrow Gap - Gas Tungsten Arc welding and for laser-hybrid welding.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-08-15

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

  15. 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

  16. Water-cooled lithium-lead box-shaped blanket concept for Demo: thermo-mechanical optimization and manufacturing sequence proposal

    International Nuclear Information System (INIS)

    Baraer, L.; Dinot, N.; Giancarli, L.; Proust, E.; Salavy, J.F.; Severi, Y.; Quintric-Bossy, J.

    1992-01-01

    The development of the water-cooled lithium-lead box-shaped blanket concept for DEMO has now reached the stage of thermo-mechanical optimization. In the previous design phases the preliminary dimensioning of the cooling circuit has permitted to define the water proportions required in the breeder region and to demonstrate, after a minimization of steel proportion and thicknesses, that this concept could reach tritium breeding self-sufficiency. In the present analysis the location of the coolant pipes has been optimized for the whole equatorial plane cross-section of both inboard and outboard segments in order to maintain the maximum Pb-17Li/steel interface temperature below 480 deg C and to minimize the thermal gradients along the steel structures. The consequent thermo-mechanical analysis has shown that the thermal stresses always remain below the allowable limits. Segment fabricability and removal are the next design issues to be analyzed. Within this strategy, a first manufactury sequence for the outboard segment is proposed

  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. Experimental results and validation of a method to reconstruct forces on the ITER test blanket modules

    International Nuclear Information System (INIS)

    Zeile, Christian; Maione, Ivan A.

    2015-01-01

    Highlights: • An in operation force measurement system for the ITER EU HCPB TBM has been developed. • The force reconstruction methods are based on strain measurements on the attachment system. • An experimental setup and a corresponding mock-up have been built. • A set of test cases representing ITER relevant excitations has been used for validation. • The influence of modeling errors on the force reconstruction has been investigated. - Abstract: In order to reconstruct forces on the test blanket modules in ITER, two force reconstruction methods, the augmented Kalman filter and a model predictive controller, have been selected and developed to estimate the forces based on strain measurements on the attachment system. A dedicated experimental setup with a corresponding mock-up has been designed and built to validate these methods. A set of test cases has been defined to represent possible excitation of the system. It has been shown that the errors in the estimated forces mainly depend on the accuracy of the identified model used by the algorithms. Furthermore, it has been found that a minimum of 10 strain gauges is necessary to allow for a low error in the reconstructed forces.

  19. Biaxial Testing of 2195 Aluminum Lithium Alloy Using Cruciform Specimens

    Science.gov (United States)

    Johnston, W. M.; Pollock, W. D.; Dawicke, D. S.; Wagner, John A. (Technical Monitor)

    2002-01-01

    A cruciform biaxial test specimen was used to test the effect of biaxial load on the yield of aluminum-lithium alloy 2195. Fifteen cruciform specimens were tested from 2 thicknesses of 2195-T8 plate, 0.45 in. and 1.75 in. These results were compared to the results from uniaxial tensile tests of the same alloy, and cruciform biaxial tests of aluminum alloy 2219-T87.

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

    International Nuclear Information System (INIS)

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

    1991-10-01

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

  1. Lithium

    Science.gov (United States)

    Bradley, Dwight C.; Stillings, Lisa L.; Jaskula, Brian W.; Munk, LeeAnn; McCauley, Andrew D.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Lithium, the lightest of all metals, is used in air treatment, batteries, ceramics, glass, metallurgy, pharmaceuticals, and polymers. Rechargeable lithium-ion batteries are particularly important in efforts to reduce global warming because they make it possible to power cars and trucks from renewable sources of energy (for example, hydroelectric, solar, or wind) instead of by burning fossil fuels. Today, lithium is extracted from brines that are pumped from beneath arid sedimentary basins and extracted from granitic pegmatite ores. The leading producer of lithium from brine is Chile, and the leading producer of lithium from pegmatites is Australia. Other potential sources of lithium include clays, geothermal brines, oilfield brines, and zeolites. Worldwide resources of lithium are estimated to be more than 39 million metric tons, which is enough to meet projected demand to the year 2100. The United States is not a major producer at present but has significant lithium resources.

  2. Liquid metal cooled blanket concept for NET

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  3. Manufacturing and testing of full scale prototype for ITER blanket shield block

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sa-Woong, E-mail: swkim12@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Duck-Hoi; Jung, Hun-Chea [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Sung-Ki [WONIL Co., Ltd., Haman (Korea, Republic of); Kang, Sung-Chan [POSCO Specialty Steel Co., Ltd., Changwon (Korea, Republic of); Zhang, Fu; Kim, Byoung-Yoon [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ahn, Hee-Jae; Lee, Hyeon-Gon; Jung, Ki-Jung [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-04-15

    Highlights: • 316L(N)-IG forged steel was successfully fabricated and qualified. • Related R&D activities were implemented to resolve the fabrication issues. • SB #8 FSP was successfully manufactured with conventional fabrication techniques. • All of the validation tests were carried out and met the acceptance criteria. - Abstract: Based on the preliminary design of the ITER blanket shield block (SB) #8, the full scale prototype (FSP) has been manufactured and tested in accordance with pre-qualification program, and related R&D was performed to resolve the technical issues of fabrication. The objective of the SB pre-qualification program is to demonstrate the acceptable manufacturing quality by successfully passing the formal test program. 316L(N)-IG stainless steel forging blocks with 1.80L × 1.12W × 0.43t (m) were developed by using an electric arc furnace, and as a result, the material properties were satisfied with technical specification. In the course of applying conventional fabrication techniques such as cutting, milling, drilling and welding of the forged stainless steel block for the manufacturing of the SB #8 FSP, several technical problems have been addressed. And also, the hydraulic connector of cross-forged material re-melted by electro slag or vacuum arc requires the application of advanced joining techniques such as automatic bore TIG and friction welding. Many technical issues – drilling, welding, slitting, non-destructive test and so on – have been raised during manufacturing. Associated R&D including the computational simulation and coupon testing has been done in collaboration with relevant industries in order to resolve these engineering issues. This paper provides technical key issues and their possible resolutions addressed during the manufacture and formal test of the SB #8 FSP, and related R&D.

  4. Fabrication and testing of small scale mock-ups of ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Hatano, Toshihisa; Sato, Satoshi; Suzuki, Satoshi; Yokoyama, Kenji; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-12-01

    Small scale mock-ups of the primary first wall, the baffle first wall, the shield block and a partial model for the edge of the primary first wall module were designed and fabricated incorporating most of the key design features of the ITER shielding blanket. All mock-ups featured the DSCu heat sink, the built-in SS coolant tubes within the heat sink and the SS shield block. CFC tiles was used as the protection armor for the baffle first wall mock-up. The small scale shield block mock-up, integrated with the first wall, was designed to have a poloidal curvature specified in the ITER design. Fabrication routes of mock-ups were decided based on the single step solid HIP of DSCu/DSCu, DSCu/SS and SS/SS reflecting the results of previous joining techniques development and testing. For attaching the CFC tiles onto DSCu heat sink in the fabrication of the baffle first wall mock-up, a two-step brazing was tried. All mock-ups and the partial model were successfully fabricated with a satisfactory dimensional accuracy. The small scale primary first wall mock-up was thermo-mechanically tested under high heat fluxes of 5-7 MW/m{sup 2} for 2500 cycles in total. Satisfactory heat removal performance and integrity of the mock-up against cyclic high heat flux loads were confirmed by measurement during the tests and destructive examination after the tests. Similar high heat flux tests were also performed with the small scale baffle first wall mock-up under 5-10 MW/m{sup 2} for 4500 cycles in total resulting in sufficient heat removal capability and integrity confirmed by measurements during the tests. (author)

  5. Fabrication and testing of small scale mock-ups of ITER shielding blanket

    International Nuclear Information System (INIS)

    Hatano, Toshihisa; Sato, Satoshi; Suzuki, Satoshi; Yokoyama, Kenji; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki; Ohara, Yoshihiro

    1998-12-01

    Small scale mock-ups of the primary first wall, the baffle first wall, the shield block and a partial model for the edge of the primary first wall module were designed and fabricated incorporating most of the key design features of the ITER shielding blanket. All mock-ups featured the DSCu heat sink, the built-in SS coolant tubes within the heat sink and the SS shield block. CFC tiles was used as the protection armor for the baffle first wall mock-up. The small scale shield block mock-up, integrated with the first wall, was designed to have a poloidal curvature specified in the ITER design. Fabrication routes of mock-ups were decided based on the single step solid HIP of DSCu/DSCu, DSCu/SS and SS/SS reflecting the results of previous joining techniques development and testing. For attaching the CFC tiles onto DSCu heat sink in the fabrication of the baffle first wall mock-up, a two-step brazing was tried. All mock-ups and the partial model were successfully fabricated with a satisfactory dimensional accuracy. The small scale primary first wall mock-up was thermo-mechanically tested under high heat fluxes of 5-7 MW/m 2 for 2500 cycles in total. Satisfactory heat removal performance and integrity of the mock-up against cyclic high heat flux loads were confirmed by measurement during the tests and destructive examination after the tests. Similar high heat flux tests were also performed with the small scale baffle first wall mock-up under 5-10 MW/m 2 for 4500 cycles in total resulting in sufficient heat removal capability and integrity confirmed by measurements during the tests. (author)

  6. 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)

  7. The testing report of the development for the lithium grains and lithium rod automatic machine

    International Nuclear Information System (INIS)

    Qian Zongkui; Kong Xianghong; Huang Yong

    2008-06-01

    With the development of lithium industry, the lithium grains and lithium rod, as additive or catalyzer, having a big comparatively acreage and a strong activated feature, have a broad application. The lithium grains and lithium rod belong to the kind of final machining materials. The principle of the lithium grains and lithium rod that how to take shape through the procedures of extrusion, cutting, anti-conglutination, threshing and so on are analysed, A sort of lithium grains and lithium rod automatic machine is developed. (authors)

  8. 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

  9. 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)

  10. Analysis of the October 5, 1979 lithium spill and fire in the Lithium Processing Test Loop

    International Nuclear Information System (INIS)

    Maroni, V.A.; Beatty, R.A.; Brown, H.L.; Coleman, L.F.; Foose, R.M.; McPheeters, C.C.; Slawecki, M.; Smith, D.L.; Van Deventer, E.H.; Weston, J.R.

    1981-12-01

    On October 5, 1979, the Lithium Processing Test Loop (LPTL) developed a lithium leak in the electromagnetic (EM) pump channel, which damaged the pump, its surrounding support structure, and the underlying floor pan. A thorough analysis of the causes and consequences of the pump failure was conducted by personnel from CEN and several other ANL divisions. Metallurgical analyses of the elliptical pump channel and adjacent piping revealed that there was a significant buildup of iron-rich crystallites and other solid material in the region of the current-carrying bus bars (region of high magnetic field), which may have resulted in a flow restriction that contributed to the deterioration of the channel walls. The location of the failure was in a region of high residual stress (due to cold work produced during channel fabrication); this failure is typical of other cold work/stress-related failures encountered in components operated in forced-circulation lithium loops. Another important result was the isolation of crystals of a compound characterized as Li/sub x/CrN/sub y/. Compounds of this type are believed to be responsible for much of the Fe, Cr, and Ni mass transfer encountered in lithium loops constructed of stainless steel. The importance of nitrogen in the mass-transfer mechanism has long been suspected, but the existence of stable ternary Li-M-N compounds (M = Fe, Cr, Ni) had not previously been verified

  11. Fast Ion Effects During Test Blanket Module Simulation Experiments in DIII-D

    International Nuclear Information System (INIS)

    Kramer, G.J.; Budny, R.V.; Ellis, R.; Gorelenkova, M.; Heidbrink, W.W.; Kurki-Suonio, T.; Nazikian, R.; Salmi, A.; Schaffer, M.J.; Shinohara, K.; Snipes, J.A.; Spong, D.A.; Koskela, T.; Van Zeeland, M.A.

    2011-01-01

    Fast beam-ion losses were studied in DIII-D in the presence of a scaled mockup of two Test Blanket Modules (TBM) for ITER. Heating of the protective tiles on the front of the TBM surface was found when neutral beams were injected and the TBM fields were engaged. The fast-ion core confinement was not significantly affected. Different orbit-following codes predict the formation of a hot spot on the TBM surface arising from beam-ions deposited near the edge of the plasma. The codes are in good agreement with each other on the total power deposited at the hot spot predicting an increase in power with decreasing separation between the plasma edge and the TBM surface. A thermal analysis of the heat flow through the tiles shows that the simulated power can account for the measured tile temperature rise. The thermal analysis, however, is very sensitive to the details of the localization of the hot spot which is predicted to be different among the various codes.

  12. Activation and afterheat analyses for the HCPB test blanket module in ITER

    International Nuclear Information System (INIS)

    Pereslavtsev, P.; Fischer, U.

    2008-01-01

    To provide a sound data basis for the safety analyses of the HCPB TBM system in ITER, the afterheat and activity inventories were assessed making use of a code system that allows performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. A suitable MCNP model of a 20 deg. ITER torus sector with an integrated TBM of the HCPB PI (plant integration) type in the horizontal test blanket port was developed and adapted to the requirements for coupled 3D neutron transport and activation calculations. Two different irradiation scenarios were considered in the coupled 3D neutron transport and activation calculations. The first one is representative for the TBM irradiation in ITER with a total of 9000 neutron pulses over a 3 (calendar) years period. The second (conservative) irradiation scenario assumes an extended irradiation time over the full anticipated lifetime of ITER. The radioactivity inventories, the afterheat and the contact gamma dose were calculated as function of the decay time. Data were processed for the total activity, afterheat and contact dose rates of the TBM, its constituting components and materials

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  14. Testing of the prototype FMIT target with liquid lithium

    International Nuclear Information System (INIS)

    Miller, W.C.; Annese, C.E.; Berg, J.D.; Miles, R.R.

    1984-01-01

    Testing of a molten lithium target was performed to evaluate hydraulic stability, determine surface evaporation rates, and map the detailed contour of the high speed, free surface wall jet. The results confirmed predictions by demonstrating acceptable performance of a prototype target

  15. Two-dimensional cross-section sensitivity and uncertainty analysis of the LBM [Lithium Blanket Module] experiments at LOTUS

    International Nuclear Information System (INIS)

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

    1988-01-01

    In a recent common Los Alamos/PSI effort, a sensitivity and nuclear data uncertainty path for the modular code system AARE (Advanced Analysis for Reactor Engineering) was developed. This path includes the cross-section code TRAMIX, the one-dimensional finite difference S/sub N/-transport code ONEDANT, the two-dimensional finite element S/sub N/-transport code TRISM, and the one- and two-dimensional sensitivity and nuclear data uncertainty code SENSIBL. Within the framework of the present work a complete set of forward and adjoint two-dimensional TRISM calculations were performed both for the bare, as well as for the Pb- and Be-preceeded, LBM using MATXS8 libraries. Then a two-dimensional sensitivity and uncertainty analysis for all cases was performed. The goal of this analysis was the determination of the uncertainties of a calculated tritium production per source neutron from lithium along the central Li 2 O rod in the LBM. Considered were the contributions from 1 H, 6 Li, 7 Li, 9 Be, /sup nat/C, 14 N, 16 O, 23 Na, 27 Al, /sup nat/Si, /sup nat/Cr, /sup nat/Fe, /sup nat/Ni, and /sup nat/Pb. 22 refs., 1 fig., 3 tabs

  16. Technical issues of RAFMs for the fabrication of ITER Test Blanket Module

    International Nuclear Information System (INIS)

    Tanigawa, Hiroyasu; Hirose, Takanori; Shiba, Kiyoyuki

    2007-01-01

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems, as it has they have been developed based on massive industrial experience of ferritic/martensitic steel replacing Mo and Nb of high chromium heat resistant martensitic steels (such as modified 9Cr-1Mo) with W and Ta, respectively. F82H and JLF-1 are RAFMs, which have been developed and studied in Japan and the various effects of irradiation were reported. F82H is designed with emphasis on high temperature property and weldability, and was provided and evaluated in various countries as a part of the IEA fusion materials development collaboration. The JAEA/US collaboration program also has been conducted with the emphasis on irradiation effects of F82H. Now, among the existing database for RAFMs the most extensive one is that for F82H. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of ITER Test Blanket Module (TBM) suggested from the recent achievements in Japan. It is desirable to make the status of RAFMs equivalent to commercial steels to use RAFMs as the ITER-TBM structural material. This would require demonstrating the reproducibility and weldability as well as providing the database. The excellent reproducibility of F82H has been demonstrated with four 5-ton-heats, and two of them were provided as F82H-IEA heats. It has been also proved that F82H could be provided as plates (thickness of 1.5 to 55 mm), pipes and rectangular tubes. It is also important to have the excellent weldability as the TBM has about 300m length of weld line, and it was proved through TIG, EB and YAG weld test performed in air atmosphere. Various mechanical and microstructural data have been accumulated including long-term tests such as creep rupture tests and aging tests. Although F82H is a well-perceived RAFM as the ITER-TBM structural material, some issues are

  17. Application of the MIT two-channel model to predict flow recirculation in WARD 61-pin blanket tests

    International Nuclear Information System (INIS)

    Huang, T.T.; Todreas, N.E.

    1983-01-01

    The preliminary application of MIT two-channel model to WARD sodium blanket tests was presented in this report. The criterion was employed to predict the recirculation for selected completed (transient and steady state) and proposed (transient only) tests. The heat loss was correlated from the results of the WARD zero power tests. The calculational results show that the criterion agrees with the WARD tests except for WARD RUN 718 for which the criterion predicts a different result from WARD data under bundle heat loss condition. However, if the test assembly is adiabatic, the calculations predict an operating point which is marginally close to the mixed-to-recirculation transition regime

  18. Application of the MIT two-channel model to predict flow recirculation in WARD 61-pin blanket tests

    International Nuclear Information System (INIS)

    Huang, T.T.; Todreas, N.E.

    1983-01-01

    The preliminary application of MIT TWO-CHANNEL MODEL to WARD sodium blanket tests was presented in this report. Our criterion was employed to predict the recirculation for selected completed (transient and steady state) and proposed (transient only) tests. The heat loss was correlated from the results of the WARD zero power tests. The calculational results show that our criterion agrees with the WARD tests except for WARD RUN 718 for which the criterion predicts a different result from WARD data under bundle heat loss condition. However, if the test assembly is adiabatic, the calculations predict an operating point which is marginally close to the mixed-to-recirculation transition regime

  19. The fabrication of a vanadium-stainless steel test section for MHD testing of insulator coatings in flowing lithium

    International Nuclear Information System (INIS)

    Reed, C.B.; Mattas, R.F.; Smith, D.L.; Chung, H.; Tsai, H.-C.; Morgan, G.D.; Wille, G.W.; Young, C.

    1996-01-01

    To test the magnetohydrodynamic (MHD) pressure drop reduction performance of candidate insulator coatings for the ITER Vanadium/Lithium Breeding Blanket, a test section comprised of a V- 4Cr-4Ti liner inside a stainless steel pipe was designed and fabricated. Theoretically, the MHD pressure drop reduction benefit resulting, from an electrically insulating coating on a vanadium- lined pipe is identical to the benefit derived from an insulated pipe fabricated of vanadium alone. A duplex test section design consisting of a V alloy liner encased in a SS pressure boundary provided protection for vanadium from atmospheric contamination during operation at high temperature and obviated any potential problems with vanadium welding while also minimizing the amount of V alloy material required. From the MHD and insulator coating- point of view, the test section outer SS wall and inner V alloy liner can be modeled simply as a wall having a sandwich construction. Two 52.3 mm OD x 2.9 m long V-alloy tubes were fabricated by Century Tubes from 64 mm x 200 mm x 1245 mm extrusions produced by Teledyne Wah Chang. The test section's duplex structure was subsequently fabricated at Century Tubes by drawing down a SS pipe (2 inch schedule 10) over one of the 53.2 mm diameter V tubes

  20. 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)

  1. Testing Conducted for Lithium-Ion Cell and Battery Verification

    Science.gov (United States)

    Reid, Concha M.; Miller, Thomas B.; Manzo, Michelle A.

    2004-01-01

    The NASA Glenn Research Center has been conducting in-house testing in support of NASA's Lithium-Ion Cell Verification Test Program, which is evaluating the performance of lithium-ion cells and batteries for NASA mission operations. The test program is supported by NASA's Office of Aerospace Technology under the NASA Aerospace Flight Battery Systems Program, which serves to bridge the gap between the development of technology advances and the realization of these advances into mission applications. During fiscal year 2003, much of the in-house testing effort focused on the evaluation of a flight battery originally intended for use on the Mars Surveyor Program 2001 Lander. Results of this testing will be compared with the results for similar batteries being tested at the Jet Propulsion Laboratory, the Air Force Research Laboratory, and the Naval Research Laboratory. Ultimately, this work will be used to validate lithium-ion battery technology for future space missions. The Mars Surveyor Program 2001 Lander battery was characterized at several different voltages and temperatures before life-cycle testing was begun. During characterization, the battery displayed excellent capacity and efficiency characteristics across a range of temperatures and charge/discharge conditions. Currently, the battery is undergoing lifecycle testing at 0 C and 40-percent depth of discharge under low-Earth-orbit (LEO) conditions.

  2. 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.)

  3. 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.

  4. Thermo-mechanical characterization of ceramic pebbles for breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Lo Frano, Rosa, E-mail: rosa.lofrano@ing.unipi.it; Aquaro, Donato; Scaletti, Luca

    2016-11-01

    Highlights: • Experimental activities to characterize the Li{sub 4}SiO{sub 4}. • Compression tests of pebbles. • Experimental evaluation of thermal conductivity of pebbles bed at different temperatures. • Experimental test with/without compression load. - Abstract: An open issue for fusion power reactor is to design a suitable breeding blanket capable to produce the necessary quantity of the tritium and to transfer the energy of the nuclear fusion reaction to the coolant. The envisaged solution called Helium-Cooled Pebble Bed (HCPB) breeding blanket foresees the use of lithium orthosilicate (Li{sub 4}SiO{sub 4}) or lithium metatitanate (Li{sub 2}TiO{sub 3}) pebble beds. The thermal mechanical properties of the candidate pebble bed materials are presently extensively investigated because they are critical for the feasibility and performances of the numerous conceptual designs which use a solid breeder. This study is aimed at the investigation of mechanical properties of the lithium orthosilicate and at the characterization of the main chemical, physical and thermo-mechanical properties taking into account the production technology. In doing that at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa adequate experiments were carried out. The obtained results may contribute to characterize the material of the pebbles and to optimize the design of the envisaged fusion breeding blankets.

  5. Remote handling of the blanket segments: testing of 1/3 scale mock-ups at the Robertino facility

    International Nuclear Information System (INIS)

    Maisonnier, D.; Amelotti, F.; Chiasera, A.; Gaggini, P.; Damiani, C.; Degli Esposti, L.; Gatti, G.; Castillo, E.; Caravati, D.; Farfalletti-Casali, F.; Gritzmann, P.; Ruiz, E.

    1995-01-01

    The remote replacement of blanket segments inside the vacuum vessel of a fusion reactor is probably the most complex task from the maintenance standpoint. Its success will rely on the definition of appropriate handling concepts and equipment, but also on a ''maintenance friendly'' reactor layout and blanket design. The key difficulty is the lack of rigidity of the segments which results in considerable deformations since they cannot be gripped above their centre of gravity. These deformations may be up to five times greater than the assembly clearance and one order of magnitude larger than the required positioning accuracy. Experimental activities have been undertaken to select appropriate handling devices and procedures, to assess the design of the components handled, and to review specific technical issues such as kinematics and dynamics performance, trajectory planning and control and sensors requirement for the handling devices. Work was performed in the Robertino facility where two handling concepts have been tested at a 1/3 scale. (orig.)

  6. 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.)

  7. Thermonuclear blankets

    International Nuclear Information System (INIS)

    Kubota, Tadashi.

    1986-01-01

    Purpose: To increase the effective thermal conductivity between lithium ceramic spheres and a metal container. Constitution: The surface of lithium ceramic spheres is coated with a nickel metal, which is further coated with a thin copper layer. Then, copper spheres with a diameter smaller than that of the lithium ceramic spheres are packed in admixture together with the lithium ceramic spheres in an appropriate volume ratio. Since, copper as a relatively soft metal is coated on the surface of the lithium ceramic spheres and the copper spheres are charged to the gaps between each of the lithium ceramic spheres, the area of contact between the lithium ceramic spheres to each other and that between the lithium ceramic spheres and the metal container are easily increased to improve the effective thermal conductivity, by which the heat removing performance of the plant can be improved. (Yoshino, Y.)

  8. Fusion Materials Irradiation Test (FMIT) facility lithium system: a design and development status

    International Nuclear Information System (INIS)

    Brackenbury, P.J.; Bazinet, G.D.; Miller, W.C.

    1983-01-01

    The design and development of the Fusion Materials Irradiation Test (FMIT) Facility lithium system is outlined. This unique liquid lithium recirculating system, the largest of its kind in the world, is described with emphasis on the liquid lithium target assembly and other important components necessary to provide lithium flow to the target. The operational status and role of the Experimental Lithium System (ELS) in the design of the FMIT lithium system are discussed. Safety aspects of operating the FMIT lithium system in a highly radioactive condition are described. Potential spillage of the lithium is controlled by cell liners, by argon flood systems and by remote maintenance features. Lithium chemistry is monitored and controlled by a side-stream loop, where impurities measured by instruments are collected by hot and cold traps

  9. Fusion Materials Irradiation Test (FMIT) facility lithium system: a design and development status

    Energy Technology Data Exchange (ETDEWEB)

    Brackenbury, P.J.; Bazinet, G.D.; Miller, W.C.

    1983-01-01

    The design and development of the Fusion Materials Irradiation Test (FMIT) Facility lithium system is outlined. This unique liquid lithium recirculating system, the largest of its kind in the world, is described with emphasis on the liquid lithium target assembly and other important components necessary to provide lithium flow to the target. The operational status and role of the Experimental Lithium System (ELS) in the design of the FMIT lithium system are discussed. Safety aspects of operating the FMIT lithium system in a highly radioactive condition are described. Potential spillage of the lithium is controlled by cell liners, by argon flood systems and by remote maintenance features. Lithium chemistry is monitored and controlled by a side-stream loop, where impurities measured by instruments are collected by hot and cold traps.

  10. Rail deployment and storage procedure and test for ITER blanket remote maintenance

    International Nuclear Information System (INIS)

    Kakudate, S.; Shibanuma, K.

    2003-01-01

    A concept of rail-mounted vehicle manipulator system has been developed to apply to the maintenance of the ITER blanket composed of ∼400 modules in the vacuum vessel. The most critical issue of the vehicle manipulator system is the feasibility of the deployment and storage of the articulated rail, composed of eight rail links without any driving mechanism in the joints. To solve this issue, a new driving mechanism and procedure for the rail deployment and storage has been proposed, taking account of the repeated operation of the multi-rail links deployed and stored in the same kinematical manner. The new driving mechanism, which is different from those of a usual articulated manipulator or 'articulated boom' equipped with actuators in every joint for movement, is composed of three external mechanisms installed outside the articulated rail, i.e. a vehicle traveling mechanism as main driver and two auxiliary driving mechanisms. A simplified synchronized control of three driving mechanisms has also been proposed, including 'torque-limit control' for suppression of the overload of the mechanisms. These proposals have been tested using a full-scale vehicle manipulator system, in order to demonstrate the proof of principle for rail deployment and storage. As a result, the articulated rail has been successfully deployed and stored within 6 h each, less than the target of 8 h, by means of the three external driving mechanisms and the proposed synchronized control. In addition, the overload caused by an unexpected mismatch of the synchronized control of three driving mechanisms has also been successfully suppressed less than the rated torque by the proposed 'torque-limit control'. It is therefore concluded that the feasibility of the rail deployment and storage of the vehicle manipulator system has been demonstrated

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

    International Nuclear Information System (INIS)

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

    1991-10-01

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

  12. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  14. Design data, liquid distributors and condenser for a distillation column to enrich tritium in metallic lithium

    International Nuclear Information System (INIS)

    Barnert, E.

    1984-01-01

    Tritium, one fuel component of the fusion reactor is bred from the reactors blanket material lithium. Before extracting the tritium from, for instance, metallic lithium by permeation it has to be enriched in the lithium by high temperature distillation. In this report design data for a column for high temperature distillation are given. About the testing of two distributors for small liquid quantities and of a condenser is reported. (orig.) [de

  15. Development and testing of a zero stitch MLI blanket using plastic pins for space use

    OpenAIRE

    畠中, 龍太; 宮北, 健; 杉田, 寛之; Saitoh, Masanori; Hirai, Tomoyuki; Hatakenaka, Ryuta; Miyakita, Takeshi; Sugita, Hiroyuki; Saitoh, Masanori; Hirai, Tomoyuki

    2014-01-01

    New types of MLI blanket have been developed to achieve high thermal performance while maintaining production and assembly workability equivalent to the conventional type. Tag-pins, which are widely used in commercial applications to hook price tags to products, are used to fix the films in place and the pin material is changed to polyetheretherketone (PEEK) for use in space. Thermal performance is measured by using a boil-off calorimeter, in which a rectangular liquid nitrogen tank is used t...

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

    International Nuclear Information System (INIS)

    Tsuru, Daigo; Enoeda, Mikio; Akiba, Masato

    2007-01-01

    Water Cooled Solid Breeder Test Blanket Module (WCSB TBM) is being designed by JAEA for the primary candidate TBM of Japan, and the safety evaluation of WCSB TBM has been performed. This reports presents summary of safety evaluation activities of the Japanese WCSB TBM, including nuclear analysis, source of RI, waste evaluation, occupational radiolysis exposure (ORE), failure mode effect analysis (FMEA) and postulated initiating event (PIE). For the purpose of basic evaluation of source terms on nuclear heating and radioactivity generation, two-dimensional nuclear analysis has been carried out. By the nuclear analysis, distributions of neutron flux, tritium breeding ratio (TBR), nuclear heat, decay heat and induced activity are calculated. Tritium production is calculated by the nuclear analysis by integrating distributions of TBR values, as about 0.2 g-T/FPD. With respect to the radioactive waste, the induced activity of the irradiated TBM is estimated. For the purpose of occupational radiolysis exposure (ORE), RI inventory is estimated. Tritium inventory in pebble bed of TBM is about 3 x 10 12 Bq, and tritium in purge gas is about 3 x 10 11 Bq. FMEA has been carried out to identify the PIEs that need safety evaluation. PIEs are summarized into three groups, i.e., heating, pressurization and release of RI. PIEs of local heating are converged without any special cares. With respect to heating of whole module, two PIEs are selected as the most severe events, i.e., loss of cooling of TBM during plasma operation and ingress of coolant into TBM during plasma operation. With respect to PIEs about pressurization, the PIEs of pressurization of the compartment nearby the pipes of cooling system are evaluated, because rupture of the pipes result pressurization of such compartments, i.e., box structure of TBM, purge gas loop, TRS, VV, port cell and TCWS vault. Box structure of TBM is designed to withstand the maximum pressure of the cooling system. At other compartments

  17. Preparation of refractory cermet structures for lithium compatibility testing

    Science.gov (United States)

    Heestand, R. L.; Jones, R. A.; Wright, T. R.; Kizer, D. E.

    1973-01-01

    High-purity nitride and carbide cermets were synthesized for compatability testing in liquid lithium. A process was developed for the preparation of high-purity hafnium nitride powder, which was subsequently blended with tungsten powder or tantalum nitride and tungsten powders and fabricated into 3 in diameter billets by uniaxial hot pressing. Specimens were then cut from the billets for compatability testing. Similar processing techniques were applied to produce hafnium carbide and zirconium carbide cermets for use in the testing program. All billets produced were characterized with respect to chemistry, structure, density, and strength properties.

  18. 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

  19. Performance Tests of a Permeation Sensor for Test Blanket Modules Using Liquid Metal

    International Nuclear Information System (INIS)

    Choi, B. G.; Lee, D. W.; Lee, E. H.; Yoon, J. S.; Kim, S. K.; Shin, K. I.; Jin, H. G.

    2013-01-01

    The tritium extraction from a breeder is one of the key technologies and its methods have been investigated. For developing the tritium extraction methods and evaluating the amount of tritium in the system, a reliable and correct sensor is required to measure the hydrogen concentration in liquid metal breeder. There are several researches for developing the sensors in the ITER participants and especially, EU has developed the permeation sensors trying to selecting materials with low Serviette's constant (solubility) and high hydrogen diffusivity coefficient. However, EU's response time is still too long time about tens of minutes to measure the tritium concentration in the online system. We have been performing the preliminary tests with designed and fabricated sensors to solve the late response of sensor. However, we could not continue the tests because of the membrane's oxidation (pure Fe) and the difficulty of welding nonferrous metals. In present study, a permeation sensor made of vacuum flanges with a porous plate inside is proposed not only to eliminate the difficulty of the fabrication but to optimize the performance of sensor. The permeation sensor to measure the hydrogen isotopes in liquid metal breeder has been proposed and evaluated to overcome the limitation of a long response time for various shapes and materials. We found that the previous sensors have limitation; the oxidation problems (pure Fe) and the difficulty in welding (nonferrous metals). Therefore we proposed a permeation sensor with the vacuum flanges filled with porous disks to eliminate the problems. By using the CF flanges, the problem caused by welding is removed. But the permeable response time of sensors took a long time to reach the pressure equivalent

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

    International Nuclear Information System (INIS)

    Wang Zeming; Chen Lu; Hu Gang

    2014-01-01

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

  1. Controlled encoding strategies in memory tests in lithium patients.

    Science.gov (United States)

    Opgenoorth, E; Karlick-Bolten, E

    1986-03-01

    The "levels of processing" theory (Craik and Lockhart) and "dual coding" theory (Paivio) provide new aspects for clinical memory research work. Therefore, an incidental learning paradigm on the basis of these two theoretical approaches was chosen to test aspects of memory performances with lithium therapy. Results of two experiments, with controlled non-semantic processing (rating experiment "comparison of size") and additive semantic processing (rating "living--non-living") indicate a slight reduction in recall (Fig. 1) and recognition performance (Fig. 2) in lithium patients. Effects on encoding strategies are of equal quality in patients and healthy subjects (Tab. 1, 2) but performance differs between both groups: poorer systematic benefit from within code repetitions ("word-word" items, "picture-picture" items) and dual coding (repeated variable item presentation "picture-word") is obtained. The less efficient encoding strategies in the speeded task are discussed with respect to cognitive rigidity and slowing of performance by emotional states. This investigation of so-called "memory deficits" with lithium is an attempt to explore impairments at an early stage of processing; the characterization of the perceptual cognitive analysis seems useful for further clinical research work on this topic.

  2. 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)

  3. Tritium recovery from lithium oxide pellets

    International Nuclear Information System (INIS)

    Bertone, P.C.; Jassby, D.L.

    1984-01-01

    The TFTR Lithium Blanket Module is an assembly containing 650 kg of lithium oxide that will be used to test the ability of neutronics codes to model the tritium breeding characteristics of limited-coverage breeding zones in a tokamak. It is required that tritium concentrations as low as 0.1 nCi/g bred in both metallic lithium samples and lithium oxide pellets be measured with an uncertainty not exceeding +- 6%. A tritium assay technique for the metallic samples which meets this criterion has been developed. Two assay techniques for the lithium oxide pellets are being investigated. In one, the pellets are heated in a flowing stream of hydrogen, while in the other, the pellets are dissolved in 12 M hydrochloric acid

  4. Radiolysis and corrosion aspects of the aqueous self-cooled blanket concept

    International Nuclear Information System (INIS)

    Bruggeman, A.; Snykers, M.; Bogaerts, W.F.; Waeben, R.; Embrechts, M.J.; Steiner, D.

    1989-01-01

    Corrosion and radiolysis aspects of the Aqueous Self-Cooled Blanket concept, proposed as a potential shielding breeding blanket for near term fusion devices and fusion reactors, have been investigated. On the basis of preliminary results for selected aqueous solutions of lithium compounds, no particular corrosion problems have been revealed for the low-temperature concept envisaged for NET and radiolysis effects might be controlled by appropriate countermeasures. For the reactor-relevant high-temperature concept particular attention has to be paid to intergranular stress-corrosion and to the synergistic radiolysis-corrosion effects. Further information is needed from tests performed in relevant operational conditions. (orig.)

  5. Studies on Flat Sandwich-type Self-Powered Detectors for Flux Measurements in ITER Test Blanket Modules

    Science.gov (United States)

    Raj, Prasoon; Angelone, Maurizio; Döring, Toralf; Eberhardt, Klaus; Fischer, Ulrich; Klix, Axel; Schwengner, Ronald

    2018-01-01

    Neutron and gamma flux measurements in designated positions in the test blanket modules (TBM) of ITER will be important tasks during ITER's campaigns. As part of the ongoing task on development of nuclear instrumentation for application in European ITER TBMs, experimental investigations on self-powered detectors (SPD) are undertaken. This paper reports the findings of neutron and photon irradiation tests performed with a test SPD in flat sandwich-like geometry. Whereas both neutrons and gammas can be detected with appropriate optimization of geometries, materials and sizes of the components, the present sandwich-like design is more sensitive to gammas than 14 MeV neutrons. Range of SPD current signals achievable under TBM conditions are predicted based on the SPD sensitivities measured in this work.

  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. Enhancement of Tokamak Fusion Test Reactor performance by lithium conditioning

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Hill, K.W.; Strachan, J.D.; Bell, M.G.; Scott, S.D.; Budny, R.; Marmar, E.S.; Snipes, J.A.; Terry, J.L.; Batha, S.; Bell, R.E.; Bitter, M.; Bush, C.E.; Chang, Z.; Darrow, D.S.; Ernst, D.; Fredrickson, E.; Grek, B.; Herrmann, H.W.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Levinton, F.M.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.; Ramsey, A.T.; Roquemore, A.L.; Skinner, C.H.; Stevenson, T.; Stratton, B.C.; Synakowski, E.; Taylor, G.; von Halle, A.; von Goeler, S.; Wong, K.L.; Zweben, S.J.

    1996-01-01

    Wall conditioning in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] by injection of lithium pellets into the plasma has resulted in large improvements in deuterium endash tritium fusion power production (up to 10.7 MW), the Lawson triple product (up to 10 21 m -3 s keV), and energy confinement time (up to 330 ms). The maximum plasma current for access to high-performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through (1) distributing the Li over the limiter surface by injection of four Li pellets into Ohmic plasmas of increasing major and minor radius, and (2) injection of four Li pellets into the Ohmic phase of supershot discharges before neutral-beam heating is begun. copyright 1996 American Institute of Physics

  8. Diagnostics of high-speed liquid lithium jet for IFMIF/EVEDA lithium test loop

    International Nuclear Information System (INIS)

    Kanemura, Takuji; Kondo, Hiroo; Furukawa, Tomohiro; Sugiura, Hirokazu; Horiike, Hiroshi; Yamaoka, Nobuo; Ida, Mizuho; Nakamura, Kazuyuki; Matsushita, Izuru

    2011-01-01

    Regarding R and Ds on the International Fusion Materials Irradiation Facility (IFMIF), hydraulic stability of the liquid Li jet simulating the IFMIF Li target is planned to be validated using EVEDA Li Test Loop (ELTL). IFMIF is an accelerator-based deuteron-lithium (Li) neutron source for research and development of fusion reactor materials. The stable Li target is required in IFMIF to maintain the quality of the neutron fluence and integrity of the Li target itself. This paper presents diagnostics of the Li jet to be implemented in validation tests of the jet stability in ELTL, and those specifications and methodologies are introduced. In the tests, the following physical parameters need to be measured; thickness of the jet; surface structure (height, length/width and frequency of free-surface waves); local flow velocity at the free surface; and Li evaporation rate. With regard to measurement of jet thickness and the surface wave height, a contact-type liquid level sensor is to be used. As for measurement of wave velocity and visual understanding of detailed free-surface structure, a high-speed video camera is to be leveraged. With respect to Li evaporation measurement, weight change of specimens installed near the free surface and frequency change of a crystal quartz are utilized. (author)

  9. Results and code prediction comparisons of lithium-air reaction and aerosol behavior tests

    International Nuclear Information System (INIS)

    Jeppson, D.W.

    1986-03-01

    The Hanford Engineering Development Laboratory (HEDL) Fusion Safety Support Studies include evaluation of potential safety and environmental concerns associated with the use of liquid lithium as a breeder and coolant for fusion reactors. Potential mechanisms for volatilization and transport of radioactive metallic species associated with breeder materials are of particular interest. Liquid lithium pool-air reaction and aerosol behavior tests were conducted with lithium masses up to 100 kg within the 850-m 3 containment vessel in the Containment Systems Test Facility. Lithium-air reaction rates, aerosol generation rates, aerosol behavior and characterization, as well as containment atmosphere temperature and pressure responses were determined. Pool-air reaction and aerosol behavior test results were compared with computer code calculations for reaction rates, containment atmosphere response, and aerosol behavior. The volatility of potentially radioactive metallic species from a lithium pool-air reaction was measured. The response of various aerosol detectors to the aerosol generated was determined. Liquid lithium spray tests in air and in nitrogen atmospheres were conducted with lithium temperatures of about 427 0 and 650 0 C. Lithium reaction rates, containment atmosphere response, and aerosol generation and characterization were determined for these spray tests

  10. In-pile test of tritium recovery from lithium oxide

    International Nuclear Information System (INIS)

    Kurasawa, Toshimasa; Yoshida, Hiroshi; Watanabe, Hitoshi; Takeshita, Hidefumi; Miyauchi, Takejiro; Matsui, Tomoaki

    1984-05-01

    In-situ tritium recovery experiment with sintered lithium oxide pellets was performed under a high neutron fluence in the JRR-2. The irradiation hole VT-10 is the vertical one in the fuel rods region of the reactor, and the neutron flux is as follows: the thermal neutron flux with the epithermal neutron; 1.12 x 10 14 n/cm 2 . sec, the fast neutron flux; 1.0 x 10 12 n/cm 2 . sec. Irradiation material is the four pellets of cylindrical Li 2 O with the size of 11mm-OD, 1.8mm-ID, 10mm-H, and their total weight is 6.67g(the apparent bulk density 86%TD). A sweep gas capsule with a inner heater was constructed for the present study. Irradiation temperatures were regulated in the high temperature range, 470 -- 760 0 C. Four cycles of irradiation tests were carried out from May to August in 1983, and the effective thermal neutron fluence and the burnup of 6 Li were 5.9 x 10 19 nvt and 0.24% of total lithium(natural abundance of Li), respectively. The amount of generated tritium was calculated to be 31.2Ci by using a value of the depression factor of the thermal neutron flux(0.148) and the effective neutron cross section(543b) for the 6 Li(n, α) 3 H reaction. Present report describes the tritium release behavior in the in-situ tritium recovery apparatus and discuss the effects of the moisture, the hydrogen spiking, the irradiation temperature, etc.. Problems relative to a real time measurement of a comparatively high tritium concentration(10 -1 -- 10 2 μCi/cm 3 ) in the helium gas stream were also investigated. (author)

  11. Charge-Control Unit for Testing Lithium-Ion Cells

    Science.gov (United States)

    Reid, Concha M.; Mazo, Michelle A.; Button, Robert M.

    2008-01-01

    A charge-control unit was developed as part of a program to validate Li-ion cells packaged together in batteries for aerospace use. The lithium-ion cell charge-control unit will be useful to anyone who performs testing of battery cells for aerospace and non-aerospace uses and to anyone who manufacturers battery test equipment. This technology reduces the quantity of costly power supplies and independent channels that are needed for test programs in which multiple cells are tested. Battery test equipment manufacturers can integrate the technology into their battery test equipment as a method to manage charging of multiple cells in series. The unit manages a complex scheme that is required for charging Li-ion cells electrically connected in series. The unit makes it possible to evaluate cells together as a pack using a single primary test channel, while also making it possible to charge each cell individually. Hence, inherent cell-to-cell variations in a series string of cells can be addressed, and yet the cost of testing is reduced substantially below the cost of testing each cell as a separate entity. The unit consists of electronic circuits and thermal-management devices housed in a common package. It also includes isolated annunciators to signal when the cells are being actively bypassed. These annunciators can be used by external charge managers or can be connected in series to signal that all cells have reached maximum charge. The charge-control circuitry for each cell amounts to regulator circuitry and is powered by that cell, eliminating the need for an external power source or controller. A 110-VAC source of electricity is required to power the thermal-management portion of the unit. A small direct-current source can be used to supply power for an annunciator signal, if desired.

  12. The accomplishments of lithium target and test facility validation activities in the IFMIF/EVEDA phase

    Science.gov (United States)

    Arbeiter, Frederik; Baluc, Nadine; Favuzza, Paolo; Gröschel, Friedrich; Heidinger, Roland; Ibarra, Angel; Knaster, Juan; Kanemura, Takuji; Kondo, Hiroo; Massaut, Vincent; Saverio Nitti, Francesco; Miccichè, Gioacchino; O'hira, Shigeru; Rapisarda, David; Sugimoto, Masayoshi; Wakai, Eiichi; Yokomine, Takehiko

    2018-01-01

    As part of the engineering validation and engineering design activities (EVEDA) phase for the international fusion materials irradiation facility IFMIF, major elements of a lithium target facility and the test facility were designed, prototyped and validated. For the lithium target facility, the EVEDA lithium test loop was built at JAEA and used to test the stability (waves and long term) of the lithium flow in the target, work out the startup procedures, and test lithium purification and analysis. It was confirmed by experiments in the Lifus 6 plant at ENEA that lithium corrosion on ferritic martensitic steels is acceptably low. Furthermore, complex remote handling procedures for the remote maintenance of the target in the test cell environment were successfully practiced. For the test facility, two variants of a high flux test module were prototyped and tested in helium loops, demonstrating their good capabilities of maintaining the material specimens at the desired temperature with a low temperature spread. Irradiation tests were performed for heated specimen capsules and irradiation instrumentation in the BR2 reactor at SCK-CEN. The small specimen test technique, essential for obtaining material test results with limited irradiation volume, was advanced by evaluating specimen shape and test technique influences.

  13. 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

  14. 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)

  15. 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)

  16. Testing of Liquid Lithium Limiters in CDX-U

    Energy Technology Data Exchange (ETDEWEB)

    R. Majeski; R. Kaita; M. Boaz; P. Efthimion; T. Gray; B. Jones; D. Hoffman; H. Kugel; J. Menard; T. Munsat; A. Post-Zwicker; V. Soukhanovskii; J. Spaleta; G. Taylor; J. Timberlake; R. Woolley; L. Zakharov; M. Finkenthal; D. Stutman; G. Antar; R. Doerner; S. Luckhardt; R. Seraydarian; R. Maingi; M. Maiorano; S. Smith; D. Rodgers

    2004-07-30

    Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid-metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid-lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm2, subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now be en performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments the liquid-lithium plasma-facing area was increased to 2000 cm2. Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described.

  17. Testing of Liquid Lithium Limiters in CDX-U

    International Nuclear Information System (INIS)

    Majeski, R.; Kaita, R.; Boaz, M.; Efthimion, P.; Gray, T.; Jones, B.; Hoffman, D.; Kugel, H.; Menard, J.; Munsat, T.; Post-Zwicker, A.; Soukhanovskii, V.; Spaleta, J.; Taylor, G.; Timberlake, J.; Woolley, R.; Zakharov, L.; Finkenthal, M.; Stutman, D.; Antar, G.; Doerner, R.; Luckhardt, S.; Seraydarian, R.; Maingi, R.; Maiorano, M.; Smith, S.; Rodgers, D.

    2004-01-01

    Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid-metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid-lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm2, subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now be en performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments the liquid-lithium plasma-facing area was increased to 2000 cm2. Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described

  18. Testing of liquid lithium limiters in CDX-U

    International Nuclear Information System (INIS)

    Majeski, R.; Kaita, R.; Boaz, M.; Efthimion, P.; Gray, T.; Jones, B.; Hoffman, D.; Kugel, H.; Menard, J.; Munsat, T.; Post-Zwicker, A.; Spaleta, J.; Taylor, G.; Timberlake, J.; Woolley, R.; Zakharov, L.; Finkenthal, M.; Stutman, D.; Antar, G.; Doerner, R.; Luckhardt, S.; Seraydarian, R.; Maingi, R.; Maiorano, M.; Smith, S.; Rodgers, D.; Soukhanovskii, V.

    2004-01-01

    Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm 2 , subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now been performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments, the liquid lithium plasma-facing area was increased to 2000 cm 2 . Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described

  19. Measurement of lithium target surface velocity in the IFMIF/EVEDA lithium test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kanemura, Takuji, E-mail: kanemura.takuji@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan); Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan); Hoashi, Eiji [Osaka University, 2-1 Yamada-oka, Suita, Osaka 565-0871 (Japan); Yoshihashi, Sachiko; Horiike, Hiroshi [Fukui University of Technology, Gakuen 3-6-1, Fukui-shi, Fukui 910-8505 (Japan); Wakai, Eiichi [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan)

    2016-11-01

    Highlights: • The objective is to measure the free-surface velocity field of the IFMIF Li target. • The Li target has an important role to remove 10 MW heat input from a deuteron beam. • The free-surface of the Li target is under the most severe heat load condition. • Measured surface velocities are almost equal to cross-sectional average velocities. • It was confirmed that the IFMIF Li target has adequate heat removal performance. - Abstract: In the framework of the Engineering Validation and Engineering Design Activities (EVEDA) project of the International Fusion Materials Irradiation Facility (IFMIF), we measured surface velocity fields of a lithium (Li) target at the EVEDA Li test loop under specifically-designated IFMIF conditions (target speeds of 10, 15, and 20 m/s, vacuum pressure of 10{sup −3} Pa, and Li temperature of 250 °C). In the current design of the IFMIF, the free surface of the Li target is under a most severe heat load condition with respect to Li boiling. The objective of this study is to measure the actual free-surface velocity under these IFMIF conditions to evaluate the heat removal performance of the Li target. The measured results (using the surface-wave tracking method that our team developed) showed two-dimensional time-averaged velocity distributions around the IFMIF beam footprint being virtually uniform, and close to the cross-sectional average velocity. The uniformity of the velocity distributions was less than 1 m/s. The comparison between the measured and analyzed surface velocity at the beam center showed that the analysis accurately predicts the measurement results within a margin of 3%. Finally, it was confirmed that the Li target delivers adequate heat removal performance in the IFMIF as designed.

  20. Synthesis and Test of 'New' Gel-Type Lithium Electrolytes

    National Research Council Canada - National Science Library

    Scrosati, Bruno

    1994-01-01

    In this 6th two-month period we have continued the characterization of PMMA-based electrolyte membranes by examining the phenomena occurring at the interface between these membranes and the lithium...

  1. Summary report for ITER task - T68: MHD facility preparation for Li/V blanket option

    International Nuclear Information System (INIS)

    Reed, C.B.; Haglund, R.C.; Miller, M.E.

    1995-08-01

    A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the question of insulator coatings. Design calculations show that an electrically insulating layer is necessary to maintain an acceptably low MHD pressure drop. To enable experimental investigations of the MHD performance of candidate insulator materials and the technology for putting them in place, the room-temperature ALEX (Argonne's Liquid Metal EXperiment) NaK facility was upgraded to a 300 degrees C lithium system. The objective of this upgrade was to modify the existing facility to the minimum extent necessary, consistent with providing a safe, flexible, and easy to operate MHD test facility which uses lithium at ITER-relevant temperatures, Hartmann numbers, and interaction parameters. The facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups. The system design description for this lithium upgrade of the ALEX facility is given in this document

  2. Studies of the ultrasonic testing scheme on bonding quality in shield blanket of ITER

    International Nuclear Information System (INIS)

    Shi Sichao; Shen Jingling; He Fengqi; Jin Wanping

    2007-01-01

    International Thermonuclear Experimental Reactor (ITER) is an international cooperative item. One of its components, the First Wall (FW) functioning as neutron shielding and cooling, is an important part. According to the component materials, structural features, testing requirements of the FW, and the ultrasonic propagation characteristics, it is suggested that Broad-band ultrasonic can be used to test the bonding quality of the FW. According to the case mentioned above, the Broad-band Ultrasonic Testing scheme was presented, and the ultrasonic testing feasibility was analyzed theoretically in this paper. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    1995-01-01

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

  5. First-wall, blanket, and shield engineering test program for magnetically confined fusion power reactors

    International Nuclear Information System (INIS)

    Maroni, V.A.

    1980-01-01

    The key engineering areas identified for early study relate to FW/B/S system thermal-hydraulics, thermomechnics, nucleonics, electromagnetics, assembly, maintenance, and repair. Programmatic guidance derived frm planning exercises involving over thirty organizations (laboratories, industries, and universities) has indicated (1) that meaningful near term engineering testing should be feasible within the bounds of a modest funding base, (2) that there are existing facilities and expertise which can be profitably utilized in this testing, and (3) that near term efforts should focus on the measurement of engineering data and the verification/calibration of predictive methods for anticipated normal operational and transient FW/B/S conditions. The remainder of this paper discusses in more detail the planning strategies, proposed approach to near term testing, and longer range needs for integrated FW/B/S test facilities

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  7. Assessment of tritiated activities in the radwaste generated from ITER Chinese helium cooled ceramic breeding test blanket module system

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Chang An, E-mail: chenchangan@caep.cn; Liu, Lingbo; Wang, Bo; Xiang, Xin; Yao, Yong; Song, Jiangfeng

    2016-11-15

    Highlights: • Approaches were developed for calculation/evaluation of tritium activities in the materials and components of a TBM system, with tritium permeation being considered for the first time. • Almost all tritiated materials and components were considered in CNHCCB TBM system including the TBM set, connection pipes, and the ancillary tritium handling systems. • Tritium activity data in HCCB TBM system were updated. Some of which in directly tritium contacted components are to be 2 or 4 magnitudes higher than the original neutron transmutation calculations. • The radwaste amount from both operation and decommission of HCCB TBM system was evaluated. - Abstract: Chinese Helium Cooled Ceramic Breeding Test blanket Module (CNHCCB TBM) will be tested in the ITER machine for the feasibility of in pile tritium production for a future magnetic confinement fusion reactor. The tritium inventories/retentions in the material/components were evaluated and updated mainly based on the tritium diffusion/permeation theory and the analysis of some reported data. Tritiated activities rank from less than 10 Bq g{sup −1} to 10{sup 9} Bq g{sup −1} for the different materials or components, which are generally higher than those from the previous neutron transmutation calculation. The amounts of tritiated radwaste were also estimated according to the operation, decommission, maintenance and replacement strategies, which vary from several tens of kilograms to tons in the different operation phases. The data can be used both for the tritium radiological safety evaluation and radwaste management of CNHCCB TBM set and its ancillary systems.

  8. Methods to enhance blanket power density

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Miller, L.G.; Bohn, T.S.; Deis, G.A.; Longhurst, G.R.; Masson, L.S.; Wessol, D.E.; Abdou, M.A.

    1982-06-01

    The overall objective of this task is to investigate the extent to which the power density in the FED/INTOR breeder blanket test modules can be enhanced by artificial means. Assuming a viable approach can be developed, it will allow advanced reactor blanket modules to be tested on FED/INTOR under representative conditions

  9. 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

  10. 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

  11. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Waganer, L.M.

    1985-01-01

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

  12. Test of lithium capillary-pore systems on the T-11M tokamak

    International Nuclear Information System (INIS)

    Evtikhin, V.A.

    2002-01-01

    In this work the divertor plate behavior has been simulated in the quasi-stationary condition. In the previous experiments on T-11M the CPS quasi-stationary heat state has not been achieved for pulse length (≤0.1 s). The T-11M tokamak up-grade allowed its performance to be increased as follows: plasma current up to 100 kA, pulse length 0.2-0.3 s. The new lithium limiter unlike the previous versions has a thermal regulation system which permits a lithium surface initial temperature to be given from -196 to 600 deg. C. This provides for an increase in test parameter range: sorption and desorption of plasma-forming gas, lithium emission into discharge, lithium erosion, limiter deposited power and so on. The first results of experiments were presented. (author)

  13. The European ITER test blanket modules: Progress in development of fabrication technologies towards standardization

    Energy Technology Data Exchange (ETDEWEB)

    Zmitko, Milan, E-mail: milan.zmitko@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla 2, Barcelona (Spain); Thomas, Noël [ATMOSTAT, F-94815 Villejuif (France); LiPuma, Antonella; Forest, Laurent [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Cogneau, Laurence [CEA-DRT, 38000 Grenoble (France); Rey, Jörg; Neuberger, Heiko [Karlsruhe Institute of Technology (KIT), Postfach 3640, Karlsruhe (Germany); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla 2, Barcelona (Spain)

    2016-11-01

    Highlights: • Significant progress on the development of welding procedures for European TBM achieved. • Fabrication processes feasibility based on diffusion and fusion welding demonstrated. • An optimized welding scenario/sequence for TBM box assembly identified. • Future qualification of pF/WPS proposed through realization of a number of QMUs. - Abstract: The paper reviews progress achieved in development of fabrication technologies and procedures applied for manufacturing of the TBM sub-components, like, HCLL and HCPB cooling plates, HCLL/HCPB stiffening plates, and HCLL/HCPB first wall and side caps. The used technologies are based on fusion and diffusion welding techniques taking into account specificities of the EUROFER97 steel. Development of a standardized procedure complying with professional codes and standards (RCC-MRx), a preliminary fabrication/welding procedure specification (pF/WPS), is described based on fabrication and non-destructive and destructive characterization of feasibility mock-ups (FMU) aimed at assessing the suitability of a fabrication process for fulfilling the design and fabrication specifications. The main FMUs characterization results are reported (e.g. pressure resistance and helium leak tightness tests, mechanical properties and microstructure at the weld joints, geometrical characteristics of the sub-components and internal cooling channels) and the key pF/WPS steps and parameters are outlined. Also, fabrication procedures for the TBM box assembly are presently under development for the establishment of an optimized assembly sequence/scenario and development of standardized welding procedure specifications. In conclusions, further steps towards the pF/WPS qualification are briefly discussed.

  14. 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

  15. 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

  16. Thermal analysis of lithium cooled natural circulation loop module for fuel rod testing in the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Eyler, L.L.; Kim, D.; Stover, R.L.; Beaver, T.R.

    1987-01-01

    Maximum heat removal capability of a lithium cooled natural circulation fuel rod test module design is determined. Loop geometry is optimized within limitations of design specifications for nominal operation temperatures, materials, and test module environment. Results provide test module operation limits and range of potential uncertainties. 3 refs., 12 figs

  17. Performance test of diamond-like carbon films for lubricating ITER blanket maintenance equipment under GPa-level high contact stress

    International Nuclear Information System (INIS)

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

    2007-01-01

    Diamond-like carbon (DLC) coating was tested as a candidate solid lubricant for transmission gears of the maintenance equipment of the blanket of the ITER instead of an oil lubricant. The wear tests using the pin-on-disk method were performed on disks with SCM440 and SNCM420 as the base materials and coated with soft, layered, and hard DLCs. All cases satisfied the required allowable contact stress (2 GPa) and lifetime (10 4 cycles), and therefore the feasibility of the DLC coating was validated. Among the three types of DLCs, the soft DLC showed the best performance. (author)

  18. Blanket of a hybrid thermonuclear reactor with liquid- metal cooling

    International Nuclear Information System (INIS)

    Terent'ev, I.K.; Fedorovich, E.P.; Paramonov, P.M.; Zhokhov, K.A.

    1982-01-01

    Blanket design of a hybrid thermopuclear reactor with a liquid metal coolant is described. To decrease MHD-resistance for uranium zone fuel elements a cylindrical shape is suggested and movement of liquid-metal coolant in fuel element packets is presumed to be in perpendicular to the magnetic field and fuel element axes direction. The first wall is cooled by water, blanket-by lithium-lead alloy

  19. 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

  20. 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

  1. 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.)

  2. Development of filler wires for welding of reduced activation ferritic martenstic steel for India's test blanket module of ITER

    International Nuclear Information System (INIS)

    Srinivasan, G.; Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K.

    2011-01-01

    Highlights: → Weld microstructure produced by RAFMS filler wires are free from delta ferrite. → Cooling rates of by weld thermal cycles influences the presence of delta ferrite. → Weld parameters modified with higher pre heat temperature and high heat input. → PWHT optimized based on correlation of hardness between base and weld metals. → Optimised mechanical properties achieved by proper tempering of the martensite. - Abstract: Indigenous development of reduced activation ferritic martensitic steel (RAFMS) has become mandatory to India to participate in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFMS is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFMS filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFMS. Purpose of this study is to develop filler wires that can be directly used for both tungsten inert gas welding (TIG) and narrow gap tungsten inert gas welding (NG-TIG), which reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, autogenous welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using TIG process at various heat inputs with a preheat temperature of 250 deg. C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimised to qualify the filler wires without the presence of delta-ferrite in

  3. Development of filler wires for welding of reduced activation ferritic martensitic steel for India's test blanket module of ITER

    International Nuclear Information System (INIS)

    Srinivasan, G.; Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K.

    2010-01-01

    Indigenous development of reduced activation ferritic-martensitic (RAFM) steel has become necessary for India as a participant in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFM steel is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFM steel filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFM steel. The purpose of this study is to develop filler wires that can be directly used for both gas tungsten arc welding (GTAW) and for narrow-gap gas tungsten arc welding (NG-GTAW) that reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser-MIG welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using GTAW process at various heat inputs with a preheat temperature of 250 C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some amount of delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimized to qualify the filler wires without the presence of delta-ferrite in the weld metal and with optimized mechanical properties. Results showed that the weld metals are free from delta-ferrite. Tensile properties at ambient temperature and at 500 C are well above the specified values, and are much higher than the base metal values. Ductile Brittle Transition Temperature (DBTT) has been evaluated as -81 C based on the 68 J criteria. The present study highlights the basis and methodology

  4. Review: Accuracy of Salivary Lithium Testing in Treatment Monitoring in Mood Disorders

    Directory of Open Access Journals (Sweden)

    Abbas Ali Asadi

    2006-04-01

    Full Text Available Lithium preparations have been used in bipolar mood disorders since 19th Century. Lethal toxic effect of lithium due to it's narrow (therapeutic index had been known from several years ago and is the most problem when is prescribed. Serum level of lithium is accepted for monitoring of toxicity, but frequency of blood testing especially in stabilizing period is stressful for patients, also it is difficult in such psychiatric patients especially in children. Many researchers worked to find a less aggressive method. One of these methods is monitoring based on salivary lithium concentration, which is controversy according to this review articles. All papers from 1949 till now are reviewed in this article and revealed this controversy. According to this review article, three ways are being suggested to solve this problem 1- Stimulation of salivary – serum ratio of lithium based on tree separated paired tests 2- Improving methods and techniques of testing. 3- Modifying of this ratio based on natural markers.

  5. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-01

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

  7. Tritium inventory and permeation in liquid breeder blankets

    International Nuclear Information System (INIS)

    Reiter, F.

    1990-01-01

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

  8. Experimental investigation of MHD pressure losses in a mock-up of a liquid metal blanket

    Science.gov (United States)

    Mistrangelo, C.; Bühler, L.; Brinkmann, H.-J.

    2018-03-01

    Experiments have been performed to investigate the influence of a magnetic field on liquid metal flows in a scaled mock-up of a helium cooled lead lithium (HCLL) blanket. During the experiments pressure differences between points on the mock-up have been recorded for various values of flow rate and magnitude of the imposed magnetic field. The main contributions to the total pressure drop in the test-section have been identified as a function of characteristic flow parameters. For sufficiently strong magnetic fields the non-dimensional pressure losses are practically independent on the flow rate, namely inertia forces become negligible. Previous experiments on MHD flows in a simplified test-section for a HCLL blanket showed that the main contributions to the total pressure drop in a blanket module originate from the flow in the distributing and collecting manifolds. The new experiments confirm that the largest pressure drops occur along manifolds and near the first wall of the blanket module, where the liquid metal passes through small openings in the stiffening plates separating two breeder units. Moreover, the experimental data shows that with the present manifold design the flow does not distribute homogeneously among the 8 stacked boxes that form the breeding zone.

  9. Lithium-lead/water interaction. Large break experiments

    International Nuclear Information System (INIS)

    Savatteri, C.; Gemelli, A.

    1991-01-01

    One current concept in fusion blanket module design is to utilize water as coolant and liquid lithium-lead as breeding/neutron-multiplier material. Considering the possibility of certain off-normal events, it is possible that water leakage into the liquid metal may occur due to a tube rupture. The lithium-lead/water contact can lead to a thermal and chemical reaction which should provoke an intolerable pressure increase in the blanket module. For realistic simulation of such in-blanket events, the Blanket Safety Test (BLAST) facility has been built. It simulates the transient event by injecting subcooled water under high pressure into a stagnant pool of about 500 kg liquid Pb-17Li. Eight fully instrumented large break tests were carried out under different conditions. The aim of the experiments is to study the chemical and thermal process and particularly: The pressurization history of the reaction vessel, the formation and deposition of the reaction products, the identification and propagation of the reaction zones and the temperature transient in the liquid metal. In this paper the results of all tests performed are presented and discussed. (orig.)

  10. Construction of a test platform for Test Blanket Module (TBM) systems integration and maintenance in ITER Port Cell #16

    Energy Technology Data Exchange (ETDEWEB)

    Vála, Ladislav, E-mail: ladislav.vala@cvrez.cz [Centrum výzkumu Řež, Hlavní 130, 250 68 Husinec-Řež (Czech Republic); Reungoat, Mathieu, E-mail: mathieu.reungoat@cvrez.cz [Centrum výzkumu Řež, Hlavní 130, 250 68 Husinec-Řež (Czech Republic); Vician, Martin [Centrum výzkumu Řež, Hlavní 130, 250 68 Husinec-Řež (Czech Republic); Poitevin, Yves; Ricapito, Italo; Zmitko, Milan; Panayotov, Dobromir [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • A non-nuclear, full size facility – TBM platform – is under construction in CVR. • It is designed for tests, optimization and validation of TBS maintenance operations. • It will allow testing and validation of specific maintenance tools and RH equipment. • It reproduces ITER Port Cell #16, as well as the TBS interfaces and main equipment. • The TBM platform will be available for full operation in the first half of 2016. - Abstract: This paper describes a project of a non-nuclear, 1:1 scale testing platform dedicated to tests, optimization and validation of integration and maintenance operations for the European TBM systems in the ITER Port Cell #16. This TBM platform is currently under construction in Centrum výzkumu Řež, Czech Republic. The facility is realized within the scope of the SUSEN project and its full operation is foreseen in the first half of 2016.

  11. A Li-particulate blanket concept for ITER

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  13. Diagnostic Accuracy of Tests for Polyuria in Lithium-Treated Patients.

    Science.gov (United States)

    Kinahan, James Conor; NiChorcorain, Aoife; Cunningham, Sean; Freyne, Aideen; Cooney, Colm; Barry, Siobhan; Kelly, Brendan D

    2015-08-01

    In lithium-treated patients, polyuria increases the risk of dehydration and lithium toxicity. If detected early, it is reversible. Despite its prevalence and associated morbidity in clinical practice, it remains underrecognized and therefore undertreated. The 24-hour urine collection is limited by its convenience and practicality. This study explores the diagnostic accuracy of alternative tests such as questionnaires on subjective polyuria, polydipsia, nocturia (dichotomous and ordinal responses), early morning urine sample osmolality (EMUO), and fluid intake record (FIR). This is a cross-sectional study of 179 lithium-treated patients attending a general adult and an old age psychiatry service. Participants completed the tests after completing an accurate 24-hour urine collection. The diagnostic accuracy of the individual tests was explored using the appropriate statistical techniques. Seventy-nine participants completed all of the tests. Polydipsia severity, EMUO, and FIR significantly differentiated the participants with polyuria (area under the receiver operating characteristic curve of 0.646, 0.760, and 0.846, respectively). Of the tests investigated, the FIR made the largest significant change in the probability that a patient experiences polyuria (3500 mL/24 hours; interval likelihood ratio, 14). Symptomatic questioning, EMUO, and an FIR could be used in clinical practice to inform the prescriber of the probability that a lithium-treated patient is experiencing polyuria.

  14. Experimental investigations of flow distribution in coolant system of Helium-Cooled-Pebble-Bed Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Ilić, M.; Schlindwein, G., E-mail: georg.schlindwein@kit.edu; Meyder, R.; Kuhn, T.; Albrecht, O.; Zinn, K.

    2016-02-15

    Highlights: • Experimental investigations of flow distribution in HCPB TBM are presented. • Flow rates in channels close to the first wall are lower than nominal ones. • Flow distribution in central chambers of manifold 2 is close to the nominal one. • Flow distribution in the whole manifold 3 agrees well with the nominal one. - Abstract: This paper deals with investigations of flow distribution in the coolant system of the Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The investigations have been performed by manufacturing and testing of an experimental facility named GRICAMAN. The facility involves the upper poloidal half of HCPB TBM bounded at outlets of the first wall channels, at outlet of by-pass pipe and at outlets of cooling channels in breeding units. In this way, the focus is placed on the flow distribution in two mid manifolds of the 4-manifold system: (i) manifold 2 to which outlets of the first wall channels and inlet of by-pass pipe are attached and (ii) manifold 3 which supplies channels in breeding units with helium coolant. These two manifolds are connected with cooling channels in vertical/horizontal grids and caps. The experimental facility has been built keeping the internal structure of manifold 2 and manifold 3 exactly as designed in HCPB TBM. The cooling channels in stiffening grids, caps and breeding units are substituted by so-called equivalent channels which provide the same hydraulic resistance and inlet/outlet conditions, but have significantly simpler geometry than the real channels. Using the conditions of flow similarity, the air pressurized at 0.3 MPa and at ambient temperature has been used as working fluid instead of HCPB TBM helium coolant at 8 MPa and an average temperature of 370 °C. The flow distribution has been determined by flow rate measurements at each of 28 equivalent channels, while the pressure distribution has been obtained measuring differential pressure at more than 250 positions. The

  15. Experiments on 18-8 stainless steels exposed to liquid lithium. I. 1,100-hour corrosion tests in lithium of 400, 500 and 6000C in natural circulation type testing apparatus

    International Nuclear Information System (INIS)

    Nihei, I.; Sumiya, I.; Fukaya, Y.; Yamazaki, Y.

    The Japan Atomic Energy Research Institute has planned and started to carry out a series of experiments concerning fusion reactor materials. This report gives the results of the first experiments. The first test materials selected were 18-8 stainless steels, and the experiments were designed to test their behavior when exposed to liquid lithium. Natural circulation type corrosion testing devices (pots) were used as the testing apparatus, and the tests were conducted with lithium temperatures up to 600 0 C

  16. Test Program For Alumina Removal And Sodium Hydroxide Regeneration From Hanford Waste By Lithium Hydrotalcite Precipitation

    International Nuclear Information System (INIS)

    Sams, T.L.; Geinesse, D.

    2011-01-01

    This test program sets a multi-phased development path to support the development of the Lithium Hydrotalcite process, in order to raise its Technology Readiness Level from 3 to 6, based on tasks ranging from laboratory scale scientific research to integrated pilot facilities.

  17. TEST PROGRAM FOR ALUMINA REMOVAL AND SODIUM HYDROXIDE REGENERATION FROM HANFORD WASTE BY LITHIUM HYDROTALCITE PRECIPITATION

    Energy Technology Data Exchange (ETDEWEB)

    SAMS TL; GEINESSE D

    2011-01-28

    This test program sets a multi-phased development path to support the development of the Lithium Hydrotalcite process, in order to raise its Technology Readiness Level from 3 to 6, based on tasks ranging from laboratory scale scientific research to integrated pilot facilities.

  18. High shock load testing of lithium-thionyl chloride batteries

    Energy Technology Data Exchange (ETDEWEB)

    Epstein, J.; Marincic, N.

    1983-10-01

    Low rate cylindrical cells have been developed, capable of withstanding mechanical shocks up to 23,000 g's for one millisecond. The cells were based on the lithium-thionyl chloride battery system and totally hermetic stainless steel hardware incorporating a glass sealed positive terminal. Four cells in series were required to deliver 25 mA pulses at a minimum voltage of 10 V before and after such exposure to one mechanical shock. Batteries were contained in a hardened steel housing and mounted within a projectile accelerated by means of a gas gun. The velocity of the projectile was measured with electronic probes immediately before impact and the deceleration was effected using a special aluminum honeycomb structure from which the g values were calculated. A high survival rate for the cells was achieved in spite of some mechanical damage to the battery housing still present.

  19. Tritium transport in HCLL and WCLL DEMO blankets

    Energy Technology Data Exchange (ETDEWEB)

    Candido, Luigi [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy); Utili, Marco [ENEA UTIS- C.R. Brasimone, Bacino del Brasimone, Camugnano, BO (Italy); Nicolotti, Iuri [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy); Zucchetti, Massimo, E-mail: massimo.zucchetti@polito.it [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy)

    2016-11-01

    Highlights: • Tritium inventories and tritium losses are the main output of the presented model for HCLL and WCLL. • A parametric study has been performed, to show the behavior of the two systems when certain parameters are changed, in order to minimize inventories and/or losses. • An improved design is needed, in order to reduce the radiological hazard related to tritium activity. According to test number 7, HCLL-BB could be able to have a tritium inventory of 33.05 g and losses of 19.55 Ci/d. • WCLL-BB shows a very low radiological risk, much lower than that suggested (inventory: 17.48 g, losses: 3.2 Ci/d). An ptimization study has been performed aiming to minimize the water flow rate for an upgraded design. • Both for HCLL and WCLL, the most critical parameters able to produce relevant variations in inventories and losses are the helium/water fraction, the CPS/WDS and the permeation reduction factors. - Abstract: The Helium-Cooled Lithium Lead (HCLL) and Water-Cooled Lithium Lead (WCLL) Breeding Blankets are two of the four blanket designs proposed for DEMO reactor. The study of tritium transport inside the blankets is fundamental to assess their preliminary design and safety features. A mathematical model has been derived, in a new form making makes easier to determine the most critical components as far as tritium losses and tritium inventories are concerned, and to model the tritium performance of the whole system. Two cases have been studied, the former with tritium generation rate constant in time and the latter considering a typical pulsed operation for a time span of 100 h. Tritium inventories and tritium losses are the main output of the model. Tritium concentrations, inventories and losses are initially calculated and compared for the two blankets, in a reference case without permeation barriers or cold traps. A parametric study to show the behavior of the two systems when certain parameters are changed, in order to minimize inventories and

  20. Experimental system design of liquid lithium-lead alloy bubbler for DFLL-TBM

    International Nuclear Information System (INIS)

    Xie Bo; Li Junge; Xu Shaomei; Weng Kuiping

    2011-01-01

    The liquid lithium-lead alloy bubbler is a very important composition in the tritium unit of Chinese Dual-Functional Lithium Lead Test Blanket Module (DFLL-TBM). In order to complete the construction and run of the bubbler experimental system,overall design of the system, main circuit design and auxiliary system design have been proposed on the basis of theoretical calculations for the interaction of hydrogen isotope with lithium-lead alloy and experiment for hydrogen extraction from liquid lithium-lead alloy by bubbling with rotational jet nozzle. The key of this design is gas-liquid exchange packed column, to achieve the measurement and extraction of hydrogen isotopes from liquid lithium-lead alloy. (authors)

  1. 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)

  2. 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.

  3. Deuteron beam interaction with lithium jet in a neutron source test facility

    International Nuclear Information System (INIS)

    Hassanein, A.

    1996-01-01

    Testing and evaluating candidate fusion reactor materials in a high-flux, high-energy neutron environment are critical to the success and economic feasibility of a fusion device. The current understanding of materials behavior in fission-like environments and existing fusion facilities is insufficient to ensure the necessary performance of future fusion reactor components. An accelerator-based deuterium-lithium system to generate the required high neutron flux for material testing is considered to be the most promising approach in the near future. In this system, a high-energy (30-40 MeV) deuteron beam impinges on a high-speed (10-20 m/s) lithium jet to produce the high-energy (≥14 MeV) neutrons required to simulate a fusion environment via the Li (d,n) nuclear stripping reaction. Interaction of the high-energy deuteron beam and the subsequent response of the high-speed lithium jet are evaluated in detail. Deposition of the deuteron beam, jet-thermal hydraulic response, lithium-surface vaporization rate, and dynamic stability of the jet are modeled. It is found that lower beam kinetic energies produce higher surface temperature and consequently higher Li vaporization rates. Larger beam sizes significantly reduce both bulk and surface temperatures. Thermal expansion and dynamic velocities (normal to jet direction) due to beam energy deposition and momentum transfer are much lower than jet flow velocity and decrease substantially at lower beam current densities. (orig.)

  4. Corrosion studies on type AISI 316L stainless steel and other materials in lithium-salt solutions

    International Nuclear Information System (INIS)

    Zheng, J.H.; Bogaerts, W.F.; Agema, K.; Phlippo, K.; Bruggeman, A.; Lorenzetto, P.; Embrechts, M.J.

    1991-01-01

    A possible concept for the blanket for next generation fusion devices is the lithium salt blanket, where lithium salt is dissolved in an aqueous coolant in order to provide for tritium. Type AISI 316L stainless steel has been considered as a structural material for such a blanket for NET (Next European Torus), and a systematic study of the corrosion behaviour of 316L stainless steel has been carried out in a number of lithium salt solutions. The experiments include cyclic potentiodynamic polarization measurement, crevice corrosion fatigue and stress corrosion cracking (SCC) tests. This paper presents a part of novel corrosion results concerning the compatibility of 316L steel and a series of other materials relevant to a fusion blanket environment. No major uniform corrosion problem has been observed, but localized corrosion, particularly corrosion fatigue and SCC, of 316L stainless steel have been found so far in a lithium hydroxide solution under some specific potential conditions. The critical electrochemical potential zones for SCC have been identified in the present study. (orig.)

  5. 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

  6. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    1980-08-01

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

  7. Tokamak blanket design study, final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

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

  8. 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)

  9. 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.

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

    International Nuclear Information System (INIS)

    Merrill, B.J.

    1980-01-01

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

  11. Reference Performance Test Methodology for Degradation Assessment of Lithium-Sulfur Batteries

    DEFF Research Database (Denmark)

    Knap, Vaclav; Stroe, Daniel-Ioan; Purkayastha, Rajlakshmi

    2018-01-01

    Lithium-Sulfur (Li-S) is an emerging battery technology receiving a growing amount of attention due to its potentially high gravimetric energy density, safety, and low production cost. However, there are still some obstacles preventing its swift commercialization. Li-S batteries are driven...... by different electrochemical processes than commonly used Lithium-ion batteries, which often results in very different behavior. Therefore, the testing and modeling of these systems have to be adjusted to reflect their unique behavior and to prevent possible bias. A methodology for a Reference Performance Test...... (RPT) for the Li-S batteries is proposed in this study to point out Li-S battery features and provide guidance to users how to deal with them and possible results into standardization. The proposed test methodology is demonstrated for 3.4 Ah Li-S cells aged under different conditions....

  12. APT target-blanket fabrication development

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, D.L.

    1997-06-13

    Concepts for producing tritium in an accelerator were translated into hardware for engineering studies of tritium generation, heat transfer, and effects of proton-neutron flux on materials. Small-scale target- blanket assemblies were fabricated and material samples prepared for these performance tests. Blanket assemblies utilize composite aluminum-lead modules, the two primary materials of the blanket. Several approaches are being investigated to produce large-scale assemblies, developing fabrication and assembly methods for their commercial manufacture. Small-scale target-blanket assemblies, designed and fabricated at the Savannah River Site, were place in Los Alamos Neutron Science Center (LANSCE) for irradiation. They were subjected to neutron flux for nine months during 1996-97. Coincident with this test was the development of production methods for large- scale modules. Increasing module size presented challenges that required new methods to be developed for fabrication and assembly. After development, these methods were demonstrated by fabricating and assembling two production-scale modules.

  13. 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.

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  15. Accelerated lifetime testing methodology for lifetime estimation of Lithium-ion batteries used in augmented wind power plants

    DEFF Research Database (Denmark)

    Stroe, Daniel Ioan; Swierczynski, Maciej Jozef; Stan, Ana-Irina

    2013-01-01

    The development of lifetime estimation models for Lithium-ion battery cells, which are working under highly variable mission profiles characteristic for wind power plant applications, requires a lot of expenditures and time resources. Therefore, batteries have to be tested under accelerated...... lifetime ageing conditions. This paper presents a three-stage methodology used for accelerated lifetime testing of Lithium-ion batteries. The results obtained at the end of the accelerated ageing process can be used for the parametrization of a performance-degradation lifetime model. In the proposed...... methodology both calendar and cycling lifetime tests are considered since both components are influencing the lifetime of Lithium-ion batteries. The methodology proposes also a lifetime model verification stage, where Lithium-ion battery cells are tested at normal operating conditions using an application...

  16. Calculations of tritium breeding ratio and inventory distributions of FEB blanket

    International Nuclear Information System (INIS)

    Deng Baiquan

    2001-01-01

    Based on the design features of FEB reactor blanket, the tritium breeding ratio and tritium concentrations in liquid lithium of each breeding zone have been calculated after 10 days full power operation for outboard blanket and one day operation for inboard blanket. The comparisons with the results calculated by Monte-Carlo code MORSE-CGT are made. Meanwhile the inventory in beryllium multiplier after one-year full power operation has also been estimated. An important conclusion has been drew the thermal hydraulic design should be careful to guarantee the blanket temperature should not rise as high as 680 degree C

  17. 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)

  18. Safety test results of lithium-thionyl chloride wound-type cells

    Energy Technology Data Exchange (ETDEWEB)

    Vallin, D.; Broussely, M. (British Columbia Univ., Vancouver (Canada))

    1989-05-01

    Increase in the use of spirally-wound, lithium-thionyl chloride cells is currently limited because of unsafe incidents which have been reported during the early stage of development of this product. Today, it is believed that these cells are safe over a wide range of operating conditions if properly designed. The paper describes the external and internal SAFT design of Li-SOCl2LSH series cells, as well as the results of safety tests. 6 refs.

  19. Safety test results of lithium-thionyl chloride wound-type cells

    Science.gov (United States)

    Vallin, D.; Broussely, M.

    1989-05-01

    Increase in the use of spirally-wound, lithium-thionyl chloride cells is currently limited because of unsafe incidents which have been reported during the early stage of development of this product. Today, it is believed that these cells are safe over a wide range of operating conditions if properly designed. The paper describes the external and internal SAFT design of Li-SOCl2LSH series cells, as well as the results of safety tests.

  20. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-12-01

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

  1. 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

  2. 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

  3. Agmatine enhances the antidepressant-like effect of lithium in mouse forced swimming test through NMDA pathway.

    Science.gov (United States)

    Mohseni, Gholmreza; Ostadhadi, Sattar; Imran-Khan, Muhammad; Norouzi-Javidan, Abbas; Zolfaghari, Samira; Haddadi, Nazgol-Sadat; Dehpour, Ahmad-Reza

    2017-04-01

    Depression is one the world leading global burdens leading to various comorbidities. Lithium as a mainstay in the treatment of depression is still considered gold standard treatment. Similar to lithium another agent agmatine has also central protective role against depression. Since, both agmatine and lithium modulate various effects through interaction with NMDA receptor, therefore, in current study we aimed to investigate the synergistic antidepressant-like effect of agmatine with lithium in mouse force swimming test. Also to know whether if such effect is due to interaction with NMDA receptor. In our present study we found that when potent dose of lithium (30mg/kg) was administered, it significantly decreased the immobility time. Also, when subeffective dose of agmatine (0.01mg/kg) was coadministered with subeffective dose of lithium (3mg/kg), it potentiated the antidepressant-like effect of subeffective dose of lithium. For the involvement of NMDA receptor in such effect, we administered NMDA receptor antagonist MK-801 (0.05mg/kg) with a combination of subeffective dose of lithium (3mg/kg) and agmatine (0.001mg/kg). A significant antidepressant-like effect was observed. Furthermore, when subeffective dose (50 and 75mg/kg) of NMDA was given it inhibited the synergistic effect of agmatine (0.01mg/kg) with lithium (3mg/kg). Hence, our finding demonstrate that agmatine have synergistic effect with lithium which is mediated by NMDA receptor pathway. Copyright © 2017 Elsevier Masson SAS. All rights reserved.

  4. Accelerated Lifetime Testing Methodology for Lifetime Estimation of Lithium-ion Batteries used in Augmented Wind Power Plants

    DEFF Research Database (Denmark)

    Stroe, Daniel Ioan; Swierczynski, Maciej Jozef; Stan, Ana-Irina

    2014-01-01

    The development of lifetime estimation models for Lithium-ion battery cells, which are working under highly variable mission profiles characteristic for wind power plant applications, requires a lot of expenditures and time resources. Therefore, batteries have to be tested under accelerated...... lifetime ageing conditions. This paper presents a three-stage methodology used for accelerated lifetime testing of Lithium ion batteries. The results obtained at the end of the accelerated ageing process were used for the parametrization of a performance-degradation lifetime model, which is able to predict...... both the capacity fade and the power capability decrease of the selected Lithium-ion battery cells. In the proposed methodology both calendar and cycling lifetime tests were considered since both components are influencing the lifetime of Lithium-ion batteries. Furthermore, the proposed methodology...

  5. 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.).

  6. 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.)

  7. HHF test with 80x80x1 Be/Cu/SS Mock-ups for verifying the joining technology of the ITER blanket First Wall

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Hong, Bong Guen; Jeong, Yong Hwan; Park, Jeong Yong; Choi, Byung Kwon; Jung, Hyun Kyu

    2008-11-15

    Through the fabrication of the Cu/SS and Be/Cu joint specimens, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The HIP conditions (1050 .deg. C, 100 MPa 2 hr for Cu/SS, 580 .deg. C 100 MPa 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint include the microstructure observation of the interface with the examination of the elemental distribution, tension test, bend test, Charpy impact test and fracture toughness test. However, since the joint should be tested under the High Heat Flux (HHF) conditions like the ITER operation for verifying its joint integrity, several HHF tests were performed like the previous HHF test with the Cu/SS, Be/Cu, Be/Cu/SS Mock-ups. In the present study, the HHF test with Be/Cu/SS Mock-ups, which have 80 mm x 80 mm single Be tile and each material depths were kept to be the same as the ITER blanket FW. The Mock-ups fabricated with three kinds of interlayers such as Cr/Ti/Cu, Ti/Cr/Cu, Ti/Cu, which were different from the developed interlayer (Cr/Cu), total 6 Mock-ups were fabricated. Preliminary analysis were performed to decide the test conditions; they were tested with up to 2.5 MW/m2 of heat fluxes and 20 cycles for each Mock-up in a given heat flux. They were tested with JUDITH-1 at FZJ in Germany. During tests, all Mock-ups showed delamination or full detachment of Be tile and it can be concluded that the joints with these interlayers have a bad joining but it can be used as a good data for developing the Be/Cu joint with HIP.

  8. The second advanced lead lithium blanket concept using ODS steel as structural material and SiCf/SiC flow channel inserts as electrical and thermal insulators (Task PPA 2.5). Final report

    International Nuclear Information System (INIS)

    Norajitra, P.; Buehler, L.; Fischer, U.

    1999-12-01

    Preparatory work on the advanced dual coolant (A-DCL) blanket concept using SiC f /SiC flow channel inserts as electrical and thermal insulators has been carried out at the Forschungszentrum Karlsruhe in co-operation with CEA as a conceptual design proposal to the EU fusion power plant study planned to be launched in 2000 within the framework of the EU fusion programme with the main objective of specifying the characteristics of an attractive and viable commercial D-T fusion power plant. The basic principles and design characteristics of this A-DCL blanket concept are presented and its potential with regard to performance (neutron wall load, lifetime, availability) is discussed in this report. The results of this study show that the A-DCL blanket concept has a high potential for further development due to its high thermal efficiency and its simple concept solution. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  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. 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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  14. Compatibility of yttria (Y{sub 2}O{sub 3}) with liquid lithium

    Energy Technology Data Exchange (ETDEWEB)

    Mitsuyama, Takaaki; Yoneoka, Toshiaki; Terai, Takayuki; Tanaka, Satoru [Tokyo Univ. (Japan). Faculty of Engineering

    1996-10-01

    Compatibility of Y{sub 2}O{sub 3} sintered specimens with liquid lithium was tested at 773K. No configuration change was observed with a slight increase of thickness for 1419 hr. Lithium-yttrium complex oxide (LiYO{sub 2}) was formed on the surface, and the inner part changed to gray or black nonstoichiometric Y{sub 2}O{sub 3-X} with lower electrical resistibility. It is concluded that Y{sub 2}O{sub 3} has a possibility as a ceramic coating material for liquid blankets if it can be made into a dense coating on the surface of piping materials. (author)

  15. EU DEMO blanket concepts safety assessment. Final report of Working Group 6a of the Blanket Concept Selection Exercise

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Porfiri, T.

    1996-06-01

    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 blanket concepts under development. Two of them use lithium ceramics, the other two concepts employ an eutectic lead-lithium alloy (Pb-17Li) as breeder material. The two most promising concepts were to select in 1995 for further development. In order to prepare the selection, a Blanket Concept Selection Exercise (BCSE) has been inititated by the participating associations under the auspices of the European Commission. This BCSE has been performed in 14 working groups which, in a comparative evaluation of the four blanket concepts, addressed specific fields. The working group safety addressed the safety implications. This report describes the methodology adopted, the safety issues identified, their comparative evaluation for the four concepts, and the results and conclusions of the working group to be entered into the overall evaluation. There, the results from all 14 working groups have been combined to yield a final ranking as a basis for the selection. In summary, the safety assessment showed that the four European blanket concepts can be considered as equivalent in terms of the safety rating adopted, each concept, however, rendering safety concerns of different quality in different areas which are substantiated in this report. (orig.) [de

  16. Fabrication of modified lithium orthosilicate pebbles by addition of titania

    Energy Technology Data Exchange (ETDEWEB)

    Knitter, R., E-mail: regina.knitter@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-WPT), Karlsruhe, 76021 (Germany); Kolb, M.H.H.; Kaufmann, U. [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-WPT), Karlsruhe, 76021 (Germany); Goraieb, A.A. [Goraieb Versuchstechnik (GVT), Karlsruhe, 76227 (Germany)

    2013-11-15

    Highlights: ► Lithium orthosilicate pebbles with additions of titania were fabricated by a modified melt-based process. ► The fabricated pebbles exhibit a very fine-grained microstructure with lithium metatitanate as a secondary phase. ► Due to the addition of titanate, the crush load of the pebbles was significantly increased. ► The closed porosity was found to be slightly increased with increasing titanate content. -- Abstract: Lithium orthosilicate pebbles are one of the ceramic tritium breeder materials destined for the European solid breeder test blanket modules of ITER, the large-scale scientific experiment intended to prove the viability of fusion as an energy source, presently under construction in Cadarache, France. While the current reference material is fabricated by melt-spraying with 2.5 wt.% excess of silica, resulting in a two-phase material of lithium orthosilicate and metasilicate, a modified melt-based process was used to fabricate breeder pebbles with additions of titania in order to obtain pebbles with lithium metatitanate as a secondary phase. The fabricated two-phase pebbles exhibit a fine-grained microstructure and increased crush loads. The optimum titanate content has yet to be evaluated, nonetheless the pebbles may have the potential to combine the advantages of both lithium orthosilicate and metatitanate breeder ceramics.

  17. 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

  18. Approach to lithium burn-up effect in lithium ceramics

    International Nuclear Information System (INIS)

    Rasneur, B.

    1994-01-01

    The lithium burn-up in Li 2 ZrO 3 is simulated by removing lithium under Li 2 O form and trapping it in high specific surface area powder while heating during 15 days or 1 month at moderate temperature so that lithium mobility be large enough without causing any sintering neither of the specimens nor of the powder. In a first treatment at 775 deg C during 1 month. 30% of the lithium content could be removed inducing a lithium concentration gradient in the specimen and the formation of a lithium-free monoclinic ZrO 2 skin. Improvements led to similar results at 650 deg C and 600 deg C, the latter temperatures are closer to the operating temperature of the ceramic breeder blanket of a fusion reactor. (author) 4 refs.; 4 figs.; 1 tab

  19. International Space Station Lithium-Ion Main Battery Thermal Runaway Propagation Test

    Science.gov (United States)

    Dalton, Penni J.; North, Tim

    2017-01-01

    In 2010, the ISS Program began the development of Lithium-Ion (Li-Ion) batteries to replace the aging Ni-H2 batteries on the primary Electric Power System (EPS). After the Boeing 787 Li-Ion battery fires, the NASA Engineering and Safety Center (NESC) Power Technical Discipline Team was tasked by ISS to investigate the possibility of Thermal Runaway Propagation (TRP) in all Li-Ion batteries used on the ISS. As part of that investigation, NESC funded a TRP test of an ISS EPS non-flight Li-Ion battery. The test was performed at NASA White Sands Test Facility in October 2016. This paper will discuss the work leading up to the test, the design of the test article, and the test results.

  20. Empirical Modeling of Lithium-ion Batteries Based on Electrochemical Impedance Spectroscopy Tests

    International Nuclear Information System (INIS)

    Samadani, Ehsan; Farhad, Siamak; Scott, William; Mastali, Mehrdad; Gimenez, Leonardo E.; Fowler, Michael; Fraser, Roydon A.

    2015-01-01

    Highlights: • Two commercial Lithium-ion batteries are studied through HPPC and EIS tests. • An equivalent circuit model is developed for a range of operating conditions. • This model improves the current battery empirical models for vehicle applications • This model is proved to be efficient in terms of predicting HPPC test resistances. - ABSTRACT: An empirical model for commercial lithium-ion batteries is developed based on electrochemical impedance spectroscopy (EIS) tests. An equivalent circuit is established according to EIS test observations at various battery states of charge and temperatures. A Laplace transfer time based model is developed based on the circuit which can predict the battery operating output potential difference in battery electric and plug-in hybrid vehicles at various operating conditions. This model demonstrates up to 6% improvement compared to simple resistance and Thevenin models and is suitable for modeling and on-board controller purposes. Results also show that this model can be used to predict the battery internal resistance obtained from hybrid pulse power characterization (HPPC) tests to within 20 percent, making it suitable for low to medium fidelity powertrain design purposes. In total, this simple battery model can be employed as a real-time model in electrified vehicle battery management systems

  1. Startup of Experimental Lithium System

    International Nuclear Information System (INIS)

    McCauley, D.L.

    1980-06-01

    The Experimental Lithium System (ELS) is designed for full-scale testing of targets and other lithium system components for the Fusion Materials Irradiation Test (FMIT) Facility. The system also serves as a test bed for development of lithium purification and characterization equipment, provides experience in operation of large lithium systems, and helps guide FMIT design

  2. Centrifugation protocols: tests to determine optimal lithium heparin and citrate plasma sample quality.

    Science.gov (United States)

    Dimeski, Goce; Solano, Connie; Petroff, Mark K; Hynd, Matthew

    2011-05-01

    Currently, no clear guidelines exist for the most appropriate tests to determine sample quality from centrifugation protocols for plasma sample types with both lithium heparin in gel barrier tubes for biochemistry testing and citrate tubes for coagulation testing. Blood was collected from 14 participants in four lithium heparin and one serum tube with gel barrier. The plasma tubes were centrifuged at four different centrifuge settings and analysed for potassium (K(+)), lactate dehydrogenase (LD), glucose and phosphorus (Pi) at zero time, poststorage at six hours at 21 °C and six days at 2-8°C. At the same time, three citrate tubes were collected and centrifuged at three different centrifuge settings and analysed immediately for prothrombin time/international normalized ratio, activated partial thromboplastin time, derived fibrinogen and surface-activated clotting time (SACT). The biochemistry analytes indicate plasma is less stable than serum. Plasma sample quality is higher with longer centrifugation time, and much higher g force. Blood cells present in the plasma lyse with time or are damaged when transferred in the reaction vessels, causing an increase in the K(+), LD and Pi above outlined limits. The cells remain active and consume glucose even in cold storage. The SACT is the only coagulation parameter that was affected by platelets >10 × 10(9)/L in the citrate plasma. In addition to the platelet count, a limited but sensitive number of assays (K(+), LD, glucose and Pi for biochemistry, and SACT for coagulation) can be used to determine appropriate centrifuge settings to consistently obtain the highest quality lithium heparin and citrate plasma samples. The findings will aid laboratories to balance the need to provide the most accurate results in the best turnaround time.

  3. Targeting as the basis for pre-test market of lithium-ion battery

    Science.gov (United States)

    Yuniaristanto, Zakaria, R.; Saputri, V. H. L.; Sutopo, W.; Kadir, E. A.

    2017-11-01

    This article discusses about market segmentation and targeting as a first step in pre-test market of a new technology. The benefits of targeting towards pre-test market are pre-test market can be conducted to focus on selected target markets so there is no bias during the pre-test market. In determining the target market then do some surveys to identify the state of market in the future, so that the marketing process is not misplaced. Lithium ion battery which is commercialized through start-up companies is the case study. This start-up companies must be able to respond the changes and bring in customers as well as maintain them so that companies can survive and evolve to achieve its objectives. The research aims to determine market segments and target market effectively. Marketing strategy (segmentation and targeting) is used to make questionnaire and cluster analysis in data processing. Respondents were selected by purposive sampling and have obtained data as many as 80 samples. As the results study, there are three segments for lithium ion battery with their own distinguished characteristics and there are two segments that can be used as the target market for the company.

  4. Lithium-ion batteries for hearing aid applications. II. Pulse discharge and safety tests

    Science.gov (United States)

    Passerini, S.; Coustier, F.; Owens, B. B.

    Rechargeable lithium-ion batteries were designed to meet the power requirements of hearing aid devices (HADs). The batteries were designed in a 312-button cell size, compatible with existing hearing aids. The batteries were tested to evaluate the design and the electrochemical performance, as they relate to a typical hearing aid application. The present report covers the pulse capabilities, cycle life and preliminary safety tests. The results are compared with other battery chemistries: secondary lithium-alloy and nickel-metal hydride batteries and primary Zn-air batteries. The cell AC impedance was stable over the frequency range between 1 and 50 kHz, ranging between 5 Ω at the higher frequency and 12 Ω at the lower extreme. Pulse tests were consistent with these values, as the cells were capable of providing a series of 100 mA pulses of 10-s duration. The safety tests suggest that the design is intrinsically safe with respect to the most common types of abuse conditions.

  5. Measurement and Analysis of the Neutron and Gamma-Ray Flux Spectra in a Neutronics Mock-Up of the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Seidel, K.; Freiesleben, H.; Poenitz, E.; Klix, A.; Unholzer, S.; Batistoni, P.; Fischer, U.; Leichtle, D.

    2006-01-01

    The nuclear parameters of a breeding blanket, such as tritium production rate, nuclear heating, activation and dose rate, are calculated by integral folding of an energy dependent cross section (or coefficient) with the neutron (or gamma-ray) flux energy spectra. The uncertainties of the designed parameters are determined by the uncertainties of both the cross section data and the flux spectra obtained by transport calculations. Also the analysis of possible discrepancies between measured and calculated integral nuclear parameter represents a two-step procedure. First, the energy region and the amount of flux discrepancies has to be found out and second, the cross section data have to be checked. To this end, neutron and gamma-ray flux spectra in a mock-up of the EU Helium-Cooled Pebble Bed (HCPB) breeder Test Blanket Module (TBM), irradiated with 14 MeV neutrons, were measured and analysed by means of Monte Carlo transport calculations. The flux spectra were determined for the energy ranges that are relevant for the most important nuclear parameters of the TBM, which are the tritium production rate and the shielding capability. The fast neutron flux which determines the tritium production on 7 Li and dominates the shield design was measured by the pulse-height distribution obtained from an organic liquid scintillation detector. Simultaneously, the gamma-ray flux spectra were measured. The neutron flux at lower energies, down to thermal, which determines the tritium production on 6 Li, was measured with time-of-arrival spectroscopy. For this purpose, the TUD neutron generator was operated in pulsed mode (pulse width 10 μs, frequency 1 kHz) and the neutrons arriving at a 3 He proportional counter in the mock-up were recorded as a function of time after the source neutron pulse. The spectral distributions for the two positions in the mock-up, where measurements were carried out, were calculated with the Monte Carlo code MCNP, version 5, and nuclear data from the

  6. 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

  7. Cyclic performance tests of Sn/MWCNT composite lithium ion battery anodes at different temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Tocoglu, U., E-mail: utocoglu@sakarya.edu.tr; Cevher, O.; Akbulut, H. [Sakarya University, Engineering Faculty, Department of Metallurgical and Materials Engineering, Esentepe Campus 54187 (Turkey)

    2016-04-21

    In this study tin-multi walled carbon nanotube (Sn-MWCNT) lithium ion battery anodes were produced and their electrochemical galvanostatic charge/discharge tests were conducted at various (25 °C, 35 °C, 50 °C) temperatures to determine the cyclic behaviors of anode at different temperatures. Anodes were produced via vacuum filtration and DC magnetron sputtering technique. Tin was sputtered onto buckypapers to form composite structure of anodes. SEM analysis was conducted to determine morphology of buckypapers and Sn-MWCNT composite anodes. Structural and phase analyses were conducted via X-ray diffraction and Raman Spectroscopy technique. CR2016 coin cells were assembled for electrochemical tests. Cyclic voltammetry test were carried out to determine the reversibility of reactions between anodes and reference electrode between 0.01-2.0 V potential window. Galvanostatic charge/discharge tests were performed to determine cycle performance of anodes at different temperatures.

  8. Numerical Analysis for Heat transfer characteristic of Helium cooling system in Helium cooled ceramic reflector Test Module Blanket (HCCR-TBM)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Lee, Dong Won; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Kim, Suk Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The main objectives of ITER project can be summarized into three types as follows - Plasma operation for a long time - Large tokamak device technology - Test blanket module (TBM) installation and verification The thermal-hydraulic analysis was performed in the He cooling channel in the BZ region of the HCCR TBM. The maximum temperature in the breeder material is equal to the limit temperature in the present design cooling channel. Nuclear fusion energy has advantage in terms of safety, resource availability, cost and waste management. There is not enough experimental results about the fusion reactor due to the severe experiments restrictions like vacuum environment, plasma production and significant nuclear heating at the same time. Much research and time is required for the commercial fusion reactor. For technical verification against the commercialization of fusion reactor, 7 countries which are EU, USA, Japan, Russia, China, India, and South Korea are building an ITER in the south of France. New designed cooling channels were proposed to improve the cooling performance. The swirl flow accelerates the mixture flow in the channels.

  9. 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.

  10. 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.

  11. 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)

  12. 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

  13. Strain-specific battery of tests for domains of mania: effects of valproate, lithium and imipramine

    Directory of Open Access Journals (Sweden)

    Shlomit Flaisher-Grinberg

    2010-04-01

    Full Text Available The lack of efficient animal models for bipolar disorder (BPD, especially for the manic pole, is a major factor hindering the research of its pathophysiology and the development of improved drug treatments. The present study was designed to identify an appropriate mouse strain for modeling some behavioral domains of mania and to evaluate the effects of drugs using this strain. The study compared the behavior of four strains: Black Swiss, C57Bl/6, CBA/J and A/J mice in a battery of tests that included spontaneous activity; sweet solution preference; light/dark box; resident-intruder; forced-swim and amphetamine-induced hyperactivity. Based on the ‘manic-like’ behavior demonstrated by the Black Swiss strain, the study evaluated the effects of the mood stabilizers valproate and lithium and of the antidepressant imipramine in the same tests using this strain. Results indicated that lithium and valproate attenuate the ‘manic-like’ behavior of Black Swiss mice whereas imipramine had no effects. These findings suggest that Black Swiss mice might be a good choice for modeling several domains of mania and distinguishing the effects of drugs on these specific domains. However, the relevance of the behavioral phenotype of Black Swiss mice to the biology of BPD is unknown at this time and future studies will investigate molecular differences between Black Swiss mice and other strains and asess the interaction between strain and mood stabilizing treatment.

  14. Safety tests of bobbin-type lithium-thionyl chloride D-cells

    Energy Technology Data Exchange (ETDEWEB)

    Okamura, Yasuyuki; Mizutani, Minoru

    1987-12-25

    Safety test was made focusing on the possibilities of explosion and leakage of bobbin-type lithium-thionyl chloride D-cells. The result indicates that there is no abnormality including explosion even at the vigorous testing including the large current charging test. Meanwhile, since the instability of battery at large current became clear, it was reconfirmed that a protective element against charging would be required. A lost of irritating, toxic, corrosive electrolyte leaking from crushed and drilled batteries is apt to injure the human respiratory organs in a closed space. It is necessary to surely implement the protective measure against abnormal temperature increase by using the battery with care taken not to throw it into flame, keep it away from a heat source as well as charge it. In addition, the protective element against charging such as charging protection diode and the protection against the destruction by external force are required. (12 figs, 2 tabs, 3 refs)

  15. Novel Field Test Equipment for Lithium-Ion Batteries in Hybrid Electrical Vehicle Applications

    Directory of Open Access Journals (Sweden)

    Goran Lindbergh

    2011-04-01

    Full Text Available Lifetime testing of batteries for hybrid-electrical vehicles (HEV is usually performed in the lab, either at the cell, module or battery pack level. Complementary field tests of battery packs in vehicles are also often performed. There are, however, difficulties related to field testing of battery-packs. Some examples are cost issues and the complexity of continuously collecting battery performance data, such as capacity fade and impedance increase. In this paper, a novel field test equipment designed primarily for lithium-ion battery cell testing is presented. This equipment is intended to be used on conventional vehicles, not hybrid vehicles, as a cheaper and faster field testing method for batteries, compared to full scale HEV testing. The equipment emulates an HEV environment for the tested battery cell by using real time vehicle sensor information and the existing starter battery as load and source. In addition to the emulated battery cycling, periodical capacity and pulse testing capability are implemented as well. This paper begins with presenting some background information about hybrid electrical vehicles and describing the limitations with today’s HEV battery testing. Furthermore, the functionality of the test equipment is described in detail and, finally, results from verification of the equipment are presented and discussed.

  16. Remote handling demonstration of ITER blanket module replacement

    International Nuclear Information System (INIS)

    Kakudate, S.; Nakahira, M.; Oka, K.; Taguchi, K.; Obara, K.; Tada, E.; Shibanuma, K.; Tesini, A.; Haange, R.; Maisonnier, D.

    2001-01-01

    In ITER, the in-vessel components such as blanket are to be maintained or replaced remotely since they will be activated by 14 MeV neutrons, and a complete exchange of shielding blanket with breeding blanket is foreseen after the Basic Performance Phase. The blanket is segmented into about seven hundred modules to facilitate remote maintainability and allow individual module replacement. For this, the remote handing equipment for blanket maintenance is required to handle a module with a dead weight of about 4 tonne within a positioning accuracy of a few mm under intense gamma radiation. According to the ITER R and D program, a rail-mounted vehicle manipulator system was developed and the basic feasibility of this system was verified through prototype testing. Following this, development of full-scale remote handling equipment has been conducted as one of the ITER Seven R and D Projects aiming at a remote handling demonstration of the ITER blanket. As a result, the Blanket Test Platform (BTP) composed of the full-scale remote handling equipment has been completed and the first integrated performance test in March 1998 has shown that the fabricate remote handling equipment satisfies the main requirements of ITER blanket maintenance. (author)

  17. Evaluation of US demo helium-cooled blanket options

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  18. 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

  19. FELIX: construction and testing of a facility to study electromagnetic effects for first wall, blanket, and shield systems

    International Nuclear Information System (INIS)

    Praeg, W.F.; Turner, L.R.; Biggs, J.A.; Knott, M.J.; Lari, R.J.; McGhee, D.G.; Wehrle, R.B.

    1983-01-01

    An experimental test facility for the study of electromagnetic effects in the FWBS systems of fusion reactors has been constructed over the past 1-1/2 years at Argonne National Laboratory (ANL). In a test volume of 0.76 m 3 a vertical pulsed 0.5 T dipole field (B < 50 T/s) is perpendicular to a 1 T solenoid field. Power supplies of 2.75 MW and 5.5 MW and a solid state switch rated 13 kV, 13.1 kA (170 MW) control the pulsed magnetic fields. The total stored energy in the coils is 2.13 MJ. The coils are designed for a future upgrade to 4 T or the solenoid and 1 T for the dipole field (a total of 23.7 MJ). This paper describes the design and construction features of the facility. These include the power supplies, the solid state switches, winding and impregnation of large dipole saddle coils, control of the magnetic forces, computer control of FELIX and of experimental data acquisition and analysis, and an initial experimental test setup to analyze the eddy current distribution in a flat disk

  20. FELIX: Construction and testing of a facility to study electromagnetic effects for First Wall, Blanket, and Shield systems

    International Nuclear Information System (INIS)

    Praeg, W.F.; Biggs, J.; Knott, M.J.; Lari, R.J.; McGhee, D.G.; Turner, L.R.; Wehrle, R.

    1983-01-01

    An experimental test facility for the study of electromagnetic effects in the FWBS systems of fusion reactors has been constructed over the past 2-1/2 years at Argonne National Laboratory (ANL). In a test volume of 0.76 m 3 a vertical pulsed 0.5 T dipole field (B < 50 T/s) is perpendicular to a 1 T solenoid field. Power supplies of 2.75 MW and 5.5 MW and a solid state switch rated 13 kV, 13.1 kA (170 MW) control the pulsed magnetic fields. The total stored energy in the coils is 2.13 MJ. The coils are designed for a future upgrade to 4 T for the solenoid and 1 T for the dipole field (a total of 23.7 MJ). This paper describes the design and construction features of the facility. These include the power supplies, the solid state switches, winding and impregnation of large dipole saddle coils, control of the magnetic forces, computer control of FELIX and of experimental data acquisition and analysis, and an initial experimental test setup to analyze the eddy current distribution in a flat disk

  1. 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

  2. Performance Model for High-Power Lithium Titanate Oxide Batteries based on Extended Characterization Tests

    DEFF Research Database (Denmark)

    Stroe, Ana-Irina; Swierczynski, Maciej Jozef; Stroe, Daniel Ioan

    2015-01-01

    Lithium-ion (Li-ion) batteries are found nowadays not only in portable/consumer electronics but also in more power demanding applications, such as stationary renewable energy storage, automotive and back-up power supply, because of their superior characteristics in comparison to other energy...... storage technologies. Nevertheless, prior to be used in any of the aforementioned application, a Li-ion battery cell must be intensively characterized and its behavior needs to be understood. This can be realized by performing extended laboratory characterization tests and developing Li-ion battery...... performance models. Furthermore, accurate performance models are necessary in order to analyze the behavior of the battery cell under different mission profiles, by simulation; thus, avoiding time and cost demanding real life tests. This paper presents the development and the parametrization of a performance...

  3. Preliminary data from lithium hydride ablation tests conducted by NASA, Ames Research Center

    International Nuclear Information System (INIS)

    Elliott, R.D.

    1970-01-01

    A series of ablation tests of lithium hydride has been made by NASA-Ames in one of their high-enthalpy arc-heated wind tunnels. Two-inch diameter cylindrical samples of the hydride, supplied by A. I., were subjected to heating on their ends for time periods up to 10 seconds. After each test, the amount of material removed from each sample was measured. The rates of loss of material were correlated with the heat input rates in terms of a heat of ablation, which ranged from 2100 to 3500 Btu/lb. The higher values were obtained when the hydride contained a matrix such as steel honeycomb of steel wool. (U.S.)

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  5. The ITER EC H&CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

    NARCIS (Netherlands)

    Gessner, R.; Aiello, G.; Grossetti, G.; Meier, A.; Ronden, D.; Spaeh, P.; Scherer, T.; Schreck, S.; Strauss, D.; Vaccaro, A.

    2013-01-01

    The final design of the structural system for the ITER EC H&CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield

  6. The transpiration cooled first wall and blanket concept

    International Nuclear Information System (INIS)

    Barleon, Leopold; Wong, Clement

    2002-01-01

    To achieve high thermal performance at high power density the EVOLVE concept was investigated under the APEX program. The EVOLVE W-alloy first wall and blanket concept proposes to use transpiration cooling of the first wall and boiling or vaporizing lithium (Li) in the blanket zone. Critical issues of this concept are: the Magnetohydrodynamic (MHD) pressure losses of the Li circuit, the evaporation through a capillary structure and the needed superheating of the Li at the first wall and blanket zones. Application of the transpiration concept to the blanket region results in the integrated transpiration cooling concept (ITCC) with either toroidal or poloidal first wall channels. For both orientations the routing of the liquid Li and the Li vapor has been modeled and the corresponding pressure losses have been calculated by varying the width of the supplying slot and the capillary diameter. The concept works when the sum of the active and passive pumping head is higher than the total system pressure losses and when the temperature at the inner side of the first wall does not override the superheating limit of the coolant. This cooling concept has been extended to the divertor design, and the removal of a surface heat flux of up to 10 MW/m 2 appears to be possible, but this paper will focus on the transpiration cooled first wall and blanket concept assessment

  7. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1983-08-01

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

  8. Breeding blanket development. Tritium release from breeder

    International Nuclear Information System (INIS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nagao, Yoshiharu

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    1984-09-01

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

  13. Acquisition of Co metal from spent lithium-ion battery using emulsion liquid membrane technology and emulsion stability test

    Science.gov (United States)

    Yuliusman; Wulandari, P. T.; Amiliana, R. A.; Huda, M.; Kusumadewi, F. A.

    2018-03-01

    Lithium-ion batteries are the most common type to be used as energy source in mobile phone. The amount of lithium-ion battery wastes is approximated by 200 – 500 ton/year. In one lithium-ion battery, there are 5 – 20% of cobalt metal, depend on the manufacturer. One of the way to recover a valuable metal from waste is leaching process then continued with extraction, which is the aim of this study. Spent lithium-ion batteries will be characterized with EDX and AAS, the result will show the amount of cobalt metal with form of LiCoO2 in the cathode. Hydrochloric acid concentration used is 4 M, temperature 80°C, and reaction time 1 hour. This study will discuss the emulsion stability test on emulsion liquid membrane. The purpose of emulsion stability test in this study was to determine optimum concentration of surfactant and extractant to produce a stable emulsion. Surfactant and extractant used were SPAN 80 and Cyanex 272 respectively with both concentrations varied. Membrane and feed phase ratios used in this experiment was 1 : 2. The optimum results of this study were SPAN 80 concentrations of 10% w/v and Cyanex 272 0.7 M.

  14. Development of filler wires for welding of reduced activation ferritic martensitic steel for India's test blanket module of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, G.; Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2010-07-01

    Indigenous development of reduced activation ferritic-martensitic (RAFM) steel has become necessary for India as a participant in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFM steel is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFM steel filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFM steel. The purpose of this study is to develop filler wires that can be directly used for both gas tungsten arc welding (GTAW) and for narrow-gap gas tungsten arc welding (NG-GTAW) that reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser-MIG welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using GTAW process at various heat inputs with a preheat temperature of 250 C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some amount of delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimized to qualify the filler wires without the presence of delta-ferrite in the weld metal and with optimized mechanical properties. Results showed that the weld metals are free from delta-ferrite. Tensile properties at ambient temperature and at 500 C are well above the specified values, and are much higher than the base metal values. Ductile Brittle Transition Temperature (DBTT) has been evaluated as -81 C based on the 68 J criteria. The present study highlights the basis and methodology

  15. Development of filler wires for welding of reduced activation ferritic martenstic steel for India's test blanket module of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, G., E-mail: gsrini@igcar.gov.in [Materials Technology Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamilnadu (India); Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K. [Materials Technology Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamilnadu (India)

    2011-06-15

    Highlights: > Weld microstructure produced by RAFMS filler wires are free from delta ferrite. > Cooling rates of by weld thermal cycles influences the presence of delta ferrite. > Weld parameters modified with higher pre heat temperature and high heat input. > PWHT optimized based on correlation of hardness between base and weld metals. > Optimised mechanical properties achieved by proper tempering of the martensite. - Abstract: Indigenous development of reduced activation ferritic martensitic steel (RAFMS) has become mandatory to India to participate in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFMS is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFMS filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFMS. Purpose of this study is to develop filler wires that can be directly used for both tungsten inert gas welding (TIG) and narrow gap tungsten inert gas welding (NG-TIG), which reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, autogenous welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using TIG process at various heat inputs with a preheat temperature of 250 deg. C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimised to qualify the filler wires without the presence of delta-ferrite in the weld

  16. The ITER EC H and CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

    Energy Technology Data Exchange (ETDEWEB)

    Gessner, Robby, E-mail: robby.gessner@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano; Grossetti, Giovanni; Meier, Andreas [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Ronden, Dennis [DIFFER – Dutch Institute for Fundamental Energy Physics, P.O. Box 1207, NL-3430 BE Nieuwegein (Netherlands); Spaeh, Peter; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2013-10-15

    Highlights: ► The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint. ► The bolts were designed as “captive” in order to avoid their accidental removal from the joint. ► The bolted flange connection using two sets of 15 captive bolts (M22 × 2) placed along the sides. ► The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. -- Abstract: The final design of the structural system for the ITER EC H and CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield Module with special perspective on Remote Handling capability. The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint conceived so that in the Hot Cell Facility, RH maintenance can be performed on internal components. The joint must be capable to resist very high Electro-Magnetic loads from disruptions, while it has to sustain substantial thermal cycling during operation. Thus the need for a rigid and reliable design is essential. Beside the set of pre-stressed bolts the flanges were therefore equipped with additional shear keys to divert radial moments away from the bolts. Main focus of the work performed was the mechanical design of the joint and the assessment of the structural integrity with respect to the loads applied and its capability for maintenance by RH procedures. To fulfill a major aspect of the RH requirements, the bolts were designed as “captive” in order to avoid their accidental removal from the joint. The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. The final approval phase of

  17. The ITER EC H and CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

    International Nuclear Information System (INIS)

    Gessner, Robby; Aiello, Gaetano; Grossetti, Giovanni; Meier, Andreas; Ronden, Dennis; Spaeh, Peter; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro

    2013-01-01

    Highlights: ► The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint. ► The bolts were designed as “captive” in order to avoid their accidental removal from the joint. ► The bolted flange connection using two sets of 15 captive bolts (M22 × 2) placed along the sides. ► The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. -- Abstract: The final design of the structural system for the ITER EC H and CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield Module with special perspective on Remote Handling capability. The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint conceived so that in the Hot Cell Facility, RH maintenance can be performed on internal components. The joint must be capable to resist very high Electro-Magnetic loads from disruptions, while it has to sustain substantial thermal cycling during operation. Thus the need for a rigid and reliable design is essential. Beside the set of pre-stressed bolts the flanges were therefore equipped with additional shear keys to divert radial moments away from the bolts. Main focus of the work performed was the mechanical design of the joint and the assessment of the structural integrity with respect to the loads applied and its capability for maintenance by RH procedures. To fulfill a major aspect of the RH requirements, the bolts were designed as “captive” in order to avoid their accidental removal from the joint. The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. The final approval phase of

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  19. Experimental lithium system experience

    International Nuclear Information System (INIS)

    Atwood, J.M.; Berg, J.D.; Kolowith, R.; Miller, W.C.

    1984-01-01

    The Experimental Lithium System is a test loop built to support design and operation of the Fusion Materials Irradiation Test Facility. ELS has achieved over 15,000 hours of safe and reliable operation. An extensive test program has demonstrated satisfactory performance of the system components, including an electromagnetic pump, lithium jet target, and vacuum system. Data on materials corrosion and behavior of lithium impurities are also presented. (author)

  20. Vibration Durability Testing of Nickel Cobalt Aluminum Oxide (NCA Lithium-Ion 18650 Battery Cells

    Directory of Open Access Journals (Sweden)

    James Michael Hooper

    2016-04-01

    Full Text Available This paper outlines a study undertaken to determine if the electrical performance of Nickel Cobalt Aluminum Oxide (NCA 3.1 Ah 18650 battery cells can be degraded by road induced vibration typical of an electric vehicle (EV application. This study investigates if a particular cell orientation within the battery assembly can result in different levels of cell degradation. The 18650 cells were evaluated in accordance with Society of Automotive Engineers (SAE J2380 standard. This vibration test is synthesized to represent 100,000 miles of North American customer operation at the 90th percentile. This study identified that both the electrical performance and the mechanical properties of the NCA lithium-ion cells were relatively unaffected when exposed to vibration energy that is commensurate with a typical vehicle life. Minor changes observed in the cell’s electrical characteristics were deemed not to be statistically significant and more likely attributable to laboratory conditions during cell testing and storage. The same conclusion was found, irrespective of cell orientation during the test.

  1. General directions and recently test modelling results of lithium capillary-pore systems as plasma facing components for tokamak-reactor

    International Nuclear Information System (INIS)

    Evtikhin, V.A.; Lyublinski, I.E.; Vertkov, A.V.; Azizov, E.A.; Mirnov, S.V.; Lazaret, V.B.; Safronov, V.M.

    2003-01-01

    Full text: At present the most promising principal solution of the divertor problem appears to be the use of liquid metals and primarily of lithium Capillary-Pore Systems (CPS) as of plasma facing material. A solid CPS filled with liquid lithium will have high resistance to surface and volume damage because of neutron radiation effects, melting, splashing and thermal stress induced cracking in steady state and during plasma transitions (disruptions, ELMs, VDEs, runaways) to provide the normal operation of divertor target plates and first wall protection elements. These materials would not be the sources of impurities inducing the raise of Z eff and they will not be collected as dust in the divertor area and in ducts. The key directions of experimental investigation of lithium CPS behaviour in first wall and divertor operation simulating conditions are considered. Experiments with lithium CPS in plasma disruption simulation conditions on the hydrogen plasma accelerator MK-200UG (∼10-15 MJ/m 2 , ∼50 μs) have been performed. Shielding lithium plasma layer formation and high stability of these systems have been shown. The new lithium limiter with a thermal regulation system tests on up graded T-11M tokamak (plasma current up to 100 kA, pulse length ∼0.3 s) have been performed. Sorption and desorption of plasma-forming gas, lithium emission into discharge, lithium erosion, limiter deposited power are investigated in this tests

  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. A 20 kw beam-on-target test of a high-power liquid lithium target for RIA

    International Nuclear Information System (INIS)

    Reed, Claude B.; Nolen, Jerry A.; Specht, James R.; Novick, Vincent J.; Plotkin, Perry

    2004-01-01

    The high-power heavy-ion beams produced by the Rare Isotope Accelerator (RIA) driver linac have large energy deposition density in solids and in many cases no solid materials would survive the full beam power. Liquid lithium technology has been proposed to solve this problem in RIA. Specifically, a windowless target for the production of radioactive ions via fragmentation, consisting of a jet of about 3 cm thickness of flowing liquid lithium, exposed to the beamline vacuum [1,2] is being developed. To demonstrate that power densities equivalent to a 200-kW RIA uranium beam, deposited in the first 4 mm of a flowing lithium jet, can be handled by the windowless target design, a high power 1 MeV Dynamitron was leased and a test stand prepared to demonstrate the target's capability of absorbing and carrying away a 20kW heat load without disrupting either the 5 mm x 10 mm flowing lithium jet target or the beam line vacuum

  4. 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.)

  5. 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)

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

    International Nuclear Information System (INIS)

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Norajitra, P.

    1995-06-01

    Two different breeding blanket concepts are being elaborated at Forschungszentrum Karlsruhe within the framework of the DEMO breeding blanket development, the concept of a helium cooled solid breeding blanket and the concept of a self-cooled liquid metal blanket. The breeder material used in the first concept is Li 4 SiO 4 as a pebble bed arranged separate from the beryllium pebble bed, which serves as multiplier. The breeder material zone is cooled by several toroidally-radially configurated helium cooling plates which, at the same time, act as reinforcements of the blanket structures. In the liquid metal blanket concept lead-lithium is used both as the breeder material and the coolant. It flows at low velocity in poloidal direction downwards and back in the blanket front zone. In both concepts the First Wall is cooled by helium gas. This report deals with the thermohydraulics design and thermomechanics analysis of the two blanket concepts. The performance data derived from the Monte-Carlo computations serve as a basis for the design calculations. The coolant inlet and outlet temperatures are chosen with the design criteria and the economics aspects taken into account. Uniform temperature distribution in the blanket structures can be achieved by suitable branching and routing of the coolant flows which contributes to reducing decisively the thermal stress. The computations were made using the ABAQUS computer code. The results obtained of the stresses have been evaluated using the ASME code. It can be demonstrated that all maximum values of temperature and stress are below the admissible limit. (orig.) [de

  8. 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

  9. Test and Analysis of Sub-Components of Aluminum-Lithium Alloy Cylinders

    Science.gov (United States)

    Haynie, Waddy T.; Chunchu, Prasad B.; Satyanarayana, Arunkumar; Hilburger, Mark W.; Smith, Russell W.

    2012-01-01

    Integrally machined blade-stiffened panels subjected to an axial compressive load were tested and analyzed to observe the buckling, crippling, and postcrippling response of the panels. The panels were fabricated from aluminum-lithium alloys 2195 and 2050, and both alloys have reduced material properties in the short transverse material direction. The tests were designed to capture a failure mode characterized by the stiffener separating from the panel in the postbuckling range. This failure mode is attributed to the reduced properties in the short transverse direction. Full-field measurements of displacements and strains using three-dimensional digital image correlation systems and local measurements using strain gages were used to capture the deformation of the panel leading up to the failure of the panel for specimens fabricated from 2195. High-speed cameras were used to capture the initiation of the failure. Finite element models were developed using an isotropic strain-hardening material model. Good agreement was observed between the measured and predicted responses for both alloys.

  10. Determination of lithium in organic matrix by potentiometric titration using fluoride ion selective electrode

    International Nuclear Information System (INIS)

    Govindan, R.; Alamelu, D.; Shah, Raju; Aggarwal, S.K.

    2008-01-01

    A method has been developed for the determination of lithium (Li) present in organic matrix containing hexa methylene tetramine (HMTA) and urea used in the sol-gel process for preparing lithium titanate microspheres, using fluoride ion selective electrode and potentiometric end point. Lithium is present in the wash solutions of the Sol-Gel process employed for the preparation of lithium titanate microspheres, proposed to be used in TBM (Test Blanket Module) of International Thermonuclear Experimental Reactor (ITER) project. Methods such as ICP-AES, AAS etc. used in aqueous solutions cannot be employed directly for lithium determination in organic matrix containing hexa methylene tetramine (HMTA), urea, NH 4 NO 3 , NH 4 Cl etc. A potentiometric method using a combination fluoride ion selective electrode for end point detection was developed and has been employed for lithium determination in the process streams from sol-gel process. The method is simple and rapid and an accuracy of about 0.5 % was achieved for the determination of Li in the range of 1 to 20 mg. The method is based on the complexation of Li by adding a known excess of NH 4 F solution, followed by potentiometric end point detection using fluoride ISE

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

    International Nuclear Information System (INIS)

    Bullis, R.; Clarkson, I.

    1983-01-01

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

  12. 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.

  13. Present development status of EUROFER and ODS-EUROFER for application in blanket concepts

    Energy Technology Data Exchange (ETDEWEB)

    Lindau, R. [Forschungszentrum Karlsruhe, Institute for Materials Research I, P.O. Box 3640, 76021 Karlsruhe (Germany)]. E-mail: rainer.lindau@imf.fzk.de; Moeslang, A. [Forschungszentrum Karlsruhe, Institute for Materials Research I, P.O. Box 3640, 76021 Karlsruhe (Germany); Rieth, M. [Forschungszentrum Karlsruhe, Institute for Materials Research I, P.O. Box 3640, 76021 Karlsruhe (Germany); Klimiankou, M. [Forschungszentrum Karlsruhe, Institute for Materials Research I, P.O. Box 3640, 76021 Karlsruhe (Germany); Materna-Morris, E. [Forschungszentrum Karlsruhe, Institute for Materials Research I, P.O. Box 3640, 76021 Karlsruhe (Germany); Alamo, A. [CEA-Saclay, SRMA/SMPX, 91191 Gif-sur-Yvette Cedex (France); Tavassoli, A.-A. F. [CEA-Saclay, SRMA/SMPX, 91191 Gif-sur-Yvette Cedex (France); Cayron, C. [CEA-Grenoble, DRT/DTEN/SMP/LS2M, 17, rue des Martyrs, 38054 Grenoble Cedex 9 (France); Lancha, A.-M. [CIEMAT, Avda. Complutense no. 22, 28040 Madrid (Spain); Fernandez, P. [CIEMAT, Avda. Complutense no. 22, 28040 Madrid (Spain); Baluc, N. [CRPP-EPFL, 5232 Villigen PSI (Switzerland); Schaeublin, R. [CRPP-EPFL, 5232 Villigen PSI (Switzerland); Diegele, E. [EFDA Close Support Unit, Boltzmannstr. 2, 85748 Garching (Germany); Filacchioni, G. [ENEA CR Casaccia, Via Anguillarese 301, 00100 S. Maria di Galeria, Rome (Italy); Rensman, J.W. [NRG, MM and I, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands); Schaaf, B. van der [NRG, MM and I, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands); Lucon, E. [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Dietz, W. [MECS, Schoenenborner Weg 15, 51789 Lindlar (Germany)

    2005-11-15

    Within the European Union, the two major breeding blanket concepts presently being developed are the helium cooled pebble bed (HCPB), and the helium cooled lithium lead (HCLL) blankets. For both concepts, different conceptual designs are being discussed with temperature windows in the range 250-550 deg. C for conservative approaches based on reduced activation ferritic-martensitic (RAFM) steels, and in the range 250-650 deg. C for more advanced versions, taking into account oxide dispersion strengthened (ODS) steels. As a final result of a systematic development of RAFM-steels in Europe, the 9% CrWVTa alloy EUROFER was specified and produced in an industrial scale with a variety of product forms. A large characterisation program is being performed including irradiation in materials test reactors between 60 and 450 deg. C ({<=}15 dpa), and in a fast breeder reactor at 330 deg. C up to 30 dpa. EUROFER is resistant to high temperature ageing, and the existing creep-rupture data ({approx}30,000 h, 450-600 deg. C) indicate long-term stability and predictability. The ODS variant of EUROFER shows superior tensile and creep properties compared to EUROFER. Applying a new production route has diminished the problem of lower ductility and inferior impact properties. A reliable joining technique for ODS and RAFM steels employing diffusion welding was successfully developed.

  14. 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.

  15. Safety tests of spiral-type lithium-thionyl chloride D-cells

    Energy Technology Data Exchange (ETDEWEB)

    Uno, Kyoji; Mizutani, Minoru (Japan Storage Battery Co., Ltd., Kyoto)

    1989-12-25

    The spiral-type Lithium-Thionyl Chloride D-cell 3360H has no problem at all on safety under normal conditions of its use, however in special severe conditions, a large current flows instantaneously due to its high performance, and danger of an explosion with abrupt heat release is produced. Safety tests have been carried out to confirm the limit of safety performance. Results show abnormal circumstances such as high-rate discharge over 7A, high-rate charging of full discharged cells, nail-penetration, compression with a wedge and heating with a heat tape over 200{degree}C result in hazardous behaviors such as venting, firing and explosion. Therefore, this cell is equipped with proper protecting devices such as overcurrent and thermal protecting fuses to avoid hazardous behaviors. However, the severe conditions of handlings such as dumping into fire and approach to heat source, deformation and rupture by adding an external forces, and applications of too much vibration and impact, should be avoided. 5 refs., 9 figs., 2 tabs.

  16. Aluminum Removal From Hanford Waste By Lithium Hydrotalcite Precipitation - Laboratory Scale Validation On Waste Simulants Test Report

    International Nuclear Information System (INIS)

    Sams, T.; Hagerty, K.

    2011-01-01

    To reduce the additional sodium hydroxide and ease processing of aluminum bearing sludge, the lithium hydrotalcite (LiHT) process has been invented by AREV A and demonstrated on a laboratory scale to remove alumina and regenerate/recycle sodium hydroxide prior to processing in the WTP. The method uses lithium hydroxide (LiOH) to precipitate sodium aluminate (NaAI(OH) 4 ) as lithium hydrotalcite (Li 2 CO 3 .4Al(OH) 3 .3H 2 O) while generating sodium hydroxide (NaOH). In addition, phosphate substitutes in the reaction to a high degree, also as a filterable solid. The sodium hydroxide enriched leachate is depleted in aluminum and phosphate, and is recycled to double-shell tanks (DSTs) to leach aluminum bearing sludges. This method eliminates importing sodium hydroxide to leach alumina sludge and eliminates a large fraction of the total sludge mass to be treated by the WTP. Plugging of process equipment is reduced by removal of both aluminum and phosphate in the tank wastes. Laboratory tests were conducted to verify the efficacy of the process and confirm the results of previous tests. These tests used both single-shell tank (SST) and DST simulants.

  17. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    1983-10-01

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

  18. 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

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  2. Detailed technical plan for Test Program Element-III (TPE-III) of the first wall/blanket shield engineering test program

    International Nuclear Information System (INIS)

    Turner, L.R.; Praeg, W.F.

    1982-03-01

    The experimental requirements, test-bed design, and computational requirements are reviewed and updated. Next, in Sections 3, 4 and 5, the experimental plan, instrumentation, and computer plan, respectively, are described. Finally, Section 6 treats other considerations, such as personnel, outside participation, and distribution of results

  3. Detailed technical plan for Test Program Element-III (TPE-III) of the first wall/blanket shield engineering test program

    Energy Technology Data Exchange (ETDEWEB)

    Turner, L.R.; Praeg, W.F.

    1982-03-01

    The experimental requirements, test-bed design, and computational requirements are reviewed and updated. Next, in Sections 3, 4 and 5, the experimental plan, instrumentation, and computer plan, respectively, are described. Finally, Section 6 treats other considerations, such as personnel, outside participation, and distribution of results.

  4. Round Robin test for the determination of nitrogen concentration in solid Lithium

    International Nuclear Information System (INIS)

    Favuzza, P.; Antonelli, A.; Furukawa, T.; Groeschel, F.; Hedinger, R.; Higashi, T.; Hirakawa, Y.; Iijima, M.; Ito, Y.; Kanemura, T.; Knaster, J.; Kondo, H.; Miccichè, G.; Nitti, F.S.; Ohira, S.; Severi, M.; Sugimoto, M.; Suzuki, A.; Traversi, R.; Wakai, E.

    2016-01-01

    Highlights: • Nitrogen contained in solid Lithium is converted into Ammonium ion. • Ammonium ion is suitably quantified by ionic chromatograph or by Ammonia sensor. • Good agreement of the partner’s results has been achieved. • Maximum operative reproducibility and blank subtraction are necessary. - Abstract: Three different partners, ENEA, JAEA ed University of Tokyo, have been involved during 2014–2015 in the Round Robin experimentation for the assessment of the soundness of the analitycal procedure for the determination of the Nitrogen impurities contained inside a solid Lithium sample. Two different kinds of Lithium samples, differing by about an order of magnitude in Nitrogen concentration (∼230 wppm; ∼20–30 wppm), have been selected for this cross analysis. The agreement of the achieved results appears very good for what concerns the most concentrated Lithium and indicates each partner’s procedure is appropriate and intrinsecally able to lead to meaningful values, characterized by a relative uncertainty of just few %. The smaller agreement in the case of the less concentrated Lithium anyway points out that particular attention must be paid to reduce as much as possible any source of external contamination and highlights the importance of the proper blank subtraction.

  5. Round Robin test for the determination of nitrogen concentration in solid Lithium

    Energy Technology Data Exchange (ETDEWEB)

    Favuzza, P., E-mail: paolo.favuzza@enea.it [ENEA Center, Via Madonna del Piano 10, 50019 Sesto Fiorentino (Italy); Antonelli, A. [ENEA Research Center, Brasimone, 40035, Camugnano (Italy); Furukawa, T. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Groeschel, F. [KIT Research Center, Hermann-von-Helmholtz-Platz 1,76344 Eggenstein-Leopoldshafen (Germany); Hedinger, R. [F4E Research Center, Boltzmannstraße 2, 85748 Garching (Germany); Higashi, T. [University of Tokyo (Japan); Hirakawa, Y.; Iijima, M.; Ito, Y.; Kanemura, T. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Knaster, J. [IFMIF-EVEDA Project Team, Rokkasho (Japan); Kondo, H. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Miccichè, G.; Nitti, F.S. [ENEA Research Center, Brasimone, 40035, Camugnano (Italy); Ohira, S. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Severi, M. [University of Firenze, Via della Lastruccia 3, 50019 Sesto Fiorentino (Italy); Sugimoto, M. [JAEA Research Center, Rokkasho (Japan); Suzuki, A. [University of Tokyo (Japan); Traversi, R. [University of Firenze, Via della Lastruccia 3, 50019 Sesto Fiorentino (Italy); Wakai, E. [JAEA Research Center, Tokai-mura, Ibaraki (Japan)

    2016-06-15

    Highlights: • Nitrogen contained in solid Lithium is converted into Ammonium ion. • Ammonium ion is suitably quantified by ionic chromatograph or by Ammonia sensor. • Good agreement of the partner’s results has been achieved. • Maximum operative reproducibility and blank subtraction are necessary. - Abstract: Three different partners, ENEA, JAEA ed University of Tokyo, have been involved during 2014–2015 in the Round Robin experimentation for the assessment of the soundness of the analitycal procedure for the determination of the Nitrogen impurities contained inside a solid Lithium sample. Two different kinds of Lithium samples, differing by about an order of magnitude in Nitrogen concentration (∼230 wppm; ∼20–30 wppm), have been selected for this cross analysis. The agreement of the achieved results appears very good for what concerns the most concentrated Lithium and indicates each partner’s procedure is appropriate and intrinsecally able to lead to meaningful values, characterized by a relative uncertainty of just few %. The smaller agreement in the case of the less concentrated Lithium anyway points out that particular attention must be paid to reduce as much as possible any source of external contamination and highlights the importance of the proper blank subtraction.

  6. 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

  7. 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

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

    Science.gov (United States)

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

    2017-12-01

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

  9. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

    A comparative study of different types of fast reactor radial blankets is presented. Included are blankets of fertile material UO 2 , THO 2 and Th-metal blankets of pure reflectors C, BeO, Ni and combinations of reflecting and fertile blankets. The results for 1000MWe cores indicate that there is no incentive to use other than fertile blankets. The most favorable fertile material is thorium due to the prospective higher price of U-233

  10. Remote-maintenance features of the Fusion Materials Irradiation Test (FMIT) facility lithium system

    International Nuclear Information System (INIS)

    Kelly, V.P.

    1981-01-01

    The FMIT Facitlity has been redesigned to allow remote maintenance of lithium system components such as the main lithium pump, heat exchanger, traps, and valves. Remote versus contact maintenance is required due to the limited effectiveness of methods planned for control of the radioisotope 7 Be which is formed in the lithium during the neutron generating process. The altered FMIT arrangement provides cubicles for isolation of the main pump and system dump valve to allow personnel access for completing welds on large diameter piping after replacing failed components. Maintenance on other components such as the main heat exchanger and traps will be done remotely. The resulting arrangement provides full capability to meet the required facility availability criteria with minimum impact on the facility and with minimum associated development costs

  11. 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

  12. Lithium mass transport in ceramic breeder materials

    International Nuclear Information System (INIS)

    Blackburn, P.E.; Johnson, C.E.

    1990-01-01

    The objective of this activity is to measure the lithium vaporization from lithium oxide breeder material under differing temperature and moisture partial pressure conditions. Lithium ceramics are being investigated for use as tritium breeding materials. The lithium is readily converted to tritium after reacting with a neutron. With the addition of 1000 ppM H 2 to the He purge gas, the bred tritium is readily recovered from the blanket as HT and HTO above 400 degree C. Within the solid, tritium may also be found as LiOT which may transport lithium to cooler parts of the blanket. The pressure of LiOT(g), HTO(g), or T 2 O(g) above Li 2 O(s) is the same as that for reactions involving hydrogen. In our experiments we were limited to the use of hydrogen. The purpose of this work is to investigate the transport of LiOH(g) from the blanket material. 8 refs., 1 fig., 3 tabs

  13. Temperature gradient compatibility tests of some refractory metals and alloys in bismuth and bismuth--lithium solutions

    International Nuclear Information System (INIS)

    DiStefano, J.R.; Cavin, O.B.

    1976-11-01

    Quartz, T-111, and Mo thermal-convection loop tests were conducted at temperatures up to 700 0 C (100 0 C ΔT) to determine the compatibility of several refractory metals/alloys with bismuth and bismuth-lithium solutions for molten salt breeder reactor applications. Methods of evaluation included weight change measurements, metallographic examination, chemical and electron microprobe analysis, and mechanical properties tests. Molybdenum, T-111, and TA--10 percent W appear to be the most promising containment materials, while niobium and iron-based alloys are unacceptable

  14. 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.)

  15. Methods of tritium recovery from molten lithium

    International Nuclear Information System (INIS)

    Farookhi, R.; Rogers, J.E.

    1968-01-01

    It is important to keep the tritium inventory in a blanket of a thermonuclear reactor at a low level both to eliminate possible hydriding of structural components and to reduce inventory cost. Removing the tritium from a lithium blanket by fractional distillation, flash vaporization, and fractional crystallization was investigated. No definitive data are available either on the vapor-liquid equilibrium between lithium and tritium at low T 2 concentrations, or on the rate of formation and decomposition of lithium tritide. The final distinction between the recovery systems discussed in this report will depend on such data, but presently distillation appears to be the best alternate to the diffusion scheme proposed by A.P. Fraas. The capital cost of equipment necessary to remove tritium by distillation appears to be greater than 10 million dollars for a 5000 MW system, whereas the capital cost associated with the diffusion process has been estimated to be 4 million dollars

  16. Design and development of ceramic breeder demo blanket

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  17. 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.

  18. Characterisation and radiolysis of modified lithium orthosilicate pebbles with noble metal impurities

    DEFF Research Database (Denmark)

    Tamulevičius, Sigitas; Zariņš, A.; Valtenbergs, O.

    2017-01-01

    Modified lithium orthosilicate (Li4SiO4) pebbles with additions of titanium dioxide (TiO2) are suggested as an alternative tritium breeding ceramic for the European solid breeder test blanket module. The noble metals – platinum (Pt), gold (Au) and rhodium (Rh), can be introduced into the modified...... Li4SiO4 pebbles during the melt-based process, due to the corrosion of Pt-Rh and Pt-Au alloy crucible components. In this study, the surface microstructure, chemical and phase composition of the modified Li4SiO4 pebbles with different contents of the noble metals was analysed. The influence...

  19. Internal short circuit and accelerated rate calorimetry tests of lithium-ion cells: Considerations for methane-air intrinsic safety and explosion proof/flameproof protection methods.

    Science.gov (United States)

    Dubaniewicz, Thomas H; DuCarme, Joseph P

    2016-09-01

    Researchers with the National Institute for Occupational Safety and Health (NIOSH) studied the potential for lithium-ion cell thermal runaway from an internal short circuit in equipment for use in underground coal mines. In this third phase of the study, researchers compared plastic wedge crush-induced internal short circuit tests of selected lithium-ion cells within methane (CH 4 )-air mixtures with accelerated rate calorimetry tests of similar cells. Plastic wedge crush test results with metal oxide lithium-ion cells extracted from intrinsically safe evaluated equipment were mixed, with one cell model igniting the chamber atmosphere while another cell model did not. The two cells models exhibited different internal short circuit behaviors. A lithium iron phosphate (LiFePO 4 ) cell model was tolerant to crush-induced internal short circuits within CH 4 -air, tested under manufacturer recommended charging conditions. Accelerating rate calorimetry tests with similar cells within a nitrogen purged 353-mL chamber produced ignitions that exceeded explosion proof and flameproof enclosure minimum internal pressure design criteria. Ignition pressures within a 20-L chamber with 6.5% CH 4 -air were relatively low, with much larger head space volume and less adiabatic test conditions. The literature indicates that sizeable lithium thionyl chloride (LiSOCl 2 ) primary (non rechargeable) cell ignitions can be especially violent and toxic. Because ignition of an explosive atmosphere is expected within explosion proof or flameproof enclosures, there is a need to consider the potential for an internal explosive atmosphere ignition in combination with a lithium or lithium-ion battery thermal runaway process, and the resulting effects on the enclosure.

  20. 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

  1. 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

  2. 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

  3. 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)

  4. Design of a new lithium ion battery test cell for in-situ neutron diffraction measurements

    Czech Academy of Sciences Publication Activity Database

    Roberts, M.; Biendicho, J. J.; Hull, S.; Beran, Přemysl; Gustafsson, T.; Svensson, G.; Edstrom, K.

    2013-01-01

    Roč. 226, MAR 15 (2013), s. 249-255 ISSN 0378-7753 R&D Projects: GA MŠk(XE) LM2011019 Institutional support: RVO:61389005 Keywords : neutron * Lithium * LiFePO4 * diffraction * in situ Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 5.211, year: 2013 http://www.sciencedirect.com/science/article/pii/S0378775312016515

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  6. 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

  7. Evaluation of T-111 forced-convection loop tested with lithium at 13700C

    International Nuclear Information System (INIS)

    DeVan, J.H.; Long, E.L. Jr.

    1975-04-01

    A T-111 alloy (Ta--8 percent W--2 percent Hf) forced-convection loop containing molten lithium was operated 3000 h at a maximum temperature of 1370 0 C. Flow velocities up to 6.3 m/s were used. The results obtained in this forced-convection loop are very similar to those observed in lower velocity thermal-convection loops of T-111 containing lithium. Weight changes were determined at 93 positions around the loop. The maximum dissolution rate occurred at the maximum wall temperature of the loop and was less than 1.3 μ m/year. Mass transfer of hafnium, nitrogen, and, to a lesser extent, carbon occurred from the hotter to cooler regions. Exposed surfaces in the highest temperature region were found to be depleted in hafnium to a depth of 60 μ m with no detectable change in tungsten content. There was some loss in room-temperature tensile strength for specimens exposed to lithium at 1370 0 C, attributable to depletion of hafnium and nitrogen and to attendant grain growth. (U.S.)

  8. Lithium ceramics: sol-gel preparation and tritium release

    International Nuclear Information System (INIS)

    Renoult, O.

    1994-04-01

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

  9. Development of blanket remote maintenance system

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou

    1998-01-01

    ITER in-vessel components such as blankets are scheduled maintenance components, including complete shield blanket replacement for breeding blankets. In-vessel components are activated by 14 MeV neutrons, so blanket maintenance requires remote handling equipment and tools able to handle heavy payloads of about 4 tons within a positioning accuracy of 2 mm under intense gamma radiation. To facilitate remote maintenance, blankets are segmented into 730 modules and rail-mounted vehicle remote maintenance was developed. According to the ITER R and D program, critical technology related to blanket maintenance was developed extensively through joint efforts of the Japan, EU, and U.S. home teams. This paper summarizes current blanket maintenance technology conducted by the Japan Home Team, including development of full-scale remote handling equipment and tools for blanket maintenance. (author)

  10. Development of blanket remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

    Kakudate, Satoshi; Nakahira, Masataka; Oka, Kiyoshi; Taguchi, Kou [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    ITER in-vessel components such as blankets are scheduled maintenance components, including complete shield blanket replacement for breeding blankets. In-vessel components are activated by 14 MeV neutrons, so blanket maintenance requires remote handling equipment and tools able to handle heavy payloads of about 4 tons within a positioning accuracy of 2 mm under intense gamma radiation. To facilitate remote maintenance, blankets are segmented into 730 modules and rail-mounted vehicle remote maintenance was developed. According to the ITER R and D program, critical technology related to blanket maintenance was developed extensively through joint efforts of the Japan, EU, and U.S. home teams. This paper summarizes current blanket maintenance technology conducted by the Japan Home Team, including development of full-scale remote handling equipment and tools for blanket maintenance. (author)

  11. 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

  12. Industrial routes for lithium zirconate elements

    International Nuclear Information System (INIS)

    Bastide, B.; Roux, N.

    1991-01-01

    Lithium metazirconate Li 2 ZrO 3 is one of the leading ceramics contemplated in solid blanket concepts. Among its merits are fair physical properties, satisfactory compatibility with structural materials and beryllium, satisfactory mechanical strength, excellent irradiation behavior as shown by a comparative irradiation of ceramics in EBR 2 reactor, and very good tritium release performance as evidence in the MOZART, and EXOTIC neutron irradiation. Pechiney and the CEA are jointly involved in developing industrial fabrication of Li 2 ZrO 3 elements to the microstructural, geometrical (pellets, rings, spheres) specifications required for their use in solid blanket conceived in the European Program

  13. 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.

  14. Progress and achievements of the ITER L-4 blanket project

    International Nuclear Information System (INIS)

    Daenner, W.; Toschi, R.; Cardella, A.

    1999-01-01

    The L-4 Blanket Project embraces the R and D of the ITER Shielding Blanket, and its main objective is the fabrication of prototype components. This paper summarises the main conclusions from the materials R and D and the development of technologies which were required for the prototype specifications and manufacturing. The main results of the ongoing testing activities, and of the component manufacture are outlined.The main objectives of the project have been achieved including improvements of the material properties and of joining technologies, which resulted in good component quality and high performance in qualification tests. (author)

  15. Progress and achievements of the ITER L-4 blanket project

    International Nuclear Information System (INIS)

    Daenner, W.; Toschi, R.; Cardella, A.

    2001-01-01

    The L-4 Blanket Project embraces the R and D of the ITER Shielding Blanket, and its main objective is the fabrication of prototype components. This paper summarises the main conclusions from the materials R and D and the development of technologies which were required for the prototype specifications and manufacturing. The main results of the ongoing testing activities, and of the component manufacture are outlined. The main objectives of the project have been achieved including improvements of the material properties and of joining technologies, which resulted in good component quality and high performance in qualification tests. (author)

  16. Examination of compression and resilience characteristics of fibrous insulation blankets

    International Nuclear Information System (INIS)

    Brislin, R.J.; Middleton, A.

    1979-08-01

    Load-deflection characteristics of alumina and alumino-silicate fibrous blankets were experimentally determined. Load retention and springback capability of combinations of these materials were measured in a 10,000-hour test at surface temperatures of 650 to 1000 0 C (1200 to 1832 0 F). Experimental results are presented and future testing plans are discussed

  17. 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.)

  18. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

    Rocaboy, Alain.

    1982-06-01

    Materials with structures suitable for this purpose are studied. A bibliographic review of the main solid and liquid lithiated compounds is then presented. Erosion, dimensioning and maintenance problems associated with the limiter and the first wall of the reactor are studied from the point of view of the constraints they impose on the design of the blankets. Detailed studies of the main solid and liquid blanket concepts enable the best technological compromises to be determined for the indispensable functions of the blanket to be assured under acceptable conditions. Our analysis leads to four classes of solution, which cannot at this stage be considered as final recommendations, but which indicate what sort of solutions it is worthwhile exploring and comparing in order to be in a position to suggest a realistic blanket at the time when plasma control is sufficiently good for power reactors to be envisaged. Some considerations on the general architecture of the reactor are indicated. Energy storage with pulsed reactors is discussed in the appendix, and a first approach made to minimizing the total tritium recovery [fr

  19. Lithium purity and characterization

    International Nuclear Information System (INIS)

    Meadows, G.E.; Keough, R.F.

    1981-02-01

    The accurate measurement of impurities in lithium is basic to the study of lithium compatibility with fusion reactor materials. In the last year the Hanford Engineering Development Laboratory (HEDL) has had the opportunity to develop sampling and analytical techniques and to apply them in support of the Experimental Lithium System (ELS) as a part of the Fusion Materials Irradiation Test Project. In this paper we present the analytical results from the fill, start-up and operation of the ELS. In addition, the analysis and purification of navy surplus ingot lithium which is being considered for use in a larger system will be discussed. Finally, the analytical techniques used in our laboratory will be summarized and the results of a recent round robin lithium analysis will be presented

  20. D-shaped configurations in FTU for testing liquid lithium limiter: Preliminary studies and experiments

    Directory of Open Access Journals (Sweden)

    G. Ramogida

    2017-08-01

    A possible alternative connection of the poloidal field coils in FTU is here proposed, with the aim of achieving a true X-point configuration with a magnetic single null well inside the plasma chamber and strike points on the lithium limiter. A preliminary assessment of this design allowed estimating the required power supply upgrade and showed its compatibility with the existing mechanical structure and cooling system, at least for plasmas with current up to 300 kA and flat-top duration up to 4s.

  1. Experimental lithium system. Final report

    International Nuclear Information System (INIS)

    Kolowith, R.; Berg, J.D.; Miller, W.C.

    1985-04-01

    A full-scale mockup of the Fusion Materials Irradiation Test (FMIT) Facility lithium system was built at the Hanford Engineering Development Laboratory (HEDL). This isothermal mockup, called the Experimental Lithium System (ELS), was prototypic of FMIT, excluding the accelerator and dump heat exchanger. This 3.8 m 3 lithium test loop achieved over 16,000 hours of safe and reliable operation. An extensive test program demonstrated satisfactory performance of the system components, including the HEDL-supplied electromagnetic lithium pump, the lithium jet target, the purification and characterization hardware, as well as the auxiliary argon and vacuum systems. Experience with the test loop provided important information on system operation, performance, and reliability. This report presents a complete overview of the entire Experimental Lithium System test program and also includes a summary of such areas as instrumentation, coolant chemistry, vapor/aerosol transport, and corrosion

  2. Solubility of lithium deuteride in liquid lithium

    International Nuclear Information System (INIS)

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

    1977-01-01

    The solubility of LiD in liquid lithium between the eutectic and monotectic temperatures was measured using a direct sampling method. Solubilities were found to range from 0.0154 mol.% LiD at 199 0 C to 3.32 mol.% LiD at 498 0 C. The data were used in the derivation of an expression for the activity coefficient of LiD as a function of temperature and composition and an equation relating deuteride solubility and temperature, thus defining the liquidus curve. Similar equations were also derived for the Li-LiH system using the existing solubility data. Extrapolation of the liquidus curves yielded the eutectic concentrations (0.040 mol.% LiH and 0.035 mol.% LiD) and the freezing point depressions (0.23 0 C for Li-LiH and 0.20 0 C for Li-LiD) at the eutectic point. The results are compared with the literature data for hydrogen and deuterium. The implications of the relatively high solubility of hydrogen isotopes in lithium just above the melting point are discussed with respect to the cold trapping of tritium in fusion reactor blankets. (Auth.)

  3. 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

  4. Absolute measurement of the responses of small lithium glass scintillators to gamma radiation

    International Nuclear Information System (INIS)

    Dalton, A.W.

    1987-04-01

    The absolute scintillation efficiency and intrinsic resolution of lithium glass scintillators for electron excitation have been determined over a range of electron energies, lithium concentrations and lithium enrichments. Measurements of these response characteristics form part of a study on the possible use of such glasses for the determination of tritium breeding in fusion reactor blanket experiments. The measurements were undertaken to establish a basis for extracting the information relating to tritium production reactions from the background signals induced within the glass scintillators by the neutron/gamma fields of a fusion reactor blanket. Criteria for the selection of glasses most suitable for tritium breeding measurements are discussed in tems of their observed responses

  5. 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

  6. Analysis of the WCLL European demo blanket concept in terms of activation and decay heat after exposure to neutron irradiation

    Directory of Open Access Journals (Sweden)

    Stankunas Gediminas

    2017-01-01

    Full Text Available This comparative paper describes the activation and decay heat calculations for water-cooled lithium-lead performed part of the EURO fusion WPSAE programme and specifications in comparison to other European DEMO blanket concepts on the basis of using a three-dimensional neutronics calculation model. Results are provided for a range of decay times of interest for maintenance activities, safety and waste management assessments. The study revealed that water-cooled lithium-lead has the highest total decay heat at longer decay times in comparison to the helium-cooled design which has the lowest total decay heat. In addition, major nuclides were identified for water-cooled lithium-lead in W armour, Eurofer, and LiPb. In addition, great attention has been dedicated to the analysis of the decay heat and activity both from the different water-cooled lithium-lead blanket modules for the entire reactor and from each water-cooled lithium-lead blanket module separately. The neutron induced activation and decay heat at shutdown were calculated by the FISPACT code, using the neutron flux densities and spectra that were provided by the preceding MCNP neutron transport calculations.

  7. 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

  8. Beryllium research on FFHR molten salt blanket

    International Nuclear Information System (INIS)

    Terai, T.; Tanaka, S.; Sze, D.-K.

    2000-01-01

    Force-free helical reactor, FFHR, is a demo-relevant heliotron-type D-T fusion reactor based on the great amount of R and D results obtained in the LHD project. Since 1993, collaboration works have made great progress in design studies of FFHR with standing on the major advantage of current-less steady operation with no dangerous plasma disruptions. There are two types of reference designs, FFHR-1 and FFHR-2, where molten Flibe (LiF-BeF2) is utilized as tritium breeder and coolant. In this paper, we present the outline of FFHR blanket design and some related R and D topics focusing on Be utilization. Beryllium is used as a neutron multiplier in the design and Be pebbles are placed in the front part of the tritium breeding zone. In a Flibe blanket, HF (TF) generated due to nuclear transmutation will be a problem because of its corrosive property. Though nickel-based alloys are thought to be intact in such a corrosive environment, FFHR blanket design does not adopt the alloys because of their induced radioactivity. The present candidate materials for the structure are low-activated ferritic steel (JLF-1), V-4Cr-4Ti, etc. They are capable to be corroded by HF in the operation condition, and Be is expected to work as a reducing agent in the system as well. Whether Be pebbles placed in a Flibe flow can work well or not is a very important matter. From this point, Be solubility in Flibe, reaction rate of the Redox reaction with TF in the liquid and on the surface of Be pebbles under irradiation, flowing behavior of Flibe through a Be pebble bed, etc. should be investigated. In 1997, in order to establish more practical and new data bases for advanced design works, we started a collaboration work of R and D on blanket engineering, where the Be research above mentioned is included. Preliminary dipping-test of Be sheets and in-situ tritium release experiment from Flibe with Be sheets have got started. (orig.)

  9. 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

  10. Low pressure lithium condensation

    International Nuclear Information System (INIS)

    Wadkins, R.P.; Oh, C.H.

    1985-01-01

    A low pressure experiment to evaluate the laminar film condensation coefficients of lithium was conducted. Some thirty-six different heat transfer tests were made at system pressures ranging from 1.3 to 26 Pa. Boiled lithium was condensed on the inside of a 7.6-cm (ID), 409 stainless-steel pipe. Condensed lithium was allowed to reflux back to the pool boiling region below the condensing section. Fourteen chromel/alumel thermocouples were attached in various regions of the condensing section. The thermocouples were initially calibrated with errors of less than one degree Celsius

  11. Simulation of the fusion materials irradiation test facility lithium and heat transport systems for abnormal events study

    International Nuclear Information System (INIS)

    Carlson, W.F.; Elyashar, N.N.

    1981-01-01

    A digital computer model of Fusion Materials Irradiation Test Facility's heat transport system has been developed. The model utilizes a set of coupled differential equations to simulate the dynamic behavior of the primary and secondary heat transport loop systems. The model has been used to investigate the stability of the proposed control schemes for lithium temperature and flow rate and for an extensive study of equipment failures and malfunction analysis. It was determined that certain equipment failures and malfunctions in the primary loop require a response from the control system within less than one second of the occurrence of the failure. The effects of equipment failures in the secondary loop were found to be less dramatic than the equivalent failures in the primary loop. The failures in the secondary loop generally required control action in time frames of the order of minutes

  12. RF DEMO ceramic helium cooled blanket, coolant and energy transformation systems

    International Nuclear Information System (INIS)

    Kovalenko, V.; Leshukov, A.; Poliksha, V.; Popov, A.; Strebkov, Yu.; Borisov, A.; Shatalov, G.; Demidov, V.; Kapyshev, V.

    2004-01-01

    RF DEMO-S reactor is a prototype of commercial fusion reactors for further generation. A blanket is the main element unit of the reactor design. The segment structure is the basis of the ceramic blanket. The segments mounting/dismounting operations are carried out through the vacuum vessel vertical port. The inboard/outboard blanket segment is the modules welded design, which are welded by back plate. The module contains the back plate, the first wall, lateral walls and breeding zone. The 9CrMoVNb steel is used as structural material. The module internal space formed by the first wall, lateral walls and back plate is used for breeding zone arrangement. The breeding zone design based upon the poloidal BIT (Breeder Inside Tube) concept. The beryllium is used as multiplier material and the lithium orthosilicate is used as breeder material. The helium at 0.1 MPa is used as purge gas. The cooling is provided by helium at 10 MPa. The coolant supply/return to the blanket modules are carrying out on the two independent circuits. The performed investigations of possible transformation schemes of DEMO-S blanket heat power into the electricity allowed to make a conclusion about the preferable using of traditional steam-turbine facility in the secondary circuit. (author)

  13. Thermal cycling tests on Li4SiO4 and beryllium pebbles

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Norajitra, P.; Weisenburger, A.

    1995-01-01

    The European B.O.T. Demo-relevant solid breeder blanket is based on the use of beds of beryllium and Li 4 SiO 4 pebbles. Particularly dangerous for the pebble integrity are the rapid temperature changes which could occur, for instance, by a sudden blanket power shut-down. A series of thermal cycle tests have been performed for various beds of beryllium and Li 4 SiO 4 pebbles. No breaking was observed in the beryllium pebbles, however the Li 4 SiO 4 pebbles broke by temperature rates of change of about -50 C/sec independently on pebbles size and lithium enrichment. This value is considerably higher than the peak temperature rates of change expected in the blanket. (orig.)

  14. Examination results on reaction of lithium

    International Nuclear Information System (INIS)

    Asada, Takashi

    2000-12-01

    Before the material corrosion tests in lithium, the reactions of lithium with air and ammonia that will be used for lithium cleaning were examined, and the results were as follows. 1. When lithium put into air, surface of lithium changes to black first but soon to white, and the white layer becomes gradually thick. The first black of lithium surface is nitride (Li 3 N) and it changes to white lithium hydroxide (LiOH) by reaction with water in air, and it grows. The growth rate of the lithium hydroxide is about 1/10 in the desiccator (humidity of about 10%) compare with in air. 2. When lithium put into nitrogen, surface of lithium changes to black, and soon changes to brown and cracks at surface. At the same time with this cracking, weight of lithium piece increases and nitridation progresses respectively rapidly. This nitridation completed during 1-2 days on lithium rod of 10 mm in diameter, and increase in weight stopped. 3. Lithium melts in liquid ammonia and its melting rate is about 2-3 hour to lithium of 1 g. The liquid ammonia after lithium melting showed dark brown. (author)

  15. Aluminum Removal And Sodium Hydroxide Regeneration From Hanford Tank Waste By Lithium Hydrotalcite Precipitation Summary Of Prior Lab-Scale Testing

    International Nuclear Information System (INIS)

    Sams, T.L.; Guillot, S.

    2011-01-01

    Scoping laboratory scale tests were performed at the Chemical Engineering Department of the Georgia Institute of Technology (Georgia Tech), and the Hanford 222-S Laboratory, involving double-shell tank (DST) and single-shell tank (SST) Hanford waste simulants. These tests established the viability of the Lithium Hydrotalcite precipitation process as a solution to remove aluminum and recycle sodium hydroxide from the Hanford tank waste, and set the basis of a validation test campaign to demonstrate a Technology Readiness Level of 3.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  17. Neutronic data consistency analysis for lithium blanket and shield design

    International Nuclear Information System (INIS)

    Reupke, W.A.; Muir, D.W.

    1976-01-01

    Using a compact least-squares treatment we analyze the consistency of evaluated cross sections with calculated and measured tritium production in /sup n/Li and 7 Li detectors embedded in a 14-MeV neutron driven /sup n/LiD sphere. The tritium production experimental error matrix is evaluated and an initial reduced chi 2 of 3.0 is calculated. A perturbation calculation of the tritium production cross section sensitivities is performed with secondary neutron energy and angular distributions held constant. The cross section error matrix is evaluated by the external consistency of available cross section measurements. A statistical adjustment of the combined data yields a reduced chi 2 of 2.3 and represents a tenfold improvement in statistical likelihood. The improvement is achieved by a decrease in the 7 Li(n,xt) 14-MeV group cross section from 328 mb to 284 mb and an adjustment of the /sup n/Li data closer to calculated values. The uncertainty in the tritium breeding ratio in pure 7 LiD is reduced by one-fifth

  18. Lithium Intoxication

    Directory of Open Access Journals (Sweden)

    Sermin Kesebir

    2011-09-01

    Full Text Available Lithium has been commonly used for the treatment of several mood disorders particularly bipolar disorder in the last 60 years. Increased intake and decreased excretion of lithium are the main causes for the development of lithium intoxication. The influence of lithium intoxication on body is evaluated as two different groups; reversible or irreversible. Irreversible damage is usually related with the length of time passed as intoxicated. Acute lithium intoxication could occur when an overdose of lithium is received mistakenly or for the purpose of suicide. Patients may sometimes take an overdose of lithium for self-medication resulting in acute intoxication during chronic, while others could develop chronic lithium intoxication during a steady dose treatment due to a problem in excretion of drug. In such situations, it is crucial to be aware of risk factors, to recognize early clinical symptoms and to conduct a proper medical monitoring. In order to justify or exclude the diagnosis, quantitative evaluation of lithium in blood and toxicologic screening is necessary. Following the monitoring schedules strictly and urgent intervention in case of intoxication would definitely reduce mortality and sequela related with lithium intoxication. In this article, the etiology, frequency, definition, clinical features and treatment approaches to the lithium intoxication have been briefly reviewed.

  19. Small scale lithium-lead/water-interaction studies

    International Nuclear Information System (INIS)

    Kranert, O.; Kottowski, H.

    1991-01-01

    One current concept in fusion blanket design is to utilize water as the coolant and liquid lithium-lead as the breeding/neutron multiplier material. Considering the complex design of the blanket module, it is likely that a water leakage into the liquid alloy may occur due to a tube rupture provoking an intolerable pressure increase in the blanket module. The pressure increase is caused by the combined chemical and thermohydraulic reaction of lithium-lead with water. Experiments which simulate such a transient event are necessary to obtain information which is important for the blanket module design. The interaction has been investigated by conducting small-scale experiments at various injection pressures, alloy- and coolant temperatures. Besides using eutectic Li 17 Pb 83 , Li 7 Pb 2 , lithium and lead have been used. Among other results, the experiments indicate increasing chemical reaction with increasing lithium concentration. At the same time, the chemical reaction inhibits violent thermohydaulic reactions due to the attenuating effect of the hydrogen produced. The preliminary epxerimental results from Li 17 Pb 83 and Li 7 Pb 2 reveal that the pressure- and temperature transients caused by the chemical and thermohydraulic reactions lie within technically manageable limits. (orig.)

  20. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    1983-10-01

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

  1. Design requirement on HYPER blanket fuel assembly

    International Nuclear Information System (INIS)

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

    2000-07-01

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

  2. Rapid thermal cycling of new technology solar array blanket coupons

    Science.gov (United States)

    Scheiman, David A.; Smith, Bryan K.; Kurland, Richard M.; Mesch, Hans G.

    1990-01-01

    NASA Lewis Research Center is conducting thermal cycle testing of a new solar array blanket technologies. These technologies include test coupons for Space Station Freedom (SSF) and the advanced photovoltaic solar array (APSA). The objective of this testing is to demonstrate the durability or operational lifetime of the solar array interconnect design and blanket technology within a low earth orbit (LEO) or geosynchronous earth orbit (GEO) thermal cycling environment. Both the SSF and the APSA array survived all rapid thermal cycling with little or no degradation in peak performance. This testing includes an equivalent of 15 years in LEO for SSF test coupons and 30 years of GEO plus ten years of LEO for the APSA test coupon. It is concluded that both the parallel gap welding of the SSF interconnects and the soldering of the APSA interconnects are adequately designed to handle the thermal stresses of space environment temperature extremes.

  3. Neutronics and activation of the preliminary reaction chamber of HiPER reactor based in a SCLL blanket

    Energy Technology Data Exchange (ETDEWEB)

    Juárez, Rafael, E-mail: rafael.juarez@upm.es [Instituto de Fusión Nuclear, UPM, Madrid (Spain); Escuela Técnica Superior de Ingenieros Industriales, UNED, Madrid (Spain); Sanz, Javier; Lopez-Revelles, A.J. [Escuela Técnica Superior de Ingenieros Industriales, UNED, Madrid (Spain); Perlado, José Manuel [Instituto de Fusión Nuclear, UPM, Madrid (Spain)

    2013-10-15

    Highlights: • Neutronic study of a proposal of a reaction chamber for HiPER reactor. • Two options for the blanket size, thin and thick, are studied and compared. • The thin blanket performs better than the thick blanket. • The proposed Vacuum Vessel is unviable as lifetime component in both cases. • Likely solutions for the Vacuum Vessel lifetime extension are explored. -- Abstract: The HiPER reactor design is exploring different reaction chambers. In this study, we tackle the neutronics and activation studies of a preliminary reaction chamber based in the following technologies: unprotected dry wall for the First Wall, self-cooled lead lithium blanket, and independent low activation steel Vacuum Vessel. The most critical free parameter in this stage is the blanket thickness, as a function of the {sup 6}Li enrichment. After a parametric study, we select for study both a “thin” and “thick” blanket, with “high” and “low” {sup 6}Li enrichment respectively, to reach a TBR = 1.1. To help to make a choice, we compute, for both blanket options, in addition to the TBR, the energy amplification factor, the tritium partial pressure, the {sup 203}Hg and {sup 210}Po total activity in the LiPb loop, and the Vacuum Vessel thickness required to guarantee the reweldability during its lifetime. The thin blanket shows a superior performance in the safety related issues and structural viability, but it operates at higher {sup 6}Li enrichment. It is selected for further improvements. The Vacuum Vessel shows to be unviable in both cases, with the thickness varying between 39 and 52 cm. Further chamber modifications, such as the introduction of a neutron reflector, are required to exploit the benefits of the thin blanket with a reasonable Vacuum Vessel.

  4. 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

  5. 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

  6. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

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

    1987-10-01

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

  7. 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.

  8. 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.

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

    Science.gov (United States)

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

    2017-11-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  11. Lithium Poisoning

    DEFF Research Database (Denmark)

    Baird-Gunning, Jonathan; Lea-Henry, Tom; Hoegberg, Lotte C G

    2017-01-01

    Lithium is a commonly prescribed treatment for bipolar affective disorder. However, treatment is complicated by lithium's narrow therapeutic index and the influence of kidney function, both of which increase the risk of toxicity. Therefore, careful attention to dosing, monitoring, and titration...... is required. The cause of lithium poisoning influences treatment and 3 patterns are described: acute, acute-on-chronic, and chronic. Chronic poisoning is the most common etiology, is usually unintentional, and results from lithium intake exceeding elimination. This is most commonly due to impaired kidney...... function caused by volume depletion from lithium-induced nephrogenic diabetes insipidus or intercurrent illnesses and is also drug-induced. Lithium poisoning can affect multiple organs; however, the primary site of toxicity is the central nervous system and clinical manifestations vary from asymptomatic...

  12. Large lithium loop experience

    International Nuclear Information System (INIS)

    Kolowith, R.; Owen, T.J.; Berg, J.D.; Atwood, J.M.

    1981-10-01

    An engineering design and operating experience of a large, isothermal, lithium-coolant test loop are presented. This liquid metal coolant loop is called the Experimental Lithium System (ELS) and has operated safely and reliably for over 6500 hours through September 1981. The loop is used for full-scale testing of components for the Fusion Materials Irradiation Test (FMIT) Facility. Main system parameters include coolant temperatures to 430 0 C and flow to 0.038 m 3 /s (600 gal/min). Performance of the main pump, vacuum system, and control system is discussed. Unique test capabilities of the ELS are also discussed

  13. LITHIUM DEPLETION IS A STRONG TEST OF CORE-ENVELOPE RECOUPLING

    International Nuclear Information System (INIS)

    Somers, Garrett; Pinsonneault, Marc H.

    2016-01-01

    Rotational mixing is a prime candidate for explaining the gradual depletion of lithium from the photospheres of cool stars during the main sequence. However, previous mixing calculations have relied primarily on treatments of angular momentum transport in stellar interiors incompatible with solar and stellar data in the sense that they overestimate the internal differential rotation. Instead, recent studies suggest that stars are strongly differentially rotating at young ages but approach a solid body rotation during their lifetimes. We modify our rotating stellar evolution code to include an additional source of angular momentum transport, a necessary ingredient for explaining the open cluster rotation pattern, and examine the consequences for mixing. We confirm that core-envelope recoupling with a ∼20 Myr timescale is required to explain the evolution of the mean rotation pattern along the main sequence, and demonstrate that it also provides a more accurate description of the Li depletion pattern seen in open clusters. Recoupling produces a characteristic pattern of efficient mixing at early ages and little mixing at late ages, thus predicting a flattening of Li depletion at a few Gyr, in agreement with the observed late-time evolution. Using Li abundances we argue that the timescale for core-envelope recoupling during the main sequence decreases sharply with increasing mass. We discuss the implications of this finding for stellar physics, including the viability of gravity waves and magnetic fields as agents of angular momentum transport. We also raise the possibility of intrinsic differences in initial conditions in star clusters using M67 as an example.

  14. Performance evaluation on force control for ITER blanket installation

    Energy Technology Data Exchange (ETDEWEB)

    Aburadani, A., E-mail: aburadani.atsushi@jaea.go.jp [Japan Atomic Energy Agency, Mukouyama 801-1, Naka, Ibaraki 311-0193 (Japan); Takeda, N.; Shigematsu, S.; Murakami, S.; Tanigawa, H.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka, Ibaraki 311-0193 (Japan); Nakahira, M.; Hamilton, D.; Tesini, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► It is crucial issues to avoid any jamming between the blanket modules and the keys. ► Force control for AC servo motor was developed to reduce excessive loads. ► This jam prevention force control method is directly measured and controlled by AC servo motor controllers. ► In the recent test, the module was passively positioned onto keys using the torque control method. -- Abstract: The most critical issue for the ITER blanket installation is to avoid any jamming between the blanket modules and the keys as a result of excessive loading during the module installation process. This is complicated by the limited clearance of 0.5 mm between the modules and the keys. To solve these technical issues, force control, such as controlling the torque for the AC servo motors, was developed to reduce excessive loads which may have an impact on the end-effector and to defer the forces acting on the groove of the blanket. This jam prevention force control method is directly measured and controlled by AC servo motor controllers. The AC servo motors are equipped to move the manipulator and end-effector during module installation.

  15. Performance evaluation on force control for ITER blanket installation

    International Nuclear Information System (INIS)

    Aburadani, A.; Takeda, N.; Shigematsu, S.; Murakami, S.; Tanigawa, H.; Kakudate, S.; Nakahira, M.; Hamilton, D.; Tesini, A.

    2013-01-01

    Highlights: ► It is crucial issues to avoid any jamming between the blanket modules and the keys. ► Force control for AC servo motor was developed to reduce excessive loads. ► This jam prevention force control method is directly measured and controlled by AC servo motor controllers. ► In the recent test, the module was passively positioned onto keys using the torque control method. -- Abstract: The most critical issue for the ITER blanket installation is to avoid any jamming between the blanket modules and the keys as a result of excessive loading during the module installation process. This is complicated by the limited clearance of 0.5 mm between the modules and the keys. To solve these technical issues, force control, such as controlling the torque for the AC servo motors, was developed to reduce excessive loads which may have an impact on the end-effector and to defer the forces acting on the groove of the blanket. This jam prevention force control method is directly measured and controlled by AC servo motor controllers. The AC servo motors are equipped to move the manipulator and end-effector during module installation

  16. 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

  17. (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

  18. Design and analysis of ITER shield blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-01

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

  19. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

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

    1997-05-01

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

  20. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

    2009-01-01

    Selecting relevant features is in demand when a large data set is of interest in a classification task. It produces a tractable number of features that are sufficient and possibly improve the classification performance. This paper studies a statistical method of Markov blanket induction algorithm...... for filtering features and then applies a classifier using the Markov blanket predictors. The Markov blanket contains a minimal subset of relevant features that yields optimal classification performance. We experimentally demonstrate the improved performance of several classifiers using a Markov blanket...... induction as a feature selection method. In addition, we point out an important assumption behind the Markov blanket induction algorithm and show its effect on the classification performance....

  1. 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

  2. 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.

  3. Lot Acceptance, Abuse and Life Testing of Varta Lithium Polymer Pouch Cells

    Directory of Open Access Journals (Sweden)

    Anderson Amy

    2017-01-01

    The tests performed involved assessing individual cell performance relating to capacity under a variety of environmental conditions as well as establishing cell safety via abuse testing for small satellite systems.

  4. 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.

  5. Evaluation of compost blankets for erosion control from disturbed lands.

    Science.gov (United States)

    Bhattarai, Rabin; Kalita, Prasanta K; Yatsu, Shotaro; Howard, Heidi R; Svendsen, Niels G

    2011-03-01

    Soil erosion due to water and wind results in the loss of valuable top soil and causes land degradation and environmental quality problems. Site specific best management practices (BMP) are needed to curb erosion and sediment control and in turn, increase productivity of lands and sustain environmental quality. The aim of this study was to investigate the effectiveness of three different types of biodegradable erosion control blankets- fine compost, mulch, and 50-50 mixture of compost and mulch, for soil erosion control under field and laboratory-scale experiments. Quantitative analysis was conducted by comparing the sediment load in the runoff collected from sloped and tilled plots in the field and in the laboratory with the erosion control blankets. The field plots had an average slope of 3.5% and experiments were conducted under natural rainfall conditions, while the laboratory experiments were conducted at 4, 8 and 16% slopes under simulated rainfall conditions. Results obtained from the field experiments indicated that the 50-50 mixture of compost and mulch provides the best erosion control measures as compared to using either the compost or the mulch blanket alone. Laboratory results under simulated rains indicated that both mulch cover and the 50-50 mixture of mulch and compost cover provided better erosion control measures compared to using the compost alone. Although these results indicate that the 50-50 mixtures and the mulch in laboratory experiments are the best measures among the three erosion control blankets, all three types of blankets provide very effective erosion control measures from bare-soil surface. Results of this study can be used in controlling erosion and sediment from disturbed lands with compost mulch application. Testing different mixture ratios and types of mulch and composts, and their efficiencies in retaining various soil nutrients may provide more quantitative data for developing erosion control plans. Copyright © 2010 Elsevier

  6. Inversion of lithium heparin gel tubes after centrifugation is a significant source of bias in clinical chemistry testing.

    Science.gov (United States)

    Lippi, Giuseppe; Salvagno, Gian Luca; Danese, Elisa; Lima-Oliveira, Gabriel; Brocco, Giorgio; Guidi, Gian Cesare

    2014-09-25

    This study was planned to establish whether random orientation of gel tubes after centrifugation may impair sample quality. Eight gel tubes were collected from 17 volunteers: 2 Becton Dickinson (BD) serum tubes, 2 Terumo serum tubes, 2 BD lithium heparin tubes and 2 Terumo lithium heparin tubes. One patient's tube for each category was kept in a vertical, closure-up position for 90 min ("upright"), whereas paired tubes underwent bottom-up inversion every 15 min, for 90 min ("inverted"). Immediately after this period of time, 14 clinical chemistry analytes, serum indices and complete blood count were then assessed in all tubes. Significant increases were found for phosphate and lipaemic index in all inverted tubes, along with AST, calcium, cholesterol, LDH, potassium, hemolysis index, leukocytes, erythrocytes and platelets limited to lithium heparin tubes. The desirable quality specifications were exceeded for AST, LDH, and potassium in inverted lithium heparin tubes. Residual leukocytes, erythrocytes, platelets and cellular debris were also significantly increased in inverted lithium heparin tubes. Lithium heparin gel tubes should be maintained in a vertical, closure-up position after centrifugation. Copyright © 2014 Elsevier B.V. All rights reserved.

  7. Lithium Batteries

    Science.gov (United States)

    National Laboratory, Materials Science and Technology Division Lithium Batteries Resources with Additional thin-film lithium batteries for a variety of technological applications. These batteries have high essentially any size and shape. Recently, Teledyne licensed this technology from ORNL to make batteries for

  8. Verification of tritium production evaluation procedure using Monte Carlo code MCNP for in-pile test of fusion blanket with JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Y. E-mail: nagao@jmtr.oarai.jaeri.go.jp; Nakamichi, K.; Tsuchiya, M.; Ishitsuka, E.; Kawamura, H

    2000-11-01

    To evaluate exactly the total amount of tritium production in tritium breeding materials during in-pile test with JMTR, the 'tritium monitor' has been produced and evaluation of total tritium generation was done by using 'tritium monitor' in preliminary in-pile mock-up, and verification of procedure concerning tritium production evaluation was conducted by using Monte Carlo code MCNP and nuclear cross section library of FSXLIBJ3R2. Li-Al alloy (Li 3.4 wt.%, 95.5% enrichment of {sup 6}Li) was selected as tritium monitor material for the evaluation on the total amount of tritium production in high {sup 6}Li enriched materials. From the results of preliminary experiment, calculated amounts of total tritium production at each 'tritium monitor', which was installed in the preliminary in-pile mock-up, were about 50-290% higher than the measured values. Concerning tritium measurement, increase of measurement error in tritium leak form measuring system to measure small amount of tritium (0.2-0.7 mCi in tritium monitor) was found in the results of present experiment. The tendency for overestimation of calculated thermal neutron flux in the range of 1-6x10{sup 13} n cm{sup -2} per s was found in JMTR and the reason may be due to the beryllium cross section data base in JENDL3.2.

  9. Verification of tritium production evaluation procedure using Monte Carlo code MCNP for in-pile test of fusion blanket with JMTR

    International Nuclear Information System (INIS)

    Nagao, Y.; Nakamichi, K.; Tsuchiya, M.; Ishitsuka, E.; Kawamura, H.

    2000-01-01

    To evaluate exactly the total amount of tritium production in tritium breeding materials during in-pile test with JMTR, the 'tritium monitor' has been produced and evaluation of total tritium generation was done by using 'tritium monitor' in preliminary in-pile mock-up, and verification of procedure concerning tritium production evaluation was conducted by using Monte Carlo code MCNP and nuclear cross section library of FSXLIBJ3R2. Li-Al alloy (Li 3.4 wt.%, 95.5% enrichment of 6 Li) was selected as tritium monitor material for the evaluation on the total amount of tritium production in high 6 Li enriched materials. From the results of preliminary experiment, calculated amounts of total tritium production at each 'tritium monitor', which was installed in the preliminary in-pile mock-up, were about 50-290% higher than the measured values. Concerning tritium measurement, increase of measurement error in tritium leak form measuring system to measure small amount of tritium (0.2-0.7 mCi in tritium monitor) was found in the results of present experiment. The tendency for overestimation of calculated thermal neutron flux in the range of 1-6x10 13 n cm -2 per s was found in JMTR and the reason may be due to the beryllium cross section data base in JENDL3.2

  10. Neutronics experiments for uncertainty assessment of tritium breeding in HCPB and HCLL blanket mock-ups irradiated with 14 MeV neutrons

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Pillon, M.; Villari, R.; Fischer, U.; Klix, A.; Leichtle, D.; Kodeli, I.; Pohorecki, W.

    2012-01-01

    Two neutronics experiments have been carried out at 14 MeV neutron sources on mock-ups of the helium cooled pebble bed (HCBP) and the helium cooled lithium lead (HCLL) variants of ITER test blanket modules (TBMs). These experiments have provided an experimental validation of the calculations of the tritium production rate (TPR) in the two blanket concepts and an assessment of the uncertainties due to the uncertainties on nuclear data. This paper provides a brief summary of the HCPB experiment and then focuses in particular on the final results of the HCLL experiment. The TPR has been measured in the HCLL mock-up irradiated for long times at the Frascati 14 MeV Neutron Generator (FNG). Redundant and well-assessed experimental techniques have been used to measure the TPR by different teams for inter-comparison. Measurements of the neutron and gamma-ray spectra have also been performed. The analysis of the experiment, carried out by the MCNP code with FENDL-2.1 and JEFF-3.1.1 nuclear data libraries, and also including sensitivity/uncertainty analysis, shows good agreement between measurements and calculations, within the total uncertainty of 5.9% at 1σ level. (paper)

  11. Tokamak blanket design study: FY 78 summary report

    International Nuclear Information System (INIS)

    1979-06-01

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

  12. Manufacturing Technology of Ceramic Pebbles for Breeding Blanket

    Directory of Open Access Journals (Sweden)

    Rosa Lo Frano

    2018-05-01

    Full Text Available An open issue for the fusion power reactor is the choice of breeding blanket material. The possible use of Helium-Cooled Pebble Breeder ceramic material in the form of pebble beds is of great interest worldwide as demonstrated by the numerous studies and research on this subject. Lithium orthosilicate (Li4SiO4 is a promising breeding material investigated in this present study because the neutron capture of Li-6 allows the production of tritium, 6Li (n, t 4He. Furthermore, lithium orthosilicate has the advantages of low activation characteristics, low thermal expansion coefficient, high thermal conductivity, high density and stability. Even if they are far from the industrial standard, a variety of industrial processes have been proposed for making orthosilicate pebbles with diameters of 0.1–1 mm. However, some manufacturing problems have been observed, such as in the chemical stability (agglomeration phenomena. The aim of this study is to provide a new methodology for the production of pebbles based on the drip casting method, which was jointly developed by the DICI-University of Pisa and Industrie Bitossi. Using this new (and alternative manufacturing technology, in the field of fusion reactors, appropriately sized ceramic pebbles could be produced for use as tritium breeders.

  13. Study of the fire behavior of high-energy lithium-ion batteries with full-scale burning test

    Science.gov (United States)

    Ping, Ping; Wang, QingSong; Huang, PeiFeng; Li, Ke; Sun, JinHua; Kong, DePeng; Chen, ChunHua

    2015-07-01

    A full-scale burning test is conducted to evaluate the safety of large-size and high-energy 50 Ah lithium-iron phosphate/graphite battery pack, which is composed of five 10 Ah single cells. The complex fire hazards associated with the combustion process of the battery are presented. The battery combustion behavior can be summarized into the following stages: battery expansion, jet flame, stable combustion, a second cycle of a jet flame followed by stable combustion, a third cycle of a jet flame followed by stable combustion, abatement and extinguishment. The multiple jets of flame indicate serious consequences for the battery and pose a challenge for battery safety. The battery ignites when the battery temperature reaches approximately 175-180 °C. This critical temperature is related to an internal short circuit of the battery, which results from the melting of the separator. The maximum temperature of the flame can reach 1500 °C. The heat release rate (HRR) varies based on the oxygen generated by the battery and the Joule effect of the internal short circuit. The HRR and heat of combustion can reach 49.4 kW and 18,195.1 kJ, respectively. The state of charge of the battery has a significant effect on the maximum HRR, the overall heat generation and the mass loss of the battery.

  14. 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

  15. Convertible shielding to ceramic breeding blanket

    International Nuclear Information System (INIS)

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

    1995-05-01

    Four concepts have been studied for the ITER convertible blanket: 1)Layered concept 2)BIT(Breeder-Inside-Tube)concept 3)BOT(Breeder-Out of-Tube)concept 4)BOT/mixed concept. All concepts use ceramic breeder and beryllium neutron multiplier, both in the shape of small spherical pebbles, 316SS structure, and H 2 O coolant (inlet/outlet temperatures : 100/150degC, pressure : 2 MPa). During the BPP, only beryllium pebbles (the primary pebble in case of BOT/mixed concept) are filled in the blanket for shielding purpose. Then, before the EPP operation, breeder pebbles will be additionally inserted into the blanket. Among possible conversion methods, wet method by liquid flow seems expecting for high and homogeneous pebble packing. Preliminary 1-D neutronics calculation shows that the BOT/mixed concept has the highest breeding and shielding performance. However, final selection should be done by R and D's and more detail investigation on blanket characteristics and fabricability. Required R and D's are also listed. With these efforts, the convertible blanket can be developed. However, the following should be noted. Though many of above R and D's are also necessary even for non-convertible blanket, R and D's on convertibility will be one of the most difficult parts and need significant efforts. Besides the installation of convertible blanket with required structures and lines for conversion will make the ITER basic machine more complicated. (author)

  16. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

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

    1993-11-01

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

  17. 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.)

  18. 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

  19. ITER breeding blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

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

  20. Analysis of the WCLL European demo blanket concept in terms of activation and decay heat after exposure to neutron irradiation

    OpenAIRE

    Stankunas Gediminas; Tidikas Andrius

    2017-01-01

    This comparative paper describes the activation and decay heat calculations for water-cooled lithium-lead performed part of the EURO fusion WPSAE programme and specifications in comparison to other European DEMO blanket concepts on the basis of using a three-dimensional neutronics calculation model. Results are provided for a range of decay times of interest for maintenance activities, safety and waste management assessments. The study revealed that water-c...

  1. Lithium extractive metallurgy

    International Nuclear Information System (INIS)

    Josa, J.M.; Merino, J.L.

    1985-01-01

    The Nuclear Fusion National Program depends on lithium supplies. Extractive metallurgy development is subordinate to the localization and evaluation of ore resources. Nowadays lithium raw materials usable with present technology consist of pegmatite ore and brine. The Instituto Geologico y Minero Espanol (IGME) found lepidolite, ambligonite and spodrimene in pegmatite ores in different areas of Spain. However, an evaluation of resources has not been made. Different Spanish surface and underground brines are to be sampled and analyzed. If none of these contain significant levels of lithium, the Junta de Energia Nuclear (JEN) will try an agreement with IGME for ENUSA (Empresa Nacional del Uranio, S.A.) to explore pegmatite-ore bodies from different locations. Different work stages, laboratory tests, pilots plants tests and commercial plant, are foreseen, if the deposits are found. (author)

  2. Beryllium R and D for blanket application

    Energy Technology Data Exchange (ETDEWEB)

    Dalle Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karls