WorldWideScience

Sample records for bayonet ifmif back-plate

  1. Preliminary evaluation of the expected radiation damage of the bayonet IFMIF back-plate

    Energy Technology Data Exchange (ETDEWEB)

    Frisoni, M. [Athena s.a.s, Via del Battiferro 3, I-40129 Bologna (Italy)], E-mail: manuela.frisoni@enea.it; Agostini, P. [ENEA CR Brasimone, Bacino del Brasimone 40032, Camugnano (Bolivia, Plurinational State of) (Italy); Fasanella, D. [Athena s.a.s, Via del Battiferro 3, I-40129 Bologna (Italy); Micciche, G. [ENEA CR Brasimone, Bacino del Brasimone 40032, Camugnano (Bolivia, Plurinational State of) (Italy)

    2009-06-15

    This paper summarises and discusses the results of a preliminary damage assessment of the non-seizure coating of the bayonet IFMIF back-plate. Neutron-induced kerma factors, dpa and gas production cross sections libraries were produced in a multigroup structure for neutron energies up to 60 MeV, by processing evaluated nuclear data files with NJOY-99.259 system. The material damage evaluations in terms of heat deposition, displacement and gas production rates were calculated using these libraries and compared with the values obtained using the data contained in the pointwise ACE format files of MCNP5 code package. The calculations were performed with MCNP5 code both using the McEnea and the McDelicious neutron source models to reproduce the energy-angle distributions of the neutrons produced in IFMIF d-Li interactions.

  2. Improvement of IFMIF/EVEDA bayonet concept back-plate design

    Energy Technology Data Exchange (ETDEWEB)

    Bernardi, D., E-mail: davide.bernardi@enea.it [ENEA Brasimone- I-40032 Camugnano (Italy); Agostini, P.; Micciche, G.; Nitti, F.S.; Tincani, A. [ENEA Brasimone- I-40032 Camugnano (Italy)

    2011-10-15

    In the frame of the Engineering Validation and Engineering Design Activities (EVEDA) phase of the International Fusion Materials Irradiation Facility (IFMIF) project, a supporting lithium loop has been designed and is currently under construction at Oarai (Japan) with the main objective to test several technological solutions to be adopted in the future IFMIF plant. Among these, the lithium target system represents one of the most critical components as it will be exposed to high-energy intense neutron flux and consequently to severe irradiation damage rates (up to 60 dpa/fpy). For this reason, it must be designed for periodic replacement. The solution proposed by ENEA is based on the so-called back-plate bayonet concept which consists of a replaceable element that can be inserted to and removed from the permanent structure of the target assembly by means of a sliding-skate mechanism. Recently, the design of the bayonet back-plate has been revised and some important modifications have been introduced in order to improve its functionality and optimize its features in terms of compactness, robustness and remote maintainability. Several design solutions have been conceived to achieve better performance including smaller overall dimensions, sealing load reduction, gasket retention system improvement, positioning and centering effectiveness and optimized detachment mechanism. Moreover, a new variable-curvature geometry for the lithium channel profile has been calculated using an analytic approach based on the simplified Navier-Stokes equations in order to avoid the fluid dynamic instabilities evidenced in the old profile. In this paper, the new design features of the back-plate are presented, along with the main outcomes obtained from the engineering assessment performed so far.

  3. Improvement of IFMIF/EVEDA bayonet concept back-plate design

    International Nuclear Information System (INIS)

    Bernardi, D.; Agostini, P.; Micciche, G.; Nitti, F.S.; Tincani, A.

    2011-01-01

    In the frame of the Engineering Validation and Engineering Design Activities (EVEDA) phase of the International Fusion Materials Irradiation Facility (IFMIF) project, a supporting lithium loop has been designed and is currently under construction at Oarai (Japan) with the main objective to test several technological solutions to be adopted in the future IFMIF plant. Among these, the lithium target system represents one of the most critical components as it will be exposed to high-energy intense neutron flux and consequently to severe irradiation damage rates (up to 60 dpa/fpy). For this reason, it must be designed for periodic replacement. The solution proposed by ENEA is based on the so-called back-plate bayonet concept which consists of a replaceable element that can be inserted to and removed from the permanent structure of the target assembly by means of a sliding-skate mechanism. Recently, the design of the bayonet back-plate has been revised and some important modifications have been introduced in order to improve its functionality and optimize its features in terms of compactness, robustness and remote maintainability. Several design solutions have been conceived to achieve better performance including smaller overall dimensions, sealing load reduction, gasket retention system improvement, positioning and centering effectiveness and optimized detachment mechanism. Moreover, a new variable-curvature geometry for the lithium channel profile has been calculated using an analytic approach based on the simplified Navier-Stokes equations in order to avoid the fluid dynamic instabilities evidenced in the old profile. In this paper, the new design features of the back-plate are presented, along with the main outcomes obtained from the engineering assessment performed so far.

  4. IFMIF Li target back-plate design integration and thermo-mechanical analysis

    International Nuclear Information System (INIS)

    Riccardi, B.; Roccella, S.; Micciche, G.

    2006-01-01

    The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-driven intense neutron source where fusion reactor candidate materials will be tested. The neutron flux is produced by means of a deuteron beam (current 250 mA, energy 40 MeV) that strikes a liquid lithium target circulating in a lithium loop. The support on which the liquid lithium flows, i.e. the back-plate, is the most heavily exposed component to neutron flux. A '' bayonet '' concept solution for the back-plate was proposed by ENEA with the objectives of improving the back-plate reliability and simplifying the remote handling procedures. On the base of this concept, a back-plate mock-up was fabricated and validated. Starting from the findings of the mock up design, a back-plate design integration exercise was carried out in order to check if the back-plate geometrical features are compatible with the target assembly and the Vertical Test Assemblies (VTA). The work carried out has demonstrated that even with the changes operated for the design integration (increase of in-plane dimensions and reduction of thickness) the bayonet concept is able to guarantee a tight connection to the target assembly. A thermo-mechanical analysis of the back-plate has been carried out by means of ABAQUS code. The thermal load used as input for the calculations, i.e. the neutron heat generation, has been estimated by means of Monte Carlo Mc-Delicious code. The two boundary constraint cases (full and minimum contact with target assembly) considered for each back-plate geometry option represent the extreme cases of the real operating condition of the plate. The influence of the contact heat exchange coefficient and the back-plate thickness has been also evaluated. For all these reasons, the results of the analysis can be considered as the domain of variability of the real working conditions. The results show that AISI 316L steel is not suitable as black-plate material: the stress induced in the plate, in

  5. Analysis of the thermomechanical behavior of the IFMIF bayonet target assembly under design loading scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Bernardi, D., E-mail: davide.bernardi@enea.it [ENEA Brasimone, Camugnano, BO (Italy); Arena, P.; Bongiovì, G.; Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy); Frisoni, M. [ENEA Bologna, Via Martiri di Monte Sole 4, Bologna (Italy); Miccichè, G.; Serra, M. [ENEA Brasimone, Camugnano, BO (Italy)

    2015-10-15

    In the framework of the IFMIF Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) phase, ENEA is responsible for the design of the European concept of the IFMIF lithium target system which foresees the possibility to periodically replace only the most irradiated and thus critical component (i.e., the backplate) while continuing to operate the rest of the target for a longer period (the so-called bayonet backplate concept). In this work, the results of the steady state thermomechanical analysis of the IFMIF bayonet target assembly under two different design loading scenarios (a “hot” scenario and a “cold” scenario) are briefly reported highlighting the relevant indications obtained with respect to the fulfillment of the design requirements. In particular, the analyses have shown that in the hot scenario the temperatures reached in the target assembly are within the material acceptable limits while in the cold scenario transition below the ductile to brittle transition temperature (DBTT) cannot be excluded. Moreover, results indicate that the contact between backplate and high flux test module is avoided and that the overall structural integrity of the system is assured in both scenarios. However, stress linearization analysis reveals that ITER Structural Design Criteria for In-vessel Components (SDC-IC) design rules are not always met along the selected paths at backplate middle plane section in the hot scenario, thus suggesting the need of a revision of the backplate design or a change of the operating conditions.

  6. Rotary bayonets for cryogenic and vacuum service

    International Nuclear Information System (INIS)

    Rucinski, R.A.; Dixon, K.D.; Krasa, R.; Krempetz, K.J.; Mulholland, G.T.; Trotter, G.R.; Urbin, J.B.

    1993-07-01

    Rotary bayonets were designed, tested, and installed for liquid nitrogen, liquid argon, and vacuum service. This paper will present the design, testing, and service record for two sizes of vacuum jacketed cryogenic rotary bayonets and two sizes of vacuum service rotary bayonets. Materials used in construction provide electrical isolation across the bayonet joint. The joint permits 360 degrees of rotation between the male and female pipe sections while maintaining integrity of service. Assemblies using three such joints were built to allow end connection points to be translated through at least 1 meter of horizontal travel while kept in service. Vacuum jacketed sizes built in-house at Fermi National Accelerator Laboratory are 1-1/2 in. inner pipe size, 3 in. vacuum jacket, and 4 in. inner pipe size, 6 in. vacuum jacket The single wall vacuum service bayonets are in 4 in. and 6 in. pipe sizes. The bayonets have successfully been in active service for over one year

  7. Rotary bayonets for cryogenic and vacuum service

    International Nuclear Information System (INIS)

    Rucinski, R.A.; Dixon, K.D.; Krasa, R.; Krempetz, K.J.; Mulholland, G.T.; Trotter, G.R.; Urbin, J.B.

    1994-01-01

    Rotary bayonets were designed, tested, and installed for liquid nitrogen, liquid argon, and vacuum service. This paper will present the design, testing, and service record for two sizes of vacuum jacketed cryogenic rotary bayonets and two sizes of vacuum service rotary bayonets. Materials used in construction provide electrical isolation across the bayonet joint. The joint permits 360 degrees of rotation between the male and female pipe sections while maintaining integrity of service. Assemblies using three such joints were built to allow end connection points to be translated through at least 1 meter of horizontal travel while kept in service. Vacuum jacketed sizes built in-house at Fermi National Accelerator Laboratory are 1 1/2 inches inner pipe size, 3 inches vacuum jacket, and 4 inches inner pipe size, 6 inches vacuum jacket. The single wall vacuum service bayonets are in 4 inch and 6 inch pipe sizes. The bayonets have successfully been in active service for over one year

  8. Users' requirements for IFMIF

    International Nuclear Information System (INIS)

    Noda, K.; Jitsukawa, S.; Ehrlich, K.; Moeslang, A.

    1998-01-01

    The International Fusion Materials Irradiation Facility (IFMIF) is a high energy neutron irradiation facility which generates an intense neutron flux with D-Li stripping reactions for fusion materials testing. The role of IFMIF is (1) development of various fusion reactor materials, (2) determination of design-relevant engineering databases for the DEMO fusion reactor, (3) calibration and validation of data generated from fission reactor irradiations and the other simulation experiments, etc. The conceptual design activity (CDA) of IFMIF was initiated in February 1995 as an IEA collaborative activity to complete a reference conceptual design of IFMIF in December 1996. Users' requirements for the conceptual design of IFMIF were developed for materials to be tested, types of experiments, small specimen test technology and irradiation conditions. Furthermore, the neutron irradiation field characteristics (spectrum, flux/volume, etc.) of IFMIF were evaluated for the conceptual design parameters and were shown to meet the essential requirements of the users. (orig.)

  9. Engineering validation for lithium target facility of the IFMIF under IFMIF/EVEDA project

    Directory of Open Access Journals (Sweden)

    E. Wakai

    2016-12-01

    Full Text Available The International Fusion Materials Irradiation Facility (IFMIF, presently in the Engineering Validation and Engineering Design Activities (EVEDA phase was started from 2007 under the frame of the Broader Approach (BA agreement. In the activities, a prototype Li loop with the world's highest flow rate of 3000L/min was constructed in 2010, and it succeeded in generating a 100mm wide and 25mm thick with a free-surface lithium flow along a concave back plate steadily at a high-speed of 15m/s at 250°C for 1300h. In the demonstration operation it was needed to develop the Li flowing measurement system with precious resolution less than 0.1mm, and a new wave height measuring method which is laser-probe method was developed for measurements of the 3D geometry of the liquid Li target surface. Using the device, the stability of the variation in the Li flowing thickness which is required in the IFMIF specification was ±1mm or less as the liquid Li target, and the result was satisfied with it and the feasibility of the long-term stable liquid Li flow was also verified. The results of the other engineering validation tests such as lithium purification tests of lithium target facility have also been evaluated and summarized.

  10. IFMIF accelerator conceptual design activities

    International Nuclear Information System (INIS)

    Jameson, R.A.; Lagniel, J.M.; Sugimoto, M.; Kein, H.; Piaszczyk, C.; Tiplyakov, V.

    1998-01-01

    A Conceptual Design Evaluation (CDE) for the International Fusion Materials Irradiation Facility (IFMIF) began in 1997 and will be completed in 1998, as an international program of the IEA involving the European Community, Japan, Russia and the United States. The IFMIF accelerator system, comprising two 125 mA, 40 MeV deuterium accelerators operating at 175 MHz, is a key element of the IFMIF facility. The objectives and accomplishments of the CDE accelerator studies are outlined

  11. Lexan Linear Shaped Charge Holder with Magnets and Backing Plate

    Science.gov (United States)

    Maples, Matthew W.; Dutton, Maureen L.; Hacker, Scott C.; Dean, Richard J.; Kidd, Nicholas; Long, Chris; Hicks, Robert C.

    2013-01-01

    A method was developed for cutting a fabric structural member in an inflatable module, without damaging the internal structure of the module, using linear shaped charge. Lexan and magnets are used in a charge holder to precisely position the linear shaped charge over the desired cut area. Two types of charge holders have been designed, each with its own backing plate. One holder cuts fabric straps in the vertical configuration, and the other charge holder cuts fabric straps in the horizontal configuration.

  12. Addendum to IFMIF-CDA interim report

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Hiroshi; Ida, Mizuho [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; eds.

    1996-08-01

    During the second IFMIF-CDA Design Integration Workshop, the conceptual design and contents of `IFMIF-CDA Interim Report` were examined and discussed at both general and group meetings. Based on these discussion, the final IFMIF-CDA Report will be modified from the `Interim Report`. This report describes the outline of these modification. (author)

  13. Addendum to IFMIF-CDA interim report

    International Nuclear Information System (INIS)

    Maekawa, Hiroshi; Ida, Mizuho

    1996-08-01

    During the second IFMIF-CDA Design Integration Workshop, the conceptual design and contents of 'IFMIF-CDA Interim Report' were examined and discussed at both general and group meetings. Based on these discussion, the final IFMIF-CDA Report will be modified from the 'Interim Report'. This report describes the outline of these modification. (author)

  14. Comparison between beryllium and diamond-backing plates in diamond-anvil cells

    DEFF Research Database (Denmark)

    Periotto, Benedetta; Nestola, Fabrizio; Balic Zunic, Tonci

    2011-01-01

    A direct comparison between two complete intensity datasets, collected on the same sample loaded in two identical diamond-anvil pressure cells equipped, respectively, with beryllium and diamond backing plates was performed. The results clearly demonstrate that the use of diamond-backing plates...

  15. IFMIF accelerators design

    International Nuclear Information System (INIS)

    Mosnier, A.; Ratzinger, U.

    2008-01-01

    The IFMIF requirement for 250 mA current of deuteron beams at a nominal energy of 40 MeV is met by means of two identical continuous wave (CW) 175 MHz linear accelerators running in parallel, each delivering a 125 mA, 40 MeV deuteron beam to the common target. This approach allows to stay within the current capability of present RF linac technology while providing operational redundancy in case of failure of one of the linacs. Each linac comprises a sequence of acceleration and beam transport/matching stages. The ion source generates a 140 mA deuteron beam at 100 keV. A low energy beam transport (LEBT) transfers the deuteron beam from the source to a radio frequency quadrupole (RFQ) cavity. The RFQ bunches and accelerates the 125 mA beam to 5 MeV. The RFQ output beam is injected through a matching section into a drift-tube-linac (DTL) where it is accelerated to the final energy of 40 MeV. In the reference design, the final acceleration stage is a conventional Alvarez-type DTL with post-couplers operating at room temperature. Operation of both the RFQ and the DTL at the same relatively low frequency is essential for accelerating the high current deuteron beam with low beam loss. The primary concern of the IFMIF linacs is the minimization of beam losses, which could limit their availability and maintainability due to excessive activation of the linac and irradiation of the environment. A careful beam dynamics design is therefore needed from the source to the target to avoid the formation of particle halo that could finally be lost in the linac or transfer lines. A superconducting solution for the high energy portion of the linac using, for example, CH-structure or coaxial-type resonators, could offer some advantages, in particular the reduction of operational costs. Careful beam dynamics simulations and comparison tests with beam during the EVEDA phase are however necessary in order to fully assess the technical feasibility of such alternative solutions

  16. Design and production of a hermetic bayonet isolation valve

    International Nuclear Information System (INIS)

    Fuerst, J.

    1993-05-01

    Fermilab is upgrading the Tevatron for lower temperature/higher beam energy operation. Portions of the satellite refrigeration system will operate below atmospheric pressure after the upgrade is complete. Contamination must be prevented by hermetically sealing the subatmospheric helium to air interfaces. Bayonet connections in the low pressure flow path require a reliable, leak tight isolation valve instead of the standard quarter turn ball valve. Design, development, and production of a new valve are described

  17. Ulno-volar bayonet hand: Its differential diagnosis from Madelung's deformity

    Energy Technology Data Exchange (ETDEWEB)

    Christ, F.

    1981-04-01

    The ulno-volar bayonet hand related to the mostly hereditary multiple exostoses is compared to Madelung's forearm deformity under clinical and roentgenological view in differential diagnosis. The ulno-volar bayonet hand is considerably more seldom, basing upon dysplasia of the lower part of the ulna, less inconvenient in function, and hardly tending to the development of early arthrosis.

  18. Ulno-volar bayonet hand: Its differential diagnosis from Madelung's deformity

    International Nuclear Information System (INIS)

    Christ, F.

    1981-01-01

    The ulno-volar bayonet hand related to the mostly hereditary multiple exostoses is compared to Madelung's forearm deformity under clinical and roentgenological view in differential diagnosis. The ulno-volar bayonet hand is considerably more seldom, basing upon dysplasia of the lower part of the ulna, less inconvenient in function, and hardly tending to the development of early arthrosis. (orig.) [de

  19. Theoretical study and experimental detection of cavitation phenomena in Liquid Lithium Target Facility for IFMIF

    International Nuclear Information System (INIS)

    Orco, G. Dell; Horiike, H.; Ida, M.; Nakamura, H.

    2006-01-01

    In the IFMIF (International Fusion Materials Irradiation Facility) testing facility, the required high energy neutrons emission will be produced by reaction of two D + beams with a free surface liquid Lithium jet target flowing along concave back-wall at 20 m/s. The Lithium height in the experimental loop and its relevant static pressure, the high flow velocities and the presence of several devices for the flow control and the pressure reduction increase the risk of cavitation onset in the target system. Special attention has to be taken in the primary pump, in the flow straightener, in the nozzle and their interconnections where the local pressure decreases and/or velocity increases or flow separations could promote the emission of cavitation vapour bubbles. The successive bubble re-implosions, in the higher pressure liquid bulk, could activate material erosion and transportation of activated particulates. These bubbles, if emitted close to the free jet flow, could also procure hydraulic instability and disturbance of the neutron field in the D + beams-Lithium target zone. Therefore, the cavitation risk must be properly foreseen along the whole IFMIF Lithium target circuit and its occurrence at different operating condition should be also monitored by special instrumentation. ENEA, in close cooperation with JAEA, has demonstrated the capability to detect the onset of the cavitation noises in liquid Lithium, by using the ENEA patented accelerometric gauge called CASBA-2000, during hydraulic test campaigns carried-out at Osaka University Lithium facility on a straight mock-up of the IFMIF back plate target. Comparison with the Thoma' cavitation similitude criteria have also determined the critical threshold limit for the estimation of the onset. Theoretical study on the conditions of cavitations generation in the IFMIF Lithium Target Circuit were also launched between ENEA and JAEA aiming at analysing the risk of the cavitation occurrence in the Lithium flow by

  20. Nuclear data for designing the IFMIF accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    The objective of the International Fusion Materials Irradiation Facility (IFMIF) and the design concept of the IFMIF accelerator system are described. The status of the nuclear data, especially for the deuteron-induced reactions, to qualify the system design is reviewed. The requests for the nuclear data compilation and/or evaluation are summarized. (author)

  1. CFD calculations on the IFMIF Li-jet fluid dynamics

    International Nuclear Information System (INIS)

    Casal, N.

    2007-01-01

    surface disturbance. The heat flux to the back plate and pressures, temperatures, and velocities maps have been obtained. The occurrence of cavitation has been assessed and sensibility analysis carried out modifying some main flow parameters like velocity. (orig.)

  2. Minutes of the IFMIF technical meeting

    International Nuclear Information System (INIS)

    Nakamura, H.; Takeda, M.; Ida, M.; Maebara, S.; Yutani, T.; Sugimoto, M.

    2004-03-01

    The IFMIF Technical Meeting was held on December 4-5, 2003 at Shiran-kaikan, Kyoto University. The main objectives are 1) to finalize the Comprehensive Design Report (CDR), 2) to discuss IFMIF cost and organization, 3) to review technical status of major systems, transition phase activities and EVEDA plan. This report presents a brief summary of the results of the meeting. Agenda, participants list and presentation materials are attached as Appendix. (author)

  3. Interaction between bubble and air-backed plate with circular hole

    Science.gov (United States)

    Liu, Y. L.; Wang, S. P.; Zhang, A. M.

    2016-06-01

    This paper investigates the nonlinear interaction between a violent bubble and an air-backed plate with a circular hole. A numerical model is established using the incompressible potential theory coupled with the boundary integral method. A double-node technique is used to solve the overdetermined problem caused by the intersection between the solid wall and the free surface. A spark-generated bubble near the air-backed plate with a circular hole is observed experimentally using a high-speed camera. Our numerical results agree well with the experimental results. Both experimental and numerical results show that a multilevel spike emerges during the bubble's expansion and contraction. Careful numerical simulation reveals that this special type of spike is caused by the discontinuity in the boundary condition. The influences of the hole size and depth on the bubble and spike dynamics are also analyzed.

  4. The Effect of Strike Face Geometry on the Dynamic Delamination of Composite Back Plates

    Science.gov (United States)

    2015-01-01

    behind the ceramic (Zuogang et al. 2010). In many cases, Kevlar , S-2 glass, ultra-high-molecular-weight polyethylene, or a similar high- performance...thin, translucent S-2 glass/SC-15 epoxy backing plate. A 0.30-cal. fragment-simulating projectile (FSP) was used to strike the front of the target... epoxy was chosen as the composite backing. Quasi-static material and high strain- rate properties for this composite are well characterized and have been

  5. Progress in IFMIF Engineering Validation and Engineering Design Activities

    International Nuclear Information System (INIS)

    Heidinger, R.; Knaster, J.; Matsumoto, H.; Sugimoto, M.; Mosnier, A.; Arbeiter, F.; Baluc, N.; Cara, P.; Chel, S.; Facco, A.; Favuzza, P.; Heinzel, V.; Ibarra, A.; Massaut, V.; Micciche, G.; Nitti, F.S.; Theile, J.

    2013-01-01

    Highlights: ► The IFMIF/EVEDA project has entered into the crucial phase of concluding the Interim IFMIF Engineering Design Report. ► The IFMIF plant configuration has been established with the definition of five IFMIF facilities and of their interfaces. ► Three major prototypes of the IFMIF main systems have been designed and are being manufactured, commissioned and operated. -- Abstract: The International Fusion Materials Irradiation Facility (IFMIF) Engineering Design and Engineering Validation Activities (EVEDA) are being developed in a joint project in the framework of the Broader Approach (BA) Agreement between EU and Japan. This project has now entered into a crucial phase as the engineering design of IFMIF is now being formulated in a series of 3 subsequent phases for delivering an Interim IFMIF Engineering Design Report (IIEDR) by mid of 2013. Content of these phases is explained, including the plant configuration detailing the 5 IFMIF facilities and their systems. Together with the Engineering Design Activities, prototyping sub-projects are pursued in the Engineering Validation Activities which consist of the design, manufacturing and testing of the following prototypical systems: Linear IFMIF Prototype Accelerator (LIPAc), EVEDA Lithium Test Loop (ELTL), and High Flux Test Module (HFTM) with the prototypical helium cooling loop (HELOKA). Highlights are described from recent experiments in the Engineering Validation Activities

  6. IFMIF suitability for evaluation of fusion functional materials

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  7. The role of users group in the IFMIF project

    International Nuclear Information System (INIS)

    Matsui, H.; Sugimoto, M.; Mdslang, A.; Garin, R.

    2007-01-01

    Full text of publication follows: Experts of materials and fusion technology areas are the major 'users' of the IFMIF, the International Fusion Materials Irradiation Facility. Now that IFMIF-EVEDA (Engineering Validation and Engineering Design Activities) Project is implemented under the Broader Approach framework as an EU-JP bilateral collaboration, the Users' Group under IEA does not have a formal interface with the IFMIF-EVEDA project. This lack of interface may cause serious problems since this situation may lead to designing, and eventually, constructing an expensive facility that does not fulfill the users' requirements in an optimal manner. Direct connection of IEA Users group and IFMIF would face a serious difficulty since the participating parties are different to each other. The Broader Approach agreement foresees the creation of a Project Committee in particular to advise IFMIF-EVEDA Project Leader on its implementation. It would be recommendable to appoint fusion-materials and -technology experts from BA participating parties as Project Committee members and let them function as interface between the Users and IFMIF-EVEDA Project. Periodic exchange of opinion involving members of this Committee with Users group members would function as an interface with scientists from all over the world. Roles of users, i.e. fusion-materials and technology experts are identified as follows: - Contribute to the reviews all along the IFMIF-EVEDA progress - Recommendations for further improvement of irradiation and test conditions - Support performance assessments of IFMIF (if needed) - Exchange of information on: - reference materials to be used for IFMIF construction - test matrices, and PIE programs for mechanical testing (at IFMIF site) - advanced microstructural analysis at suitable international laboratories - Ensure the link with DEMO design criteria experts as well as breeding blanket expert group - Support establishment of 'materials design limit data' for IFMIF

  8. Overview of the IFMIF test facility design in IFMIF/EVEDA phase

    International Nuclear Information System (INIS)

    Tian, Kuo; Abou-Sena, Ali; Arbeiter, Frederik; García, Ángela; Gouat, Philippe; Heidinger, Roland; Heinzel, Volker; Ibarra, Ángel; Leysen, Willem; Mas, Avelino; Mittwollen, Martin; Möslang, Anton; Theile, Jürgen; Yamamoto, Michiyoshi; Yokomine, Takehiko

    2015-01-01

    Highlights: • This paper summarizes the current design status of IFMIF EVEDA test facility. • The principle functions of the test facility and key components are described. • The brief specifications of the systems and key components are addressed. - Abstract: The test facility (TF) is one of the three major facilities of the International Fusion Material Irradiation Facility (IFMIF). Engineering designs of TF main systems and key components have been initiated and developed in the IFMIF EVEDA (Engineering Validation and Engineering Design Activities) phase since 2007. The related work covers the designs of a test cell which is the meeting point of the TF and accelerator facility and lithium facility, a series of test modules for experiments under different irradiation conditions, an access cell to accommodate remote handling systems, four test module handling cells for test module processing and assembling, and test facility ancillary systems for engineering support on energy, media, and control infrastructure. This paper summarizes the principle functions, brief specifications, and the current design status of the above mentioned IFMIF TF systems and key components.

  9. Numerical Investigation of the IFMIF Lithium Target

    International Nuclear Information System (INIS)

    Gordeev, S.; Heinzel, V.; Slobodchuk, V.; Leichtle, D.; Anton Moeslang, A.

    2006-01-01

    The International Fusion Materials Facility (IFMIF) facility uses a high speed (10-20 m/s) Lithium (Li) jet flow as a target for two 40 MeV / 125 mA deuteron beams. The major function of the Li target is to provide a stable Li jet for the production of an intense neutron flux. For the understanding the lithium jet behaviour and elimination of the free-surface flow instabilities a detailed analysis of the Li jet flow is necessary. Numerical investigations of the IFMIF Li - Target have been performed with the CFD code Star-CD. A number of turbulence models were tested on the experimental data obtained at the water jet test facility of the Institute for Physics and Power Engineering (IPPE), Obninsk, Russia. Calculated and measured velocity profiles and thickness of the flow cross sections have been compared. The most suitable turbulence models were used for numerical investigations of the IFMIF Li-jet. For the analysis of the IFMIF Li target 3D models of the nozzle and jet flows have been developed. In the first part of analyses the nozzle flow effects, such as relaminarization of the accelerated flow, secondary motions and their influence on the development of the viscous layer and velocity profile have been investigated. Further evaluation of turbulence models was performed and recommendations for suitable turbulence models are given. Calculations predict the complete laminarization of the flow at the nozzle outlet for velocities less than 10 m/s. Within the transition region of velocities between 10 and 20 m/s calculations show the laminarization only in the first convergent part. In this case the acceleration dose not suppress secondary flows in the straight part near the nozzle exit. The second task is devoted to the stability of the Li jet flow. To this end, the influence of the nozzle outlet boundaries, jet curvature effects, gravity and surface tension on the free surface stability has been analysed. First calculations show, that such factors as gravity and

  10. Exploration of reliability databases and comparison of former IFMIF's results

    International Nuclear Information System (INIS)

    Tapia, Carlos; Dies, Javier; Abal, Javier; Ibarra, Angel; Arroyo, Jose M.

    2011-01-01

    There is an uncertainty issue about the applicability of industrial databases to new designs, such as the International Fusion Materials Irradiation Facility (IFMIF), as they usually contain elements for which no historical statistics exist. The exploration of common components reliability data in Accelerator Driven Systems (ADS) and Liquid Metal Technologies (LMT) frameworks is the milestone to analyze the data used in IFMIF reliability's reports and for future studies. The comparison between the reliability accelerator results given in the former IFMIF's reports and the databases explored has been made by means of a new accelerator Reliability, Availability, Maintainability (RAM) analysis. The reliability database used in this analysis is traceable.

  11. Study of the thermo-mechanical performances of the IFMIF-EVEDA Lithium Test Loop target assembly

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A., E-mail: dimaio@din.unipa.it [Dipartimento dell' Energia, Universita di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Arena, P.; Bongiovi, G. [Dipartimento dell' Energia, Universita di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Giammusso, R.; Micciche, G.; Tincani, A. [ENEA C. R. Brasimone, 40032 Camugnano, Bologna (Italy)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer IFMIF-EVEDA target assembly thermo-mechanical behavior has been investigated. Black-Right-Pointing-Pointer Finite element method has been followed and a commercial code has been used. Black-Right-Pointing-Pointer Nominal, design and pressure test steady state scenarios and start-up transient conditions have been investigated. Black-Right-Pointing-Pointer Steady state results have shown that back-plate yielding may occur only under the design scenario. Black-Right-Pointing-Pointer Transient analysis has indicated that TA start-up lasts for {approx}60 h. - Abstract: Within the framework of the IFMIF R and D program and in close cooperation with ENEA-Brasimone, at the Department of Energy of the University of Palermo a research campaign has been launched to investigate the thermo-mechanical behavior of the target assembly under both steady state and start-up transient conditions. A theoretical approach based on the finite element method (FEM) has been followed and a well-known commercial code has been adopted. A realistic 3D FEM model of the target assembly has been set-up and optimized by running a mesh independency analysis. A proper set of loads and boundary conditions, mainly concerned with radiation heat transfer between the target assembly external walls and the inner walls of its containment vessel, have been considered and the target assembly thermo-mechanical behavior under nominal, design and pressure test steady state scenarios and start-up transient conditions has been investigated. Results are herewith reported and discussed.

  12. Overview of the IFMIF test cell design

    International Nuclear Information System (INIS)

    Moeslang, A.; Daum, E.; Jitsukawa, S.; Noda, K.; Viola, R.

    1996-01-01

    The Conceptual Design Activity (CDA) for the International Fusion Materials Irradiation Facility (IFMIF) has entered its second and final year, and an outline design has been developed. Initial evaluations of the potential of this high flux, high intensity D-Li source have shown that the main materials testing needs can be fulfilled. According to these needs, Vertical Test Assemblies will accommodate test modules for the high flux (0.5 liter, 20 dpa/a, 250-1000 C), the medium flux (6 liter, 1-20 dpa/a, 250-1000 C), the low flux (7.5 liter, 0.1-1 dpa/a), and the very low flux (> 100 liter, 0.01-0.1 dpa/a) regions. Detailed test matrices have been defined for the high and medium flux regions, showing that on the basis of small specimen test technologies, a database for an engineering design of an advanced fusion reactor (DEMO) can be established for a variety of structural materials and ceramic breeders. The design concepts for the Test Cell, including test assemblies, remote handling equipment and Hot Cell Facilities with capacity for investigating all irradiation specimens at the IFMIF site are described

  13. LMFBR steam generator development: duplex bayonet tube steam generator. Volume II

    International Nuclear Information System (INIS)

    DeFur, D.D.

    1975-04-01

    This report represents the culmination of work performed in fulfillment of ERDA Contract AT(11-1)-2426, Task Agreement 2, in which alternate steam generator designs were developed and studied. The basic bayonet tube generator design previously developed by C-E under AEC Contract AT(11-1)-3031 was expanded by incorporating duplex heat transfer tubes to enhance the unit's overall safety and reliability. The effort consisted of providing and evaluating conceptual designs of the evaporator, superheater and reheater components for a large plant LMFBR steam generator (950 MWt per heat transport loop)

  14. Fluid dynamics characteristics of IFMIF Li-jet under deuteron load

    International Nuclear Information System (INIS)

    Fuertes, F.M.; Casal, N.; Barbero, R.; Garcia, A.; Branas, B.; Riccardi, B.

    2006-01-01

    IFMIF is an accelerator-based neutron source with the purpose of testing and fully qualify fusion candidate materials. Two 40 MeV deuteron beams, 125 mA current each, strike a target of liquid lithium flowing over a concave back-plate. The deuteron-lithium stripping reactions produce an intense high energy neutron flux which simulates the fusion reactor irradiation. To remove the beam power deposited on it (up to 10 MW), the lithium jet must have a speed around 20 m/s, which may give rise to flow instabilities. However, a stable liquid free surface is a very critical requirement of the target system, otherwise the neutron field could be altered. Therefore, the possible occurrences that could affect the hydrodynamical stability of the lithium jet are being examined in the frame of EFDA Technology Workprogramme. This paper summarizes the studies of the fluid dynamics characteristics of the lithium jet under the deuteron heat load, based on applications of the CFX 5.7 code, a commercial Navier-Stokes equations solver with specific modelling of turbulence, like the classical k - ε among others. Significant effort has been dedicated to develop an optimized and reliable numerical mesh, able to illustrate the behaviour of the lithium free surface and other issues like heat transport along the stream and to the back-plate, and lithium vaporization. First activities were dedicated to explore the effects on the results of a three-dimensional unstructured numerical mesh covering the area from the nozzle upstream the target to the exit of the target region. Subsequently, a more effective approach to this issue has been undertaken by developing a fine two-dimensional mesh along the longitudinal flow direction, with refined areas in the free surface and close to the wall regions. The numerical convergence criteria have been found to be strongly sensitive with respect to small modifications of the adopted unstructured mesh. Owing to the uncertainties associated with modelling

  15. Conceptual design of the IFMIF Start-Up monitoring module

    Energy Technology Data Exchange (ETDEWEB)

    Gouat, Philippe, E-mail: philippe.gouat@sckcen.be [SCK-CEN – The Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium); Leysen, Willem; Goussarov, Andrei; Galledou, Papa Sally [SCK-CEN – The Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium); Rapisarda, David; Mota, Fernando; Garcia, Angela [CIEMAT – Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, Avda. Complutense 40, 28040 Madrid (Spain)

    2013-10-15

    Highlights: ► IFMIF test module conceptual design. ► IFMIF test module foreseen instrumentation. ► Cerenkov photon flux monitor. -- Abstract: The preliminary engineering design of the test facilities, including the various test modules to be used in the IFMIF plant is a part of the IFMIF/EVEDA (Engineering Validation and Engineering Design Activities) project from the Broader Approach to fusion. One presents the current status of the conceptual development of the IFMIF Start-Up Monitoring Module, a dedicated device used in the IFMIF test cell during the commissioning phase of the installation, in order to completely characterise the irradiation conditions behind the target on which the beam of deuterons will be focused. This STUMM embarks a lot of instrumentation to precisely characterise the neutron field, the nuclear heating and the temperatures in the test cell. One briefly describes the measuring instruments (including a specific radiation flux monitor under development), the possible layouts and the possible positioning. One also defines which types of measurements are expected by this especially dedicated commissioning module.

  16. Conceptual design study of IFMIF target system

    International Nuclear Information System (INIS)

    Kato, Y.; Nakamura, H.; Ida, M.; Maekawa, H.; Katsuta, H.; Hua, T.; Cevolani, S.

    1997-01-01

    IFMIF-CDA (International Fusion Materials Irradiation Facility - Conceptual Design Activity) had been carried out during 1995 and 1996, under the auspices of the IEA. The mission of this facility is to provide an accelerator based deuterium-lithium (D-Li) neutron source to test the candidate materials of radiation - resistant and low - activation materials up to about a full lifetime of anticipated use in fusion energy reactors. The neutrons of about 14 MeV are obtained by the stripping reaction of the deuteron of Max. 40 MeV with target lithium. Total deuteron beam current is about 250 mA and beam footprint is 20 cm x 5 cm on the free surface of lithium jet. In this report general characteristics of the target lithium system and the results of thermal and flow analysis for the target lithium jet are described. (author)

  17. Hardware availability calculations and results of the IFMIF accelerator facility

    International Nuclear Information System (INIS)

    Bargalló, Enric; Arroyo, Jose Manuel; Abal, Javier; Beauvais, Pierre-Yves; Gobin, Raphael; Orsini, Fabienne; Weber, Moisés; Podadera, Ivan; Grespan, Francesco; Fagotti, Enrico; De Blas, Alfredo; Dies, Javier; Tapia, Carlos; Mollá, Joaquín; Ibarra, Ángel

    2014-01-01

    Highlights: • IFMIF accelerator facility hardware availability analyses methodology is described. • Results of the individual hardware availability analyses are shown for the reference design. • Accelerator design improvements are proposed for each system. • Availability results are evaluated and compared with the requirements. - Abstract: Hardware availability calculations have been done individually for each system of the deuteron accelerators of the International Fusion Materials Irradiation Facility (IFMIF). The principal goal of these analyses is to estimate the availability of the systems, compare it with the challenging IFMIF requirements and find new paths to improve availability performances. Major unavailability contributors are highlighted and possible design changes are proposed in order to achieve the hardware availability requirements established for each system. In this paper, such possible improvements are implemented in fault tree models and the availability results are evaluated. The parallel activity on the design and construction of the linear IFMIF prototype accelerator (LIPAc) provides detailed design information for the RAMI (reliability, availability, maintainability and inspectability) analyses and allows finding out the improvements that the final accelerator could have. Because of the R and D behavior of the LIPAc, RAMI improvements could be the major differences between the prototype and the IFMIF accelerator design

  18. Hardware availability calculations and results of the IFMIF accelerator facility

    Energy Technology Data Exchange (ETDEWEB)

    Bargalló, Enric, E-mail: enric.bargallo-font@upc.edu [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC), Barcelona (Spain); Arroyo, Jose Manuel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, Madrid (Spain); Abal, Javier [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC), Barcelona (Spain); Beauvais, Pierre-Yves; Gobin, Raphael; Orsini, Fabienne [Commissariat à l’Energie Atomique, Saclay (France); Weber, Moisés; Podadera, Ivan [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, Madrid (Spain); Grespan, Francesco; Fagotti, Enrico [Istituto Nazionale di Fisica Nucleare, Legnaro (Italy); De Blas, Alfredo; Dies, Javier; Tapia, Carlos [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC), Barcelona (Spain); Mollá, Joaquín; Ibarra, Ángel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, Madrid (Spain)

    2014-10-15

    Highlights: • IFMIF accelerator facility hardware availability analyses methodology is described. • Results of the individual hardware availability analyses are shown for the reference design. • Accelerator design improvements are proposed for each system. • Availability results are evaluated and compared with the requirements. - Abstract: Hardware availability calculations have been done individually for each system of the deuteron accelerators of the International Fusion Materials Irradiation Facility (IFMIF). The principal goal of these analyses is to estimate the availability of the systems, compare it with the challenging IFMIF requirements and find new paths to improve availability performances. Major unavailability contributors are highlighted and possible design changes are proposed in order to achieve the hardware availability requirements established for each system. In this paper, such possible improvements are implemented in fault tree models and the availability results are evaluated. The parallel activity on the design and construction of the linear IFMIF prototype accelerator (LIPAc) provides detailed design information for the RAMI (reliability, availability, maintainability and inspectability) analyses and allows finding out the improvements that the final accelerator could have. Because of the R and D behavior of the LIPAc, RAMI improvements could be the major differences between the prototype and the IFMIF accelerator design.

  19. Evaluation of RF properties by orifice design for IFMIF RFQ

    International Nuclear Information System (INIS)

    Maebara, Sunao; Sugimoto, Masayoshi

    2005-03-01

    Orifices for the IFMIF RFQ have been designed and fabricated, and RF properties have been evaluated by a network analyzer. The designed orifices were installed into a vacuum port of the 1.1m-long RFQ mock-up module, and the resonant frequency and the phase difference between cavities were measured for a quadrupole operation mode of TE 210 . It was found that the RF properties are not affected on condition that slit direction with the same direction of current flow at the RFQ wall. Orifice conductance from 0.22 to 0.25 m 3 /sec by nitrogen conversion at room temperature was designed, and an ultimate pressure level of 5x10 -7 [Pa] was evaluated for the 4.1m-long central module for the IFMIF RFQ. It was concluded that the designed orifices are effective for RF properties and vacuum conductance in the IFMIF RFQ. (author)

  20. Investigation of IFMIF target assembly structure design

    International Nuclear Information System (INIS)

    Ida, Mizuho; Nakamura, Hiroo; Sugimoto, Masayoshi; Yamamura, Toshio

    2006-10-01

    In the International Fusion Materials Irradiation Facility (IFMIF), the back-wall of target assembly is the part suffered the highest neutron-flux. The back-wall and the assembly are designed to have lips for cutting/welding at the back-wall replacement. To reduce thermal stress and deformation of the back-wall under neutron irradiation, contact pressure between the back-wall and the assembly is one of dominant factors. Therefore, an investigation was performed for feasible clamping pressure of a mechanical clamp set in limited space around the back-wall. It was clarified that the clamp can give a pressure difference up to 0.4 MPa between the contact pressure and atmosphere pressure in the test cell room. Also a research was performed for the dissimilar metal welding in the back-wall. Use of 309 steel was found adequate as the intermediate filler metal through the research of previous welding. Maintaining a temperature of the target assembly so as to avoid a freezing of liquid lithium is needed at the lithium charge into the loop before the beam injection. The assembly is covered with thermal insulation. Therefore, a research and an investigation were performed for compact and light thermal-insulation effective even under helium (i.e. high heat-conduction) condition of the test cell room. The result was as follows; in the case that a thermal conductivity 0.008 W/m·K of one of found insulation materials is available in the temperature range up to 300degC of the IFMIF target assembly, needed thickness and weight of the insulation were respectively only 8.2 mm and 32 kg. Also a research was performed for high-heat-density heaters to maintain temperature of the back-wall which can not be cover with insulation due to limited space. A heater made of silicon-nitride was found to be adequate. Total heat of 8.4 kW on the back-wall was found to be achievable through an investigations of heater arrange. Also an investigation was performed for remote-handling device to

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

    International Nuclear Information System (INIS)

    Sato, Shinichi; Osaki, Toshio; Koga, Shinji

    1996-01-01

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

  2. Present status of the Liquid Breeder Validation Module for IFMIF

    International Nuclear Information System (INIS)

    Casal, Natalia; Mas, Avelino; Mota, Fernando; García, Ángela; Rapisarda, David; Nomen, Oriol; Arroyo, Jose Manuel; Abal, Javier; Mollá, Joaquín; Ibarra, Ángel

    2013-01-01

    Highlights: • The LBVM will be used to perform irradiation experiments on functional materials for fusion reactors. • It houses 16 experimental rigs, each one containing a EUROFER capsule partially filled with lithium lead, at 300–550 °C. • A helium purge gas will sweep the tritium permeated through the capsule walls to a tritium measuring station. • A helium cooling system will keep tritium diffusion within safe margins and guarantee its mechanical integrity. • Thermal hydraulic and mechanical calculations, the module instrumentation and aspects as safety or RAMI are presented. -- Abstract: One of the objectives of IFMIF (International Fusion Materials Irradiation Facility), as stated in its specifications, is the validation of breeder blanket concepts for DEMO design. The so-called Liquid Breeder Validation Module (LBVM) will be used in IFMIF to perform experiments under irradiation on functional materials related to liquid breeder concepts for future fusion reactors. This module, not considered in previous IFMIF design phases, is currently under design by CIEMAT in the framework of the IFMIF/EVEDA project. In this paper, the present status of the design of the LBVM is presented

  3. Present status of the Liquid Breeder Validation Module for IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Casal, Natalia, E-mail: natalia.casal@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Mas, Avelino; Mota, Fernando; García, Ángela; Rapisarda, David [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Nomen, Oriol [Institut de Recerca en Energia de Catalunya (IREC), Barcelona (Spain); Centre de Disseny d’Equips Industrials (CDEI), Technical University of Catalonia (UPC), Barcelona (Spain); Arroyo, Jose Manuel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Abal, Javier [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Mollá, Joaquín; Ibarra, Ángel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain)

    2013-10-15

    Highlights: • The LBVM will be used to perform irradiation experiments on functional materials for fusion reactors. • It houses 16 experimental rigs, each one containing a EUROFER capsule partially filled with lithium lead, at 300–550 °C. • A helium purge gas will sweep the tritium permeated through the capsule walls to a tritium measuring station. • A helium cooling system will keep tritium diffusion within safe margins and guarantee its mechanical integrity. • Thermal hydraulic and mechanical calculations, the module instrumentation and aspects as safety or RAMI are presented. -- Abstract: One of the objectives of IFMIF (International Fusion Materials Irradiation Facility), as stated in its specifications, is the validation of breeder blanket concepts for DEMO design. The so-called Liquid Breeder Validation Module (LBVM) will be used in IFMIF to perform experiments under irradiation on functional materials related to liquid breeder concepts for future fusion reactors. This module, not considered in previous IFMIF design phases, is currently under design by CIEMAT in the framework of the IFMIF/EVEDA project. In this paper, the present status of the design of the LBVM is presented.

  4. 1996 Design effort for IFMIF HEBT

    International Nuclear Information System (INIS)

    Blind, B.

    1997-01-01

    The paper details the 1996 design effort for the IFMIF HEBT. Following a brief overview, it lists the primary requirements for the beam at the target, describes the design approach and design tools used, introduces the beamline modules, gives the results achieved with the design at this stage, points out possible improvements and gives the names and computer locations of the TRACE3-D and PARMILA files that sum up the design work. The design does not fully meet specifications in regards to the flatness of the distribution at the target. With further work, including if necessary some backup options, the flatness specifications may be realized. It is not proposed that the specifications, namely flatness to ±5% and higher-intensity ridges that are no more than 15% above average, be changed at this time. The design also does not meet the requirement that the modules of all beamlines should operate at the same settings. However, the goal of using identical components and operational procedures has been met and only minor returning is needed to produce very similar beam distributions from all beamlines. Significant further work is required in the following areas: TRACE3-D designs and PARMILA runs must be made for the beams coming from accelerators No. 3 and No. 4. Transport of 30-MeV and 35-MeV beams to the targets and beam dump must be studied. Comprehensive error studies must be made. These must result in tolerance specifications and may require design iterations. Detailed interfacing with target-spot instrumentation is required. This instrumentation must be able to check all aspects of the specifications

  5. Design principles of a nuclear and industrial HVAC of IFMIF

    International Nuclear Information System (INIS)

    Pruneri, Giuseppe; Ibarra, A.; Heidinger, R.; Knaster, J.; Sugimoto, M.

    2016-01-01

    Highlights: • Parameter of Derivate air Contamination (DAC) allows to associate the type of air ventilation. • The construction and operation of IFMIF will be subjected to the regulations of the country in which it will be sited. • Structures, systems and components are assigned a particular safety important components (SIC, 1–2 and Non-SIC) clarification that is based on the consequences of their failure. • Reliability, Availability, Maintainability and Inspectability (RAMI) analysis has given a great contribution of the facility to optimize the configuration, particularly for the HVAC system. - Abstract: In 2013, the IFMIF, the International Fusion Material Irradiation Facility, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase, framed by the Broader Approach Agreement between Japan and EURATOM, accomplished in 2013 its mandate to provide the engineering design of the plant on schedule [1]. The IFMIF aims to qualify and characterize materials that are capable of withstanding the intense neutron flux originated in D-T reactions of future fusion reactors due to a neutron flux with a broad peak at 14 MeV, which is able to provide >20 dpa/fpy on small specimens in this EVEDA phase. The successful operation of such a challenging plant demands a careful assessment of the Conventional Facilities (CF), which have adequate redundancies to allow for the target plant availability [2]. The present paper addresses the design proposed in the IFMIF Intermediate Engineering Design Report regarding the CF, particularly the IFMIF's Nuclear and Industrial HVAC design. A preliminary feasibility study, including the initial configuration, calculations and reliability/availability analysis, were performed. The nuclear HVAC design was developed progressively; first, by establishing a conceptual design, starting from the system functional description, followed by the identification of the corresponding interfacing systems and their

  6. Design principles of a nuclear and industrial HVAC of IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Pruneri, Giuseppe [IFMIF/EVEDA, Project Team, Rokkasho (Japan); Ibarra, A. [CIEMAT, Madrid (Spain); Heidinger, R. [F4E, Garching (Germany); Knaster, J. [IFMIF/EVEDA Project Team, Rokkasho (Japan); Sugimoto, M. [JAEA, Rokkasho (Japan)

    2016-02-15

    Highlights: • Parameter of Derivate air Contamination (DAC) allows to associate the type of air ventilation. • The construction and operation of IFMIF will be subjected to the regulations of the country in which it will be sited. • Structures, systems and components are assigned a particular safety important components (SIC, 1–2 and Non-SIC) clarification that is based on the consequences of their failure. • Reliability, Availability, Maintainability and Inspectability (RAMI) analysis has given a great contribution of the facility to optimize the configuration, particularly for the HVAC system. - Abstract: In 2013, the IFMIF, the International Fusion Material Irradiation Facility, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase, framed by the Broader Approach Agreement between Japan and EURATOM, accomplished in 2013 its mandate to provide the engineering design of the plant on schedule [1]. The IFMIF aims to qualify and characterize materials that are capable of withstanding the intense neutron flux originated in D-T reactions of future fusion reactors due to a neutron flux with a broad peak at 14 MeV, which is able to provide >20 dpa/fpy on small specimens in this EVEDA phase. The successful operation of such a challenging plant demands a careful assessment of the Conventional Facilities (CF), which have adequate redundancies to allow for the target plant availability [2]. The present paper addresses the design proposed in the IFMIF Intermediate Engineering Design Report regarding the CF, particularly the IFMIF's Nuclear and Industrial HVAC design. A preliminary feasibility study, including the initial configuration, calculations and reliability/availability analysis, were performed. The nuclear HVAC design was developed progressively; first, by establishing a conceptual design, starting from the system functional description, followed by the identification of the corresponding interfacing systems and their

  7. RFQ Designs and Beam-Loss Distributions for IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Jameson, Robert A [ORNL

    2007-01-01

    The IFMIF 125 mA cw 40 MeV accelerators will set an intensity record. Minimization of particle loss along the accelerator is a top-level requirement and requires sophisticated design intimately relating the accelerated beam and the accelerator structure. Such design technique, based on the space-charge physics of linear accelerators (linacs), is used in this report in the development of conceptual designs for the Radio-Frequency-Quadrupole (RFQ) section of the IFMIF accelerators. Design comparisons are given for the IFMIF CDR Equipartitioned RFQ, a CDR Alternative RFQ, and new IFMIF Post-CDR Equipartitioned RFQ designs. Design strategies are illustrated for combining several desirable characteristics, prioritized as minimum beam loss at energies above ~ 1 MeV, low rf power, low peak field, short length, high percentage of accelerated particles. The CDR design has ~0.073% losses above 1 MeV, requires ~1.1 MW rf structure power, has KP factor 1.7,is 12.3 m long, and accelerates ~89.6% of the input beam. A new Post-CDR design has ~0.077% losses above 1 MeV, requires ~1.1 MW rf structure power, has KP factor 1.7 and ~8 m length, and accelerates ~97% of the input beam. A complete background for the designs is given, and comparisons are made. Beam-loss distributions are used as input for nuclear physics simulations of radioactivity effects in the IFMIF accelerator hall, to give information for shielding, radiation safety and maintenance design. Beam-loss distributions resulting from a ~1M particle input distribution representative of the IFMIF ECR ion source are presented. The simulations reported were performed with a consistent family of codes. Relevant comparison with other codes has not been possible as their source code is not available. Certain differences have been noted but are not consistent over a broad range of designs and parameter range. The exact transmission found by any of these codes should be treated as indicative, as each has various sensitivities in

  8. Effects of the Back Plate Inner Diameter on the Frictional Heat Input and General Performance of Brush Seals

    Directory of Open Access Journals (Sweden)

    Manuel Hildebrandt

    2018-05-01

    Full Text Available Reducing losses in the secondary air system of gas and steam turbines can significantly increase the efficiency of such machines. Meanwhile, brush seals are a widely used alternative to labyrinth seals. Their most valuable advantage over other sealing concepts is the very small gap between the sealing package and the rotor and thus reduced leakage mass flow. This small gap can be achieved due to the great radial flexibility without running the risk of severe detrimental deterioration in case of rubbing. Rubbing between rotor and seal during operation might occur as a result of e.g., an unequal thermal expansion of the rotor and stator or a rotor elongation due to centrifugal forces or manoeuvre forces. Thanks to the flexible structure of the brush seal, the contact forces during a rubbing event are reduced; however, the frictional heat input can still be considerable. Particularly in aircraft engines with their thin and lightweight rotor structures, the permissible material stresses can easily be exceeded by an increased heat input and thus harm the engine’s integrity. The geometry of the seal has a decisive influence on the resulting contact forces and consequently the heat input. This paper is a contribution to further understand the influence of the geometrical parameters of the brush seal on the heat input and the leakage during the rubbing of the seal on the rotor. In this paper, a total of three seals with varied back plate inner diameter are examined in more detail. The experimental tests were carried out on the brush seal test rig of the Institute of Thermal Turbomachinery (ITS under machine-relevant conditions. These are represented by pressure differences of 1 to 7 bar, surface speeds of 30 to 180 m/s and radial interferences of 0.1 to 0.4 mm. For a better interpretation, the results were compared with those obtained at the static test rig of the Institute of Jet Propulsion and Turbomachinery (IFAS at the Technical University of

  9. Radiation effects in IFMIF Li target diagnostic systems

    International Nuclear Information System (INIS)

    Molla, J.; Vila, R.; Shikama, T.; Horiike, H.; Simakov, S.; Ciotti, M.; Ibarra, A.

    2009-01-01

    Diagnostics for the lithium target will be crucial for the operation of IFMIF. Several parameters as the lithium temperature, target thickness or wave pattern must be monitored during operation. Radiation effects may produce malfunctioning in any of these diagnostics due to the exposure to high radiation fields. The main diagnostic systems proposed for the operation of IFMIF are reviewed in this paper from the point of view of radiation damage. The main tools for the assessment of the performance of these diagnostics are the neutronics calculations by using specialised codes and the information accumulated during the last decades on the radiation effects in functional materials, components and diagnostics for ITER. This analysis allows to conclude that the design of some of the diagnostic systems must be revised to assure the high availability required for the target system.

  10. Present status of the conceptual design of IFMIF target facility

    International Nuclear Information System (INIS)

    Katsuta, H.; Kato, Y.; Konishi, S.; Miyauchi, Y.; Smith, D.; Hua, T.; Green, L.; Benamati, G.; Cevolani, S.; Roehrig, H.; Schutz, W.

    1998-01-01

    The conceptual design activity (CDA) for the international fusion materials irradiation facility (IFMIF) has been conducted. For the IFMIF target facility, the conceptual designs of the following two main components have been performed. The design concept of IFMIF utilizes a high energy deuteron beam of 30-40 MeV and total current of 250 mA, impinging on a flowing lithium jet to produce high energy neutrons for irradiation of candidate fusion materials. (1) The target assembly: The kinetic energy of the deuteron beam is deposited on a Li-jet target and neutrons are produced through the d-Li stripping reaction in this target. The assembly is designed to get a stable lithium jet and to prevent the onset of lithium boiling. For 40-MeV deuteron beam (total current of 250 mA) and a beam footprint of 5 x 20 cm 2 lithium jet dimensions are designed to be 2.5 cm thick and 26 cm wide. The lithium jet parameters are given. (2) Lithium loop: The loop circulates the lithium to and from the target assembly and removes the heat deposited by the deuteron beam containing systems for maintaining the-high purity of the lithium required for radiological safety and to minimize corrosion. The maximum lithium flow rate is 130 l/s and the total lithium inventory is about 21 m 3 . The IFMIF policy requires that the lithium loop system be designed to guarantee no combustion of lithium in the event of a lithium leak. This can be achieved by use of multiple confinement of the lithium carrying components. The radioactive waste generated by the target facilities is estimated. (orig.)

  11. Safety design of the international fusion materials irradiation facility (IFMIF)

    International Nuclear Information System (INIS)

    Konishi, Satoshi; Yamaki, Daiju; Katsuta, Hiroji; Moeslang, Anton; Jameson, R.A.; Martone, Marcello; Shannon, T.E.

    1997-11-01

    In the Conceptual Design Activity of the IFMIF, major subsystems, as well as the entire facility is carefully designed to satisfy the safety requirements for any possible construction sites. Each subsystem is qualitatively analyzed to identify possible hazards to the workers, public and environments using Failure Mode and Effect Analysis (FMEA). The results are reflected in the design and operation procedure. Shielding of radiation, particularly neutron around the test cell is one of the most important issue in normal operation. Radiation due to beam halo and activation is a hazard for operation personnel in the accelerator system. For the maintenance, remote handling technology is designed to be applied in various facilities of the IFMIF. Lithium loop and target system hold the majority of the radioactive material in the facility. Tritium and beryllium-7 are generated by the nuclear reaction during operation and thus needed to be removed continuously. They are also the potential hazards of airborne source in off-normal events. Minimization of inventory, separation and immobilization, and multiple confinement are considered in the design. Generation of radioactive waste is anticipated to be minor, but waste treatment systems for gas, liquid and solid wastes are designed to minimize the environmental impact. Lithium leak followed by a fire is a major concern, and extensive prevention plan is made in the target design. One of the design option considered is composed of; primary enclosure of the lithium loop, secondary containment filled with positive pressure argon, and an air tight lithium cell made of concrete with a steel lining. This study will report some technical issues considered in the design of IFMIF. It was concluded that the IFMIF can be designed and constructed to meet or exceed current safely standards for workers, public and the environment with existing technology and reasonable construction cost. (J.P.N.)

  12. IFMIF-CDA technical workshop on lithium target system. Proceedings

    International Nuclear Information System (INIS)

    1995-09-01

    An intense neutron source, International Fusion Materials Irradiation Facility (IFMIF) is planned under the collaborative program by International Energy Agency (IEA), and the Conceptual Design Activity (CDA) started in February 1995. US, Japan and EU are responsible to take a lead in coordinating accelerator, target and test cell design, respectively. In order to exchange the current results of the study and to coordinate the activities for the design integration, the first technical workshop on the lithium target system was held in the period of July 18-21 at the Tokai Research Establishment of the JAERI. This publication summarizes the materials presented in this meeting. The presentations and discussions were organized with the identified CDA tasks. It was confirmed that the reference design of the IFMIF target based on the previous studies under FMIT and ESNIT, elaborated to meet IFMIF parameters, is reasonable and feasible. It was pointed out that the interface between accelerator and test cell subsystems should be carefully investigated to avoid technical conflicts. Some design options such as nozzle, backwall and lithium jet geometry, lithium purity control, and lithium vapor control, based on the current technology were proposed to improve the integral target system function, and further R and D studies were suggested for design integration. (author)

  13. IFMIF-LIPAc Beam Diagnostics. Profiling and Loss Monitoring Systems

    International Nuclear Information System (INIS)

    Egberts, J.

    2012-01-01

    The IFMIF accelerator will accelerate two 125 mA continuous wave (cw) deuteron beams up to 40 MeV and blasts them onto a liquid lithium target to release neutrons. The very high beam power of 10 MW pose unprecedented challenges for the accelerator development. Therefore, it was decided to build a prototype accelerator, the Linear IFMIF Prototype Accelerator (LIPAc), which has the very same beam characteristic, but is limited to 9 MeV only. In the frame of this thesis, diagnostics devices for IFMIF and LIPAc have been developed. The diagnostics devices consist of beam loss monitors and interceptive as well as non-interceptive profile monitors. For the beam loss monitoring system, ionization chambers and diamond detectors have been tested and calibrated for neutron and γ radiation in the energy range expected at LIPAc. During these tests, for the first time, diamond detectors were successfully operated at cryogenic temperatures. For the interceptive profilers, thermal simulations were performed to ensure safe operation. For the non-interceptive profiler, Ionization Profile Monitors (IPMs) were developed. A prototype has been built and tested, and based on the findings, the final IPMs were designed and built. To overcome the space charge of accelerator beam, a software algorithm was written to reconstruct the actual beam profile. (author) [fr

  14. The IFMIF-EVEDA accelerator beam dump design

    International Nuclear Information System (INIS)

    Iglesias, D.; Arranz, F.; Arroyo, J.M.; Barrera, G.; Branas, B.; Casal, N.; Garcia, M.; Lopez, D.; Martinez, J.I.; Mayoral, A.; Ogando, F.; Parro, M.; Oliver, C.; Rapisarda, D.; Sanz, J.; Sauvan, P.; Ibarra, A.

    2011-01-01

    The IFMIF-EVEDA accelerator will be a 9 MeV, 125 mA cw deuteron accelerator prototype for verifying the validity of the 40 MeV accelerator design for IFMIF. A beam dump designed for maximum power of 1.12 MW will be used to stop the beam at the accelerator exit. The conceptual design for the IFMIF-EVEDA accelerator beam dump is based on a conical beam stop made of OFE copper. The cooling system uses an axial high velocity flow of water pressurized up to 3.4 x 10 5 Pa to avoid boiling. The design has been shown to be compliant with ASME mechanical design rules under nominal full power conditions. A sensitivity analysis has been performed to take into account the possible margin on the beam properties at the beam dump entrance. This analysis together with the study of the maintenance issues and the mounting and dismounting operations has led to the complete design definition.

  15. Preliminary analyses of Li jet flows for the IFMIF target

    International Nuclear Information System (INIS)

    Ida, Mizuho; Kato, Yoshio; Nakamura, Hideo; Maekawa, Hiroshi

    1997-03-01

    The characteristics of liquid lithium (Li) plane jet flowing along a concave wall were studied using a multi-dimensional numerical code, FLOW-3D, as part of the two-year conceptual design activity (CDA) of the International Fusion Materials Irradiation Facility (IFMIF) that started in February 1995. The IFMIF will provide high flux, high energy (∼14MeV) neutron irradiation field by deuteron-Li reaction in the Li jet target for testing and development of low-activation and damage-resistant fusion materials. The Li jet target flow at high-velocity (≤ 20m/s) in vacuum, and should adequately remove the intense deuteron beam power (≤ 10MW). The two-dimensional analyses on the thermal and hydraulic responses of the target flow, under the conditions proposed in the IFMIF-CDA, indicated enough temperature margins to avoid significant vaporization and voiding respectively at the jet free surface and the peak temperature location in the jet by keeping the flow stability. (author)

  16. IFMIF LLRF control system architecture based on EPICS

    International Nuclear Information System (INIS)

    Calvo, J.; Ibarra, A.; Miguel Angel Patricio; Rivers, M.

    2012-01-01

    The IFMIF-EVEDA (International Fusion Materials Irradiation Facility - Engineering Validation and Engineering Design Activity) linear accelerator will be a 9 MeV, 125 mA CW (Continuous Wave) deuteron accelerator prototype to validate the technical options of the accelerator design for IFMIF. The primary mission of such facility is to test and verify materials performance when subjected to extensive neutron irradiation of the type encountered in a fusion reactor. The RF (Radio Frequency) power system of IFMIF-EVEDA consists of 18 RF chains working at 175 MHz with three amplification stages each. The LLRF (Low-Level Radio Frequency) controls the amplitude and phase of the signal to be synchronized with the beam and it also controls the resonance frequency of the cavities. The system is based on a commercial cPCI (Compact Peripheral Component Interconnect) FPGA (Field Programmable Gate Array) board provided by Lyrtech and controlled by a Windows Host PC. For this purpose, it is mandatory to communicate the cPCI FPGA Board with an EPICS Channel Access, building an IOC (Input Output Controller). A new software architecture to design a device support, using AsynPortDriver class and CSS as a GUI (Graphical User Interface), is also presented. (authors)

  17. Thermal-hydraulic analysis of bayonet cooling thimble in fuel drain tank of ORNL 10 MW MSRE

    International Nuclear Information System (INIS)

    Sun Lu; Sun Licheng; Yan Changqi

    2012-01-01

    The residual heat removal system of molten salt reactor designed by ORNL, using molten salt as fuel and draining the fuel into fuel drain tank after shutdown of the reactor, removes the decay heat by the circulation of water through the bayonet cooling thimbles in the fuel drain tank. According to structural features of the bayonet cooling thimbles in ORNL 10 MW molten salt reactor experiment (MSRE), this paper presents the analytical results of the influence of the width of gas gap and the width of steam riser on the heat removal ability and the natural circulation of the cooling water, etc. The analysis results show that, when the width of gas gap range from 3.1 mm to 5.1 mm, the change of heat dissipation power and natural circulation flow rate are both less than 5%; when the width of steam riser changes from 3.6 mm to 5.1 mm, the flow mass of the natural circulation change from 1.9 kg/s to 4.79 kg/s, with a slightly effect on the heat transfer efficiency of the system. (authors)

  18. Availability simulation software adaptation to the IFMIF accelerator facility RAMI analyses

    International Nuclear Information System (INIS)

    Bargalló, Enric; Sureda, Pere Joan; Arroyo, Jose Manuel; Abal, Javier; De Blas, Alfredo; Dies, Javier; Tapia, Carlos; Mollá, Joaquín; Ibarra, Ángel

    2014-01-01

    Highlights: • The reason why IFMIF RAMI analyses needs a simulation is explained. • Changes, modifications and software validations done to AvailSim are described. • First IFMIF RAMI results obtained with AvailSim 2.0 are shown. • Implications of AvailSim 2.0 in IFMIF RAMI analyses are evaluated. - Abstract: Several problems were found when using generic reliability tools to perform RAMI (Reliability Availability Maintainability Inspectability) studies for the IFMIF (International Fusion Materials Irradiation Facility) accelerator. A dedicated simulation tool was necessary to model properly the complexity of the accelerator facility. AvailSim, the availability simulation software used for the International Linear Collider (ILC) became an excellent option to fulfill RAMI analyses needs. Nevertheless, this software needed to be adapted and modified to simulate the IFMIF accelerator facility in a useful way for the RAMI analyses in the current design phase. Furthermore, some improvements and new features have been added to the software. This software has become a great tool to simulate the peculiarities of the IFMIF accelerator facility allowing obtaining a realistic availability simulation. Degraded operation simulation and maintenance strategies are the main relevant features. In this paper, the necessity of this software, main modifications to improve it and its adaptation to IFMIF RAMI analysis are described. Moreover, first results obtained with AvailSim 2.0 and a comparison with previous results is shown

  19. Availability simulation software adaptation to the IFMIF accelerator facility RAMI analyses

    Energy Technology Data Exchange (ETDEWEB)

    Bargalló, Enric, E-mail: enric.bargallo-font@upc.edu [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Sureda, Pere Joan [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Arroyo, Jose Manuel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, Madrid (Spain); Abal, Javier; De Blas, Alfredo; Dies, Javier; Tapia, Carlos [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Mollá, Joaquín; Ibarra, Ángel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, Madrid (Spain)

    2014-10-15

    Highlights: • The reason why IFMIF RAMI analyses needs a simulation is explained. • Changes, modifications and software validations done to AvailSim are described. • First IFMIF RAMI results obtained with AvailSim 2.0 are shown. • Implications of AvailSim 2.0 in IFMIF RAMI analyses are evaluated. - Abstract: Several problems were found when using generic reliability tools to perform RAMI (Reliability Availability Maintainability Inspectability) studies for the IFMIF (International Fusion Materials Irradiation Facility) accelerator. A dedicated simulation tool was necessary to model properly the complexity of the accelerator facility. AvailSim, the availability simulation software used for the International Linear Collider (ILC) became an excellent option to fulfill RAMI analyses needs. Nevertheless, this software needed to be adapted and modified to simulate the IFMIF accelerator facility in a useful way for the RAMI analyses in the current design phase. Furthermore, some improvements and new features have been added to the software. This software has become a great tool to simulate the peculiarities of the IFMIF accelerator facility allowing obtaining a realistic availability simulation. Degraded operation simulation and maintenance strategies are the main relevant features. In this paper, the necessity of this software, main modifications to improve it and its adaptation to IFMIF RAMI analysis are described. Moreover, first results obtained with AvailSim 2.0 and a comparison with previous results is shown.

  20. Cooling system for the IFMIF-EVEDA radiofrequency system

    International Nuclear Information System (INIS)

    Perez Pichel, G. D.

    2012-01-01

    The IFMIF-EVEDA project consists on an accelerator prototype that will be installed at Rokkasho (Japan). Through CIEMAT, that is responsible of the development of many systems and components. Empresarios Agrupados get the responsibility of the detailed design of the cooling system for the radiofrequency system (RF system) that must feed the accelerator. the RF water cooling systems is the water primary circuit that provides the required water flow (with a certain temperature, pressure and water quality) and also dissipates the necessary thermal power of all the radiofrequency system equipment. (Author) 4 refs.

  1. Deuteron-induced activation data in EAF for IFMIF calculations

    International Nuclear Information System (INIS)

    Forrest, R.; Cook, I.

    2006-01-01

    The main type of activation calculations needed for fusion technology deals with the interaction of neutrons with materials. The road map for development of fusion as an electricity producing technology is based on ITER and IFMIF followed by DEMO. IFMIF is a materials testing facility that will enable materials planned to be used in DEMO to be irradiated to very high fluences, so providing the database of material properties required for the licensing of DEMO. IFMIF will use intense beams of high energy deuterons striking a flowing lithium target to produce the neutron field. Although the neutron spectrum is a good match to those produced in a D-T fusion device, there is a significant high energy tail extending up to 55 MeV. These high energy neutrons were the motivation for increasing the upper energy limit in the neutron-induced part of EAF-2005 so that activation calculations could be made in IFMIF. The deuterons themselves will also make a contribution to activation especially in the target where they strike the lithium but also due to beam losses in the accelerator. It was realised that because of corrosion in the lithium loop there is the potential for a wide range of elements to be present in the target region and it is therefore necessary to have a complete library of deuteron-induced cross section data, just as in the neutron case. A preliminary library based on model calculations with TALYS using global parameters was used to construct a deuteron-induced library and this was released as part of the maintenance release of EAF-2005.1 at the beginning of this year. This data library has been used with an updated version of the inventory code FISPACT to calculate the activation in the lithium target due to reactions of the deuterons with the corrosion products. These calculations show that deuterons are much more important than neutrons (about a factor of 70) in activating the elements other than lithium. This work shows the importance of the effect and means

  2. IFMIF (International Fusion Materials Irradiation Facility) key element technology phase interim report

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-03-01

    Activities of International Fusion Materials Irradiation Facility (IFMIF) have been performed under an IEA collaboration since 1995. IFMIF is an accelerator-based deuteron (D{sup +})-lithium (Li) neutron source designed to produce an intense neutron field (2 MW/m{sup 2}, 20 dpa/year for Fe) in a volume of 500 cm{sup 3} for testing candidate fusion materials. In 2000, a 3 year Key Element technology Phase (KEP) of IFMIF was started to reduce the key technology risk factors. This interim report summarizes the KEP activities until mid 2001 in the major project work-breakdown areas of accelerator, target, test facilities and design integration. (author)

  3. Results of the RAMI analyses performed for the IFMIF accelerator facility in the engineering design phase

    Energy Technology Data Exchange (ETDEWEB)

    Bargalló, Enric, E-mail: enric.bargallo@esss.se [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Arroyo, Jose Manuel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, Madrid (Spain); Abal, Javier; Dies, Javier; De Blas, Alfredo; Tapia, Carlos [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Moya, Joaquin; Ibarra, Angel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, Madrid (Spain)

    2015-10-15

    Highlights: • RAMI methodology used for IFMIF accelerator facility is presented. • Availability analyses and results are shown. • Main accelerator design changes are proposed. • Consequences and conclusions of the RAMI analyses are described. - Abstract: This paper presents a summary of the RAMI (Reliability Availability Maintainability Inspectability) analyses done for the IFMIF (International Fusion Materials Irradiation Facility) Accelerator facility in the Engineering Design Phase. The methodology followed, the analyses performed, the results obtained and the conclusions drawn are described. Moreover, the consequences of the incorporation of the RAMI studies in the IFMIF design are presented and the main outcomes of these analyses are shown.

  4. Results of the RAMI analyses performed for the IFMIF accelerator facility in the engineering design phase

    International Nuclear Information System (INIS)

    Bargalló, Enric; Arroyo, Jose Manuel; Abal, Javier; Dies, Javier; De Blas, Alfredo; Tapia, Carlos; Moya, Joaquin; Ibarra, Angel

    2015-01-01

    Highlights: • RAMI methodology used for IFMIF accelerator facility is presented. • Availability analyses and results are shown. • Main accelerator design changes are proposed. • Consequences and conclusions of the RAMI analyses are described. - Abstract: This paper presents a summary of the RAMI (Reliability Availability Maintainability Inspectability) analyses done for the IFMIF (International Fusion Materials Irradiation Facility) Accelerator facility in the Engineering Design Phase. The methodology followed, the analyses performed, the results obtained and the conclusions drawn are described. Moreover, the consequences of the incorporation of the RAMI studies in the IFMIF design are presented and the main outcomes of these analyses are shown.

  5. IFMIF (International Fusion Materials Irradiation Facility) key element technology phase interim report

    International Nuclear Information System (INIS)

    Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki

    2002-03-01

    Activities of International Fusion Materials Irradiation Facility (IFMIF) have been performed under an IEA collaboration since 1995. IFMIF is an accelerator-based deuteron (D + )-lithium (Li) neutron source designed to produce an intense neutron field (2 MW/m 2 , 20 dpa/year for Fe) in a volume of 500 cm 3 for testing candidate fusion materials. In 2000, a 3 year Key Element technology Phase (KEP) of IFMIF was started to reduce the key technology risk factors. This interim report summarizes the KEP activities until mid 2001 in the major project work-breakdown areas of accelerator, target, test facilities and design integration. (author)

  6. Cooling system for the IFMIF-EVEDA radiofrequency system; Sistema de refrigeracion del sistema de radiofrecuencia del IFMIF-EVEDA

    Energy Technology Data Exchange (ETDEWEB)

    Perez Pichel, G. D.

    2012-07-01

    The IFMIF-EVEDA project consists on an accelerator prototype that will be installed at Rokkasho (Japan). Through CIEMAT, that is responsible of the development of many systems and components. Empresarios Agrupados get the responsibility of the detailed design of the cooling system for the radiofrequency system (RF system) that must feed the accelerator. the RF water cooling systems is the water primary circuit that provides the required water flow (with a certain temperature, pressure and water quality) and also dissipates the necessary thermal power of all the radiofrequency system equipment. (Author) 4 refs.

  7. Neutronics of the IFMIF neutron source: development and analysis

    International Nuclear Information System (INIS)

    Wilson, P.P.H.

    1999-01-01

    The accurate analysis of this system required the development of a code system and methodology capable of modelling the various physical processes. A generic code system for the neutronics analysis of neutron sources has been created by loosely integrating existing components with new developments: the data processing code NJOY, the Monte Carlo neutron transport code MCNP, and the activation code ALARA were supplemented by a damage data processing program, damChar, and integrated with a number of flexible and extensible modules for the Perl scripting language. Specific advances were required to apply this code system to IFMIF. Based on the ENDF-6 data format requirements of this system, new data evaluations have been implemented for neutron transport and activation. Extensive analysis of the Li(d, xn) reaction has led to a new MCNP source function module, M c DeLi, based on physical reaction models and capable of accurate and flexible modelling of the IFMIF neutron source term. In depth analyses of the neutron flux spectra and spatial distribution throughout the high flux test region permitted a basic validation of the tools and data. The understanding of the features of the neutron flux provided a foundation for the analyses of the other neutron responses. (orig./DGE) [de

  8. ENERGY EFFICIENCY LIMITS FOR A RECUPERATIVE BAYONET SULFURIC ACID DECOMPOSITION REACTOR FOR SULFUR CYCLE THERMOCHEMICAL HYDROGEN PRODUCTION

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.; Edwards, T.

    2009-06-11

    A recuperative bayonet reactor design for the high-temperature sulfuric acid decomposition step in sulfur-based thermochemical hydrogen cycles was evaluated using pinch analysis in conjunction with statistical methods. The objective was to establish the minimum energy requirement. Taking hydrogen production via alkaline electrolysis with nuclear power as the benchmark, the acid decomposition step can consume no more than 450 kJ/mol SO{sub 2} for sulfur cycles to be competitive. The lowest value of the minimum heating target, 320.9 kJ/mol SO{sub 2}, was found at the highest pressure (90 bar) and peak process temperature (900 C) considered, and at a feed concentration of 42.5 mol% H{sub 2}SO{sub 4}. This should be low enough for a practical water-splitting process, even including the additional energy required to concentrate the acid feed. Lower temperatures consistently gave higher minimum heating targets. The lowest peak process temperature that could meet the 450-kJ/mol SO{sub 2} benchmark was 750 C. If the decomposition reactor were to be heated indirectly by an advanced gas-cooled reactor heat source (50 C temperature difference between primary and secondary coolants, 25 C minimum temperature difference between the secondary coolant and the process), then sulfur cycles using this concept could be competitive with alkaline electrolysis provided the primary heat source temperature is at least 825 C. The bayonet design will not be practical if the (primary heat source) reactor outlet temperature is below 825 C.

  9. IFMIF, a fusion relevant neutron source for material irradiation current status

    International Nuclear Information System (INIS)

    Knaster, J.; Chel, S.; Fischer, U.; Groeschel, F.; Heidinger, R.; Ibarra, A.; Micciche, G.; Möslang, A.; Sugimoto, M.; Wakai, E.

    2014-01-01

    The d-Li based International Fusion Materials Irradiation Facility (IFMIF) will provide a high neutron intensity neutron source with a suitable neutron spectrum to fulfil the requirements for testing and qualifying fusion materials under fusion reactor relevant irradiation conditions. The IFMIF project, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the Broader Approach (BA) Agreement between Japan Government and EURATOM, aims at the construction and testing of the most challenging facility sub-systems, such as the first accelerator stage, the Li target and loop, and irradiation test modules, as well as the design of the entire facility, thus to be ready for the IFMIF construction with a clear understanding of schedule and cost at the termination of the BA mid-2017. The paper reviews the IFMIF facility and its principles, and reports on the status of the EVEDA activities and achievements

  10. IFMIF-KEP. International fusion materials irradiation facility key element technology phase report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-03-01

    The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m{sup 2}, 20 dpa/y in Fe, in a volume of 500 cm{sup 3} and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration. (author)

  11. IFMIF-KEP. International fusion materials irradiation facility key element technology phase report

    International Nuclear Information System (INIS)

    2003-03-01

    The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m 2 , 20 dpa/y in Fe, in a volume of 500 cm 3 and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration. (author)

  12. 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.

  13. Tritium transport calculations for the IFMIF Tritium Release Test Module

    Energy Technology Data Exchange (ETDEWEB)

    Freund, Jana, E-mail: jana.freund@kit.edu; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

    2014-10-15

    Highlights: • Delivery of material data for the tritium balance in the IFMIF Tritium Release Test Module. • Description of the topological models in TMAP and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). • Computation of release of tritium from the breeder solid material into the purge gas. • Computation of the loss of tritium over the capsule wall, rig hull, container wall and purge gas return line. - Abstract: The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the

  14. Tritium transport calculations for the IFMIF Tritium Release Test Module

    International Nuclear Information System (INIS)

    Freund, Jana; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

    2014-01-01

    Highlights: • Delivery of material data for the tritium balance in the IFMIF Tritium Release Test Module. • Description of the topological models in TMAP and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). • Computation of release of tritium from the breeder solid material into the purge gas. • Computation of the loss of tritium over the capsule wall, rig hull, container wall and purge gas return line. - Abstract: The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the

  15. IFMIF - Design Study for in Situ Creep Fatigue Tests

    International Nuclear Information System (INIS)

    Gordeev, S.; Heinzel, V.; Simakov, St.; Stratmanns, E.; Vladimirov, P.; Moeslang, A.

    2006-01-01

    While the high flux volume (20-50 dpa/fpy) of the International Fusion Materials Irradiation Facility (IFMIF) is dedicated to the irradiation of ∼ 1100 qualified specimens that will be post irradiation examined after disassembling in dedicated Hot Cells, various in situ experiments are foreseen in the medium flux volume (1-20 dpa/fpy). Of specific importance for structural lifetime assessments of fusion power reactors are instrumented in situ creep-fatigue experiments, as they can simulate realistically a superposition of thermal fatigue or creep fatigue and irradiation with fusion relevant neutrons. Based on former experience with in situ fatigue tests under high energy light ion irradiation, a design study has been performed to evaluate the feasibility of in situ creep fatigue tests in the IFMIF medium flux position. The vertically arranged test module for such experiments consists basically of a frame similar to a universal testing machine, but equipped with three pulling rods, driven by independent step motors, instrumentation systems and specimen cooling systems. Therefore, three creep fatigue specimens may be tested at one time in this apparatus. Each specimen is a hollow tube with coolant flow in the specimen interior to maintain individual specimen temperatures. The recently established IFMIF global 3D geometry model was used together the latest McDeLicious code for the neutral and charged particle transport calculations. These comprehensive neutronics calculations have been performed with a fine special resolution of 0.25 cm 3 , showing among others that the specimens will be irradiated with a homogeneous damage rate of up to 13(∼ 9%) dpa/fpy and a fusion relevant damage to helium ratio of 10-12 appm He/dpa. In addition, damage and gas production rates as well as the heat deposition in structural parts of the test module have been calculated. Despite of the vertical gradients in the nuclear heating, CFD code calculations with STAR-CD revealed very

  16. RAMI strategies in the IFMIF Test Facilities design

    Energy Technology Data Exchange (ETDEWEB)

    Abal, Javier, E-mail: javier.abal@upc.edu [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Dies, Javier [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Arroyo, José Manuel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Bargalló, Enric [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Casal, Natalia; García, Ángela [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Martínez, Gonzalo; Tapia, Carlos; De Blas, Alfredo [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Mollá, Joaquín; Ibarra, Ángel [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain)

    2013-10-15

    Highlights: • We have implemented fault tolerant design strategies so that the strong availability requirements are met. • The evolution to the present design of the signal and cooling lines inside the TTC has also been compared. • The RAMI analyses have demonstrated a strong capability in being a complementary tool in the design of IFMIF Test Facilities. -- Abstract: In this paper, a RAMI analysis of the different stages in Test Facilities (TF) design is described. The comparison between the availability results has been a milestone not only to evaluate the major unavailability contributors in the updates but also to implement fault tolerant design strategies when possible. These strategies encompass a wide range of design activities: from the definition of degraded modes of operation in the Test Facilities to specific modifications in the test modules in order to guarantee their fail safe operation.

  17. RAMI strategies in the IFMIF Test Facilities design

    International Nuclear Information System (INIS)

    Abal, Javier; Dies, Javier; Arroyo, José Manuel; Bargalló, Enric; Casal, Natalia; García, Ángela; Martínez, Gonzalo; Tapia, Carlos; De Blas, Alfredo; Mollá, Joaquín; Ibarra, Ángel

    2013-01-01

    Highlights: • We have implemented fault tolerant design strategies so that the strong availability requirements are met. • The evolution to the present design of the signal and cooling lines inside the TTC has also been compared. • The RAMI analyses have demonstrated a strong capability in being a complementary tool in the design of IFMIF Test Facilities. -- Abstract: In this paper, a RAMI analysis of the different stages in Test Facilities (TF) design is described. The comparison between the availability results has been a milestone not only to evaluate the major unavailability contributors in the updates but also to implement fault tolerant design strategies when possible. These strategies encompass a wide range of design activities: from the definition of degraded modes of operation in the Test Facilities to specific modifications in the test modules in order to guarantee their fail safe operation

  18. CEA-DSM-DAPNIA-SACM contribution to IFMIF 2004

    International Nuclear Information System (INIS)

    2005-01-01

    A neutron source from the D-Li stripping reaction has been selected as the basic concept of the international fusion materials irradiation facility (IFMIF). The objective requires generation by a linear accelerator of 250 mA continuous current of deuterons at a nominal energy of 40 MeV, with provision for operation at lower energy. The basic approach is to provide 2 linacs modules, each delivering 125 mA to a common lithium target. This approach presents availability and operational flexibility advantages. This paper reviews the technical choices that have been made for different components such as the ECR source, beam diagnostics, the radio frequency quadrupole, the RF system, the drift-tube-linac and the high energy beam transport line. (A.C.)

  19. Qualification of Sub-Atmospheric Pressure Sensors for the Cryomagnet Bayonet Heat Exchangers of the Large Hadron Collider

    Science.gov (United States)

    Bager, T.; Casas-Cubillos, J.; Jeanmonod, N.

    2006-04-01

    The superconducting magnets of the Large Hadron Collider (LHC) will be cooled at 1.9 K by distributed cooling loops working with saturated two-phase superfluid helium flowing in 107 m long bayonet heat exchangers located in each magnet cold-mass cell. The temperature of the magnets could be difficult to control because of the large dynamic heat load variations. Therefore, it is foreseen to measure the heat exchangers pressure to feed the regulation loops with the corresponding saturation temperature. The required uncertainty of the sub-atmospheric saturation pressure measurement shall be of the same order of the one associated to the magnet thermometers, in pressure it translates as ±5 Pa at 1.6 kPa. The transducers shall be radiation hard as they will endure, in the worst case, doses up to 10 kGy and 1015 neutronsṡcm-2 over 10 years. The sensors under evaluation were installed underground in the dump section of the SPS accelerator with a radiation environment close to the one expected for the LHC. The monitoring equipment was installed in a remote radiation protected area. This paper presents the results of the radiation qualification campaign with emphasis on the reliability and accuracy of the pressure sensors under the test conditions.

  20. Qualification of Sub-atmospheric Pressure Sensors for the Cryomagnet Bayonet Heat Exchangers of the Large Hadron Collider

    CERN Document Server

    Jeanmonod, N; Casas-Cubillos, J

    2006-01-01

    The superconducting magnets of the Large Hadron Collider (LHC) will be cooled at 1.9 K by distributed cooling loops working with saturated two-phase superfluid helium flowing in 107 m long bayonet heat exchangers [1] located in each magnet cold-mass cell. The temperature of the magnets could be difficult to control because of the large dynamic heat load variations. Therefore, it is foreseen to measure the heat exchangers pressure to feed the regulation loops with the corresponding saturation temperature. The required uncertainty of the sub-atmospheric saturation pressure measurement shall be of the same order of the one associated to the magnet thermometers, in pressure it translates as ±5 Pa at 1.6 kPa. The transducers shall be radiation hard as they will endure, in the worst case, doses up to 10 kGy and 10**15 neutrons·cm**-2 over 10 years. The sensors under evaluation were installed underground in the dump section of the SPS accelerator with a radiation environment close to the one expected for the L...

  1. An efficient hybrid sulfur process using PEM electrolysis with a bayonet decomposition reactor - HTR2008-58207

    International Nuclear Information System (INIS)

    Gorensek, M. B.; Summers, W. A.; Lahoda, E. J.; Bolthrunis, C. O.; Greyvenstein, R.

    2008-01-01

    The Hybrid Sulfur (HyS) Process is being developed to produce hydrogen by water-splitting using heat from advanced nuclear reactors. It has the potential for high efficiency and competitive hydrogen production cost, and has been demonstrated at a laboratory scale. As a two-step process, the HyS is one of the simplest thermochemical cycles. The sulfuric acid decomposition reaction is common to all sulfur cycles, including the Sulfur-Iodine (SI) cycle. What distinguishes the HyS Process from the other sulfur cycles is the use of sulfur dioxide (SO 2 ) to depolarize the anode of a water electrolyzer. The two critical HyS Process components are the SO 2 - depolarized electrolyzer (SDE), and the high-temperature decomposition reactor. A proton exchange membrane (PEM)- type SDE and a silicon carbide bayonet-type high-temperature decomposition reactor are being developed for DOE's Nuclear Hydrogen Initiative (NHI) by Savannah River National Laboratory (SRNL) and by Sandia National Laboratories (SNL), respectively. The ultimate goal of the NHI-sponsored work is to couple the SDE and the reactor in an integrated laboratory scale experiment to prove the technical readiness of the HyS cycle for the NGNP demonstration. This paper describes the flowsheet that is being prepared to combine these two components into a viable process and presents the latest performance projections and economics for a HyS Process coupled to a PBMR heat source. The basic flowsheet for this process has been described elsewhere [4]. It requires an acid concentration section because the SDE product, which is limited to no more than 50% H 2 SO 4 by cell voltage considerations, is too dilute to be fed directly to the bayonet, which needs at least 65% H 2 SO 4 in the feed for acceptable performance. Optimization involved trade-offs between decomposition reaction and acid concentration heat requirements. The PBMR heat source can split its heat output between the decomposition reaction and either steam

  2. Overview of the main challenges for the engineering design of the test facilities system of IFMIF

    International Nuclear Information System (INIS)

    Molla, J.; Nakamura, K.

    2009-01-01

    High intense radiation fields were demanded to IFMIF to address the lack of information on effects in materials due to radiation fields with fusion reactor features. Such intense radiation fields will also produce a number of unwanted effects in exposed materials and components. The main difficulties to achieve a reliable engineering design of the Test Facilities System during the Engineering Validation and the Engineering Design phase of IFMIF now under development are reviewed in this paper. The most challenging activities will be the design of the high flux test module, the creep fatigue test module, the test cell and the remote handling system. The intense radiation fields in the irradiation area and the high availability required for IFMIF (70%) are the main reasons for these difficulties.

  3. Numerical examination of temperature control in helium-cooled high flux test module of IFMIF

    International Nuclear Information System (INIS)

    Ebara, Shinji; Yokomine, Takehiko; Shimizu, Akihiko

    2007-01-01

    For long term irradiation of the International Fusion Materials Irradiation Facility (IFMIF), test specimens are needed to retain constant temperature to avoid change of its irradiation characteristics. The constant temperatures control is one of the most challenging issues for the IFMIF test facilities. We have proposed a new concept of test module which is capable of precisely measuring temperature, keeping uniform temperature with enhanced cooling performance. In the system according to the new design, cooling performances and temperature distributions of specimens were examined numerically under diverse conditions. Some transient behaviors corresponding to the prescribed temperature control mode were perseveringly simulated. It was confirmed that the thermal characteristics of the new design satisfied the severe requirement of IFMIF

  4. Irradiation facilities for materials research: IFMIF and small scale installations

    International Nuclear Information System (INIS)

    Perlado, J. M.; Victoria, M.

    2007-01-01

    The research of advance materials in nuclear fields such as new fission reactors (Generation-IV), Accelerator Driven Systems for Transmutation of Radioactive Wastes and Nuclear Fusion, is becoming very much common in the types of low activation and radiation resistant Materials. Ferritic-Martensitic Steels (based in 9-12 Cr) with or without Oxide Dispersion Techniques (Ytria Nanoparticles), Composites materials are becoming the new generation to answer requirements of high temperature, high radiation resistance of structural materials. Special dedication is appearing in general research programmes to this area of Materials. The understanding of their final performance needs a wider knowledge of the mechanisms of radiation damage in these materials from the atomistic scale to the macroscopic responses. New extensive campaigns are being funded to irradiate from simple elements to model alloys and finally the complex materials themselves. That sequence and its state of art will be presented One clear technique for that understanding is the Multi scale Modelling which includes simulation techniques from quantum mechanics, molecular dynamics, defects diffusion, mesoscopic modelling and finally the macroscopic constitutive relations for macroscopic analysis. However, in each one of these steps is necessary a systematic and well established program of experiments that combines the irradiation and the very detailed analysis with techniques such as Transmission Electron Microscope, Positron Annihilation, SIMS, Atom Probe, Nanoindebntation. A key aspect that wants to be presented in this work is the state of art and discussion of Irradiation Facilities for Materials studies. Those facilities goes from ion implantation sources, small accelerator, Experimental Reactors such High Flux Reactor, sophisticated Triple Beams Sources as JANNUS in France to generate at the same time displacements-hydrogen-helium, and projected very large neutron installation such as IFMIF. The role to

  5. The European contribution to the development and validation activities for the design of IFMIF lithium facility

    Energy Technology Data Exchange (ETDEWEB)

    Miccichè, Gioacchino, E-mail: gioacchino.micciche@enea.it [EURATOM-ENEA, CR Brasimone I-40035 Camugnano, BO (Italy); Aiello, Antonio; Bernardi, Davide; Favuzza, Paolo; Agostini, Pietro [EURATOM-ENEA, CR Brasimone I-40035 Camugnano, BO (Italy); Frisoni, Manuela [EURATOM-ENEA, CR Bologna I-40129, BO Italy (Italy); Pinna, Tonio; Porfiri, MariaTeresa [EURATOM-ENEA, CR Frascati I-0044 Frascati, Roma (Italy); Tincani, Amelia [EURATOM-ENEA, CR Brasimone I-40035 Camugnano, BO (Italy); Di Maio, PieroAlessandro [University of Palermo, I-90128 Palermo (Italy); Knaepen, Bernard [Université libre de Bruxelles, I-1050 Bruxelles (Belgium)

    2013-10-15

    Highlights: • Engineering design of the target assembly. • Erosion, corrosion phenomena promoted by the lithium are studied. • Purification system implemented in the LiFus6 loop. • Study of the remote handling maintenance for the IFMIF TA. -- Abstract: The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-driven intense neutron source where candidate materials for fusion reactors will be tested and validated. The high energy neutron flux is produced by means of two deuteron beams (total current of 250 mA, energy of 40 MeV) that strikes a liquid lithium target circulating in a lithium loop of IFMIF plant. The European (EU) contribution to the development of the lithium facility comprises five procurement packages, as follow: (1) participation to the experimental activities of the EVEDA lithium test loop in Oarai (Japan); (2) study aimed at evaluating the corrosion and erosion phenomena, promoted by lithium, for structural fusion reference materials like AISI 316L and Eurofer; (3) design and validation of the lithium purification method with the aim to provide input data for the design of the purification system of IFIMF lithium loop; (4) design and validation of the remote handling (RH) procedures for the refurbishment/replacement of the EU concept of IFMIF target assembly including the design of the remote handling tools; (5) the engineering design of the European target assembly for IFMIF and the safety and RAMI analyses for the entire IFMIF lithium facility. The paper gives an overview of the status of the activities and of the main outcomes achieved so far.

  6. The European contribution to the development and validation activities for the design of IFMIF lithium facility

    International Nuclear Information System (INIS)

    Miccichè, Gioacchino; Aiello, Antonio; Bernardi, Davide; Favuzza, Paolo; Agostini, Pietro; Frisoni, Manuela; Pinna, Tonio; Porfiri, MariaTeresa; Tincani, Amelia; Di Maio, PieroAlessandro; Knaepen, Bernard

    2013-01-01

    Highlights: • Engineering design of the target assembly. • Erosion, corrosion phenomena promoted by the lithium are studied. • Purification system implemented in the LiFus6 loop. • Study of the remote handling maintenance for the IFMIF TA. -- Abstract: The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-driven intense neutron source where candidate materials for fusion reactors will be tested and validated. The high energy neutron flux is produced by means of two deuteron beams (total current of 250 mA, energy of 40 MeV) that strikes a liquid lithium target circulating in a lithium loop of IFMIF plant. The European (EU) contribution to the development of the lithium facility comprises five procurement packages, as follow: (1) participation to the experimental activities of the EVEDA lithium test loop in Oarai (Japan); (2) study aimed at evaluating the corrosion and erosion phenomena, promoted by lithium, for structural fusion reference materials like AISI 316L and Eurofer; (3) design and validation of the lithium purification method with the aim to provide input data for the design of the purification system of IFIMF lithium loop; (4) design and validation of the remote handling (RH) procedures for the refurbishment/replacement of the EU concept of IFMIF target assembly including the design of the remote handling tools; (5) the engineering design of the European target assembly for IFMIF and the safety and RAMI analyses for the entire IFMIF lithium facility. The paper gives an overview of the status of the activities and of the main outcomes achieved so far

  7. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Executive summary

    International Nuclear Information System (INIS)

    1997-01-01

    This report is a summary of the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member

  8. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Executive summary

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

    This report is a summary of the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member.

  9. Hydraulics and heat transfer in the IFMIF liquid lithium target: CFD calculations

    OpenAIRE

    Peña, A.; Esteban, G.A.; Sancho, J.; Kolesnik, V.; Abánades Velasco, Alberto

    2009-01-01

    CFD (Computational fluid dynamics) calculation turns out to be a good approximation to the real behavior of the lithium (Li) flow of the target of the international fusion materials irradiation facility (IFMIF). A three-dimensional (3D) modelling of the IFMIF design Li target assembly, made with the CFD commercial code ANSYS-FLUENT has been carried out. The simulation by a structural mesh is focused on the thermal-hydraulic analysis inside the Li jet flow. For, this purpose, the two deuteron ...

  10. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Final report

    International Nuclear Information System (INIS)

    Martone, M.

    1997-01-01

    This report documents the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member

  11. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Martone, M [ENEA, Centro Ricerche Frascati, Rome (Italy)

    1997-01-01

    This report documents the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member.

  12. Safety managements of the linear IFMIF/EVEDA prototype accelerator

    International Nuclear Information System (INIS)

    Takahashi, Hiroki; Maebara, Sunao; Kojima, Toshiyuki; Narita, Takahiro; Tsutsumi, Kazuyoshi; Sakaki, Hironao; Suzuki, Hiromitsu; Sugimoto, Masayoshi

    2014-01-01

    Highlights: •Safety management is needed to secure the personnel safety from high dose rate. •The management of access to the accelerator vault is mainly performed by PPS. •The operation management is needed for safety during Injector and RFQ commissioning. •Pulse Duty Management system is newly developed for Injector commissioning for operation management. •PDM system is useful to reduce the radioactivation of equipment and the radiation exposure during and after beam operation. -- Abstract: On the Linear IFMIF/EVEDA Prototype Accelerator (LIPAc), the validation up to 9 MeV deuteron beam with 125 mA continuous wave is planned in Rokkasho, Aomori, Japan. Since the deuteron beam power exceeds 1 MW, safety issue related to γ-ray and neutron production is critical. To establish the safety management indispensable to reduce radiation exposure for personnel and activation of accelerator equipment, Personnel Protection System (PPS) of LIPAc control system, which works together with Radiation Monitoring System and Access Control System, was developed for LIPAc. The management of access to the accelerator vault by PPS and the beam duty management of PPS are presented in details

  13. Experimental investigation of the IFMIF target mock-up

    International Nuclear Information System (INIS)

    Loginov, N.; Mikheyev, A.; Morozov, V.; Aksenov, Yu.; Arnol'dov, M.; Berensky, L.; Fedotovsky, V.; Chernov, V.; Nakamura, H.

    2009-01-01

    The international fusion materials irradiation facility (IFMIF) lithium neutron target mock-ups have been constructed and tested at water and lithium test facilities in the IPPE of Russia. Jet velocity in both mock-ups was up to 20 m/s. Calculations and experiments showed lithium flow instability at conjunction point of straight and concave sections of the mock-up back wall. Water velocity profile across the mock-up width, jet thickness, and wave height were measured. The significant increase of thickness of both water and lithium jets near the mock-up sidewalls was observed. The influence of shape of the nozzle outlet part on jet stability was investigated. Lithium evaporation from the jet free surface was investigated as well as lithium deposition on vacuum pipe walls of the target mock-up. It was shown that these phenomena are not very critical for the target efficiency. The possibility of lithium denitration down to 2 ppm (at 10 ppm requested) by means of aluminium getter was shown. Two types of cold traps and plug indicators of impurities were tested. The results are presented in the paper.

  14. Hydraulic numerical analyses of the IFMIF target performance

    International Nuclear Information System (INIS)

    Gordeev, S.; Heinzel, V.; Stieglitz, R.

    2011-01-01

    The target of the International Fusion Material Irradiation Facility (IFMIF) is one of the most crucial components of the planned facility, since due the high power densities only a free surface operated liquid lithium target is potentially capable to match the constraints set by the user community. Therefore in the design process two design options for such a free surface target have been elaborated. Both are analyzed by means of a time dependent large eddy simulation in this paper. The analysis of the numerical data rapidly shows that not only the nozzle but also the flow straighteners installed upstream the two step acceleration nozzle impact the flow especially in the boundary layers of the duct flow domain. Those boundary layers formulate the inlet conditions into the technically relevant target domain. In general both target design options serve an acceptable film thickness in the irradiation area. However, the design option with a straight to curved back wall is less preferable than a design solution with an entirely curved back wall since the surface shape exhibits larger surface fluctuations and a higher fluctuation intensity although the thermal behaviour close to the back wall is marginally better than for the entirely curved option.

  15. IFMIF (International Fusion Materials Irradiation Facility) key element technology phase task description

    Energy Technology Data Exchange (ETDEWEB)

    Ida, M.; Nakamura, H.; Sugimoto, M.; Yutani, T.; Takeuchi, H. [eds.] [Japan Atomic Energy Research Inst., Tokai Research Establishment, Fusion Neutron Laboratory, Tokai, Ibaraki (Japan)

    2000-08-01

    In 2000, a 3 year Key Element technology Phase (KEP) of the International Fusion Materials Irradiation Facility (IFMIF) has been initiated to reduce the key technology risk factors needed to achieve continuous wave (CW) beam with the desired current and energy and to reach the corresponding power handling capabilities in the liquid lithium target system. In the KEP, the IFMIF team (EU, Japan, Russian Federation, US) will perform required tasks. The contents of the tasks are described in the task description sheet. As the KEP tasks, the IFMIF team have proposed 27 tasks for Test Facilities, 12 tasks for Target, 26 tasks for Accelerator and 18 tasks for Design Integration. The task description by RF is not yet available. The task items and task descriptions may be added or revised with the progress of KEP activities. These task description sheets have been compiled in this report. After 3 years KEP, the results of the KEP tasks will be reviewed. Following the KEP, 3 years Engineering Validation Phase (EVP) will continue for IFMIF construction. (author)

  16. Activation of the IFMIF prototype accelerator and beam dump by deuterons and protons

    Czech Academy of Sciences Publication Activity Database

    Simakov, S. P.; Bém, Pavel; Burjan, Václav; Fischer, U.; Forrest, R.A.; Götz, Miloslav; Honusek, Milan; Klein, H.; Kroha, Václav; Novák, Jan; Sauer, A.; Šimečková, Eva; Tiede, R.

    2008-01-01

    Roč. 83, 10-12 (2008), s. 1543-1547 ISSN 0920-3796 R&D Projects: GA MPO 2A-1TP1/101 Institutional research plan: CEZ:AV0Z10480505 Keywords : IFMIF * Protons and deuterons accelerator * Beam dump Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 0.828, year: 2008

  17. 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)

  18. Development of activation foils method for the IFMIF neutron flux chracterization

    Czech Academy of Sciences Publication Activity Database

    Simakov, S. P.; Bém, Pavel; Burjan, Václav; Fischer, U.; Götz, Miloslav; Honusek, Milan; Kroha, Václav; Novák, Jan; Šimečková, Eva

    2007-01-01

    Roč. 82, 15-24 (2007), s. 2510-2517 ISSN 0920-3796 R&D Projects: GA MPO 2A-1TP1/101 Institutional research plan: CEZ:AV0Z10480505 Keywords : IFMIF Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.058, year: 2007

  19. The engineering design evolution of IFMIF: From CDR to EDA phase

    Energy Technology Data Exchange (ETDEWEB)

    Pérez, Mario, E-mail: mario.perez@ifmif.org

    2015-10-15

    Highlights: • Brief description of International Fusion Materials Irradiation Facility (IFMIF), its background and scope its Engineering Design and Validation Activities (EVEDA) phase. • Description and justification of the main design evolutions from previous phases; and in particular from the baseline described in the “Comprehensive Design Report” (CDR). - Abstract: The International Fusion Materials Irradiation Facility (IFMIF), presently in its Engineering Design and Engineering Validation Activities (EVEDA) phase, started in 2007 under the framework of the Broader Approach (BA) Agreement between Japanese Government and EURATOM. The mandate assigned was to develop an integrated engineering design of IFMIF together with accompanying sub-projects to validate the major technological challenges that included the construction of either full scale prototypes or cleverly devised scaled down facilities, which are essential to reliably face the construction of IFMIF on schedule and cost. The Engineering Design Activities were accomplished on-schedule with the release of its “Intermediate Engineering Design Report (IIEDR)” in June 2013 compliant with our mandate. This paper highlights the design improvements implemented from the previous Conceptual Design Phase.

  20. Activation of the IFMIF Lithium Loop Corrosion Products

    Energy Technology Data Exchange (ETDEWEB)

    Cambi, G [Department of Physics, Bologna University, Via Irnerio 46, 40126 Bologna (Italy); Cepraga, D G [ENEA FIS-MET, Via Don Fiammelli 2, 40128 Bologna (Italy); Frisoni, M [Athena s.a.s., Via del Battiferro 3, 40129 Bologna (Italy); Pinna, T [Associazione EURATOM- ENEA sulla Fusione, Via Enrico Fermi 45, I-00044 Frascati (RM), (Italy)

    2006-07-01

    The assessment of the activation of steel corrosion products generated in one year of IFMIF lithium loop operation due to the interaction between lithium and Stainless Steel SS-304 has been performed. This paper is mainly focused on the neutron activation and it describes the approach used for and present the results obtained. A preliminary estimate of the accelerator deuteron beam contribute to the activation is also presented. The study was accomplished through the following phases: 1) neutron spectrum calculation in the lithium target via MCNP-4C2 with McEnea neutron source model based on the measurements of neutron emission spectra produced in Li(d,n) reactions for a thick lithium target performed at the '' Cyclotron and Radioisotope Center (CYRIC) '', Tohoku University, Japan; 2) inventories calculations and decay gamma sources production via ANITA-IEAF activation code package; the calculations were performed by considering a lithium mix composition containing lithium impurities and corrosion products referred to 200 wppm of Steel SS-304 corresponding to a corrosion rate of 0.2 {mu}m/y and a SS-304 wetted surface of 572 m{sup 2} ; an irradiation scenario reproducing the integrated (in eleven months of operation) neutron flux responsible for the activation of the circulating corrosion products facing the deuteron beam was considered; 3) decay gamma transport analysis for dose rate evaluations via both VITENEA-IEF/SCALENEA-1 and MCNP-4C2 systems for the Longest Pipe of the Lithium loop. The following conclusions can be drawn by the results analysis: {center_dot} dose rates at 50 cm from the Longest Pipe are 198 {mu}Sv/h and 85{mu}Sv/h at 1 day and 1 week from the plant shutdown, respectively {center_dot} considering the average 20 mSv/a regulatory limit in Europe for '' Radiation Worker '' and the four-week period of annual maintenance activities in Li loop, the zone around the piping, exceeding 125 mSv/h, has to be declared '' Restricted Access Area '' {center

  1. Activation of the IFMIF Lithium Loop Corrosion Products

    International Nuclear Information System (INIS)

    Cambi, G.; Cepraga, D.G.; Frisoni, M.; Pinna, T.

    2006-01-01

    The assessment of the activation of steel corrosion products generated in one year of IFMIF lithium loop operation due to the interaction between lithium and Stainless Steel SS-304 has been performed. This paper is mainly focused on the neutron activation and it describes the approach used for and present the results obtained. A preliminary estimate of the accelerator deuteron beam contribute to the activation is also presented. The study was accomplished through the following phases: 1) neutron spectrum calculation in the lithium target via MCNP-4C2 with McEnea neutron source model based on the measurements of neutron emission spectra produced in Li(d,n) reactions for a thick lithium target performed at the '' Cyclotron and Radioisotope Center (CYRIC) '', Tohoku University, Japan; 2) inventories calculations and decay gamma sources production via ANITA-IEAF activation code package; the calculations were performed by considering a lithium mix composition containing lithium impurities and corrosion products referred to 200 wppm of Steel SS-304 corresponding to a corrosion rate of 0.2 μm/y and a SS-304 wetted surface of 572 m 2 ; an irradiation scenario reproducing the integrated (in eleven months of operation) neutron flux responsible for the activation of the circulating corrosion products facing the deuteron beam was considered; 3) decay gamma transport analysis for dose rate evaluations via both VITENEA-IEF/SCALENEA-1 and MCNP-4C2 systems for the Longest Pipe of the Lithium loop. The following conclusions can be drawn by the results analysis: · dose rates at 50 cm from the Longest Pipe are 198 μSv/h and 85μSv/h at 1 day and 1 week from the plant shutdown, respectively · considering the average 20 mSv/a regulatory limit in Europe for '' Radiation Worker '' and the four-week period of annual maintenance activities in Li loop, the zone around the piping, exceeding 125 mSv/h, has to be declared '' Restricted Access Area '' · the worker radiation protection

  2. Experimental Investigation of the IFMIF Target Mock-up

    International Nuclear Information System (INIS)

    Loginov, N.; Mikheyev, A.; Morozov, V.; Aksenov, Y.; Arnoldov, M.; Berensky, L.; Fedotovsky, V.; Chernov, V.M.; Nakamura, H.

    2007-01-01

    Full text of publication follows: The IFMIF lithium neutron target mock-ups have been constructed and tested at the water and lithium test facilities. Description of the mock-ups and test facilities is presented in the paper, as well as the main results obtained. Reference geometry was used but the mockup flow cross-section was decreased. Velocity of water and lithium was up to reference value of 20 m/s. Features of lithium and water hydrodynamics were observed. The calculations and experiments showed that conjunction point of back wall straight and concave sections generated instability of lithium flow because of centrifugal force sudden change at this place. Therefore, it was proposed to use parabolic shape of the target back wall. Generation of wakes at the corners of cross-section of the Shima nozzle outlet was observed, and, as a result, surface waves appeared on the lithium jet. Observations of lithium and water jets and measurements of water jet thickness showed significant increasing the thickness near sidewalls of the mock-up concave section. It is because of absence of the centrifugal force at these places. Very large instability of the water jet surface was observed when outlet part of the Shima nozzle was divergent slightly (about 1 deg.), and vice versa very smooth jet surface occurred in confusing case (of about 0.5 deg.). So, nozzle outlet shape is very critical. Evaporation of lithium from the jet surface was investigated as well as deposition of vapor on vacuum pipe wall. It turned out to be not so critical. Significant part of the work concerned purification of lithium and monitoring impurities. The possibility of denitration of lithium down to 2 ppm by means of aluminum soluble getter was showed. Two types of both cold traps and plug indicators of impurities were tested. The results are presented in the paper. (authors)

  3. Investigation of high flux test module for the international fusion materials irradiation facilities (IFMIF)

    International Nuclear Information System (INIS)

    Miyashita, Makoto; Sugimoto, Masayoshi; Yutani, Toshiaki

    2007-03-01

    This report describes investigation on structure of a high neutron flux test module (HFTM) for the International Fusion Materials Irradiation Facilities (IFMIF). The HFTM is aimed for neutron irradiation of a specimen in a high neutron flux domain of the test cell for irradiation ground of IFMIF. We investigated the overall structure of the HFTM that was able to include specimens in a rig and thermocouple arrangement, an interface of control signal and support structure. Moreover, pressure and the amount of the bend in the module vessel (a rectangular section pressure vessel) were calculated. The module vessel did a rectangular section from limitation of a high neutron flux domain. Also, we investigated damage of thermocouples under neutron irradiation, which was a temperature sensor of irradiation materials temperature control demanded high precision. Based on these results, drawings on the HTFM structure. (author)

  4. Cea-DSM-DAPNIA-SACM contribution to IFMIF KEP phase June 2000 to December 2002

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    The international fusion materials irradiation facility (IFMIF) requires the generation by a linear accelerator (LINAC) of 250 mA continuous current of deuterons at a nominal energy of 40 MeV. The basic approach is to provide 2 linac modules, each delivering 125 mA to a common target. The accelerators begin with a deuteron ion source and a low-energy beam transport to a radio-frequency quadrupole (RFQ), a buncher and a pre-accelerator up to 5 MeV. The key element technology phase (KEP) was initiated in 2000 with the objective of reducing some key technology risk factors. The IFMIF KEP is carried out at the Cea and it focuses on 5 issues: the ECR source, the 4-vanes RFQ design, the radio-frequency system, the DTL (drift tube linac) design, and high power diagnostics. The present report reviews progress made in the 5 issues quoted above. (A.C.)

  5. Issues to be verified by IFMIF prototype accelerator for engineering validation

    International Nuclear Information System (INIS)

    Sugimoto, M.; Imai, T.; Okumura, Y.; Nakayama, K.; Suzuki, S.; Saigusa, M.

    2002-01-01

    The validation of the accelerator technology providing the 250 mA/40 MeV continuous-wave (CW) deuteron beam with the required quality is a key issue to realize the international fusion materials irradiation facility (IFMIF). As the difficulty of high current accelerator generally comes from the low energy section due to space-charge effects, a prototype test of such a part is planned in the next development phase. The optimal choice of the prototype consists of a full-scale injector, a full-modelled radiofrequency quadrupole, and a short drift tube linear accelerator associated with a beam diagnostics/dump. Through prototype tests, the stable control of the CW accelerator at the various operational conditions will be addressed, and the technical risks of IFMIF accelerator construction can be significantly reduced

  6. Development of a high brightness ion source for IFMIF and preliminary test results

    International Nuclear Information System (INIS)

    Iga, Takashi; Okumura, Yoshikazu; Kashiwagi, Mieko

    2001-05-01

    Development of a high brightness ion source for the 40MeV/250mA deuteron beam accelerator, IFMIF, is in progress at JAERI. A prototype ion source using hot filament cathodes has been developed. This ion source consists of a multi-cusp plasma generator and a two-stage accelerator. Beam optics has been investigated at the energy of up to 60keV. Experimental results of the beam optics agreed well with the simulation by assuming that the equivalent ion mass is 2.38. Ion beam of 60keV/100mA H+, which corresponds to ion beam of 100keV/220mA D+, was obtained with optimum perveance (minimum divergence). This result indicates that the current requirement for the IFMIF ion source would be satisfied with this ion source. (author)

  7. Minutes of the IFMIF technical meetings, May 17-20, 2005, Tokyo, Japan

    International Nuclear Information System (INIS)

    Ida, Mizuho; Nakamura, Hiroo; Yutani, Toshiaki; Maebara, Sunao; Umetsu, Tomotake; Sugimoto, Masayoshi

    2005-08-01

    The International Fusion Materials Irradiation Facility (IFMIF) Technical Meetings were held on May 17-20, 2005 at Japan Atomic Energy Research Institute (JAERI) Tokyo. The main objectives were 1) to review technical status of the subsystems; accelerator, target and test facilities, 2) to technically discuss interface issues between target and test facilities, 3) to review results of peer-reviews performed in the EU and Japan, 4) to harmonize design/experimental activities among the subsystems, 5) to review and discuss the Engineering Validation and Engineering Design Activity (EVEDA) tasks, and 6) to make a report of 1) - 5) to the IFMIF Executive Subcommittee. This report presents a brief summary of the Target Technical, Meeting, Test Facilities Technical Meeting, Target/Test Facilities Interface Meeting, Accelerator Technical Meeting and the Technical Integration Meeting. (author)

  8. Nuclear irradiation parameters of beryllium under fusion, fission and IFMIF irradiation conditions

    International Nuclear Information System (INIS)

    Fischer, U.; Chen, Y.; Leichtle, D.; Simakov, S.; Moeslang, A.; Vladimirov, P.

    2004-01-01

    A computational analysis is presented of the nuclear irradiation parameters for Beryllium under irradiation in typical neutron environments of fission and fusion reactors, and of the presently designed intense fusion neutron source IFMIF. The analysis shows that dpa and Tritium production rates at fusion relevant levels can be achieved with existing high flux fission reactors while the achievable Helium production is too low. The resulting He-Tritium and He/dpa ratios do not meet typical fusion irradiation conditions. Irradiation simulations in the medium flux test modules of the IFMIF neutron source facility were shown to be more suitable to match fusion typical irradiation conditions. To achieve sufficiently high production rates it is suggested to remove the creep-fatigue testing machine together with the W spectra shifter plate and move the tritium release module upstream towards the high flux test module. (author)

  9. Advances in liquid metal cooled ADS systems, and useful results for the design of IFMIF

    International Nuclear Information System (INIS)

    Massaut, V.; Debruyn, D.; Decreton, M.

    2007-01-01

    Full text of publication follows: Liquid metal cooled Accelerator Driven Systems (ADS) have a lot of design commonalities with the design of IFMIF. The use of a powerful accelerator and a liquid metal spallation source makes it similar to the main features of the IFMIF irradiator. Developments in the field of liquid metal ADS can thus be very useful for the design phase of IFMIF, and synergy between both domains should be enhanced to avoid dubbing work already done. The liquid metal ADS facilities are developed for testing materials under high fast (> 1 MeV) neutron flux, and also for studying the transmutation of actinides as foreseen in the P and T (Partitioning and Transmutation) strategy of future fission industry. The ADS are mostly constituted of a sub-critical fission fuel assembly matrix, a spallation source (situated at the centre of the fuel arrangement) and a powerful accelerator targeting the spallation source. In liquid metal ADS, the spallation source is a liquid metal (like Pb-Bi) which is actively cooled to remove the power generated by the particle beam, spallation reactions and neutrons. Based on an advanced ADS design (e.g. the MYRRHA/XT-ADS facility), the paper shows the various topics which are common for both facilities (ADS and IFMIF) and highlights their respective specificities, leading to focused R and D activities. This would certainly cover the common aspects related to high power accelerators, liquid metal targets and beam-target coupling. But problems of safety, radioprotection, source heating and cooling, neutrons shielding, etc... lead also to common features and developments. Results already obtained for the ADS development will illustrate this synergy. This paper will therefore allow to take profit of recent developments in both fission and fusion programs and enhance the collaboration among the R and D teams in both domains. (authors)

  10. Advances in liquid metal cooled ADS systems, and useful results for the design of IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Massaut, V.; Debruyn, D. [SCK CEN, Mol (Belgium); Decreton, M. [Ghent Univ., Dept. of Applied Physics (Belgium)

    2007-07-01

    Full text of publication follows: Liquid metal cooled Accelerator Driven Systems (ADS) have a lot of design commonalities with the design of IFMIF. The use of a powerful accelerator and a liquid metal spallation source makes it similar to the main features of the IFMIF irradiator. Developments in the field of liquid metal ADS can thus be very useful for the design phase of IFMIF, and synergy between both domains should be enhanced to avoid dubbing work already done. The liquid metal ADS facilities are developed for testing materials under high fast (> 1 MeV) neutron flux, and also for studying the transmutation of actinides as foreseen in the P and T (Partitioning and Transmutation) strategy of future fission industry. The ADS are mostly constituted of a sub-critical fission fuel assembly matrix, a spallation source (situated at the centre of the fuel arrangement) and a powerful accelerator targeting the spallation source. In liquid metal ADS, the spallation source is a liquid metal (like Pb-Bi) which is actively cooled to remove the power generated by the particle beam, spallation reactions and neutrons. Based on an advanced ADS design (e.g. the MYRRHA/XT-ADS facility), the paper shows the various topics which are common for both facilities (ADS and IFMIF) and highlights their respective specificities, leading to focused R and D activities. This would certainly cover the common aspects related to high power accelerators, liquid metal targets and beam-target coupling. But problems of safety, radioprotection, source heating and cooling, neutrons shielding, etc... lead also to common features and developments. Results already obtained for the ADS development will illustrate this synergy. This paper will therefore allow to take profit of recent developments in both fission and fusion programs and enhance the collaboration among the R and D teams in both domains. (authors)

  11. IFMIF, International Fusion Materials Irradiation Facility conceptual design activity cost report

    International Nuclear Information System (INIS)

    Rennich, M.J.

    1996-12-01

    This report documents the cost estimate for the International Fusion Materials Irradiation Facility (IFMIF) at the completion of the Conceptual Design Activity (CDA). The estimate corresponds to the design documented in the Final IFMIF CDA Report. In order to effectively involve all the collaborating parties in the development of the estimate, a preparatory meeting was held at Oak Ridge National Laboratory in March 1996 to jointly establish guidelines to insure that the estimate was uniformly prepared while still permitting each country to use customary costing techniques. These guidelines are described in Section 4. A preliminary cost estimate was issued in July 1996 based on the results of the Second Design Integration Meeting, May 20--27, 1996 at JAERI, Tokai, Japan. This document served as the basis for the final costing and review efforts culminating in a final review during the Third IFMIF Design Integration Meeting, October 14--25, 1996, ENEA, Frascati, Italy. The present estimate is a baseline cost estimate which does not apply to a specific site. A revised cost estimate will be prepared following the assignment of both the site and all the facility responsibilities

  12. Preliminary definition of the remote handling system for the current IFMIF Test Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Queral, V., E-mail: vicentemanuel.queral@ciemat.es [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Urbon, J. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain); Garcia, A.; Cuarental, I.; Mota, F. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Micciche, G. [CR ENEA Brasimone, I-40035 Camugnano (BO) (Italy); Ibarra, A. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Rapisarda, D. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain); Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain); Casal, N. [Laboratorio Nacional de Fusion, EURATOM-CIEMAT, 28040 Madrid (Spain)

    2011-10-15

    A coherent design of the remote handling system with the design of the components to be manipulated is vital for reliable, safe and fast maintenance, having a decisive impact on availability, occupational exposures and operational cost of the facility. Highly activated components in the IFMIF facility are found at the Test Cell, a shielded pit where the samples are accurately located. The remote handling system for the Test Cell reference design was outlined in some past IFMIF studies. Currently a new preliminary design of the Test Cell in the IFMIF facility is being developed, introducing important modifications with respect to the reference one. This recent design separates the previous Vertical Test Assemblies in three functional components: Test Modules, shielding plugs and conduits. Therefore, it is necessary to adapt the previous design of the remote handling system to the new maintenance procedures and requirements. This paper summarises such modifications of the remote handling system, in particular the assessment of the feasibility of a modified commercial multirope crane for the handling of the weighty shielding plugs for the new Test Cell and a quasi-commercial grapple for the handling of the new Test Modules.

  13. Review of JAEA activities on the IFMIF liquid lithium target in FY2006

    International Nuclear Information System (INIS)

    Ida, Mizuho; Nakamura, Hiroo; Miyashita, Makoto; Sugimoto, Masayoshi; Chida, Teruo; Furuya, Kazuyuki; Yoshida, Eiichi; Hirakawa, Yasuhi; Miyake, Osamu; Hirabayashi, Masaru; Ara, Kuniaki

    2008-03-01

    Engineering Validation Design and Engineering Design Activity (EVEDA) of the International Fusion Materials Irradiation Facility (IFMIF) is under going. IFMIF is an accelerator-based Deuterium-Lithium (D-Li) neutron source to produce intense high energy neutrons and a sufficient irradiation volume for testing candidate materials for fusion reactors. To realize such a condition, 40 MeV deuteron beam with a current of 250 mA is injected into high speed liquid Li flow with a speed of 20 m/s. In target system, nuclear heating due to neutron causes thermal stress especially on a back-wall of the target assembly. In addition, radioactive species such as beryllium-7, tritium and activated corrosion products are generated. In this report, thermal stress analyses of the back-wall, mechanical tests on weld specimen made of the back-wall material, estimations of beryllium-7 behavior and worker dose at the IFMIF Li loop and consideration on major EVEDA tasks are summarized. (author)

  14. Preliminary definition of the remote handling system for the current IFMIF Test Facilities

    International Nuclear Information System (INIS)

    Queral, V.; Urbon, J.; Garcia, A.; Cuarental, I.; Mota, F.; Micciche, G.; Ibarra, A.; Rapisarda, D.; Casal, N.

    2011-01-01

    A coherent design of the remote handling system with the design of the components to be manipulated is vital for reliable, safe and fast maintenance, having a decisive impact on availability, occupational exposures and operational cost of the facility. Highly activated components in the IFMIF facility are found at the Test Cell, a shielded pit where the samples are accurately located. The remote handling system for the Test Cell reference design was outlined in some past IFMIF studies. Currently a new preliminary design of the Test Cell in the IFMIF facility is being developed, introducing important modifications with respect to the reference one. This recent design separates the previous Vertical Test Assemblies in three functional components: Test Modules, shielding plugs and conduits. Therefore, it is necessary to adapt the previous design of the remote handling system to the new maintenance procedures and requirements. This paper summarises such modifications of the remote handling system, in particular the assessment of the feasibility of a modified commercial multirope crane for the handling of the weighty shielding plugs for the new Test Cell and a quasi-commercial grapple for the handling of the new Test Modules.

  15. Comparative study of the tungsten irradiation conditions in IFMIF and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Simakov, S.P.; Pereslavtsev, P.; Fischer, U. [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology; Moeslang, A. [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen (Germany). Inst. for Material Research I

    2010-05-15

    The International Fusion Material Irradiation Facility (IFMIF) [1] will provide an accelerator based intense neutron source with a white spectrum extending up to 55 MeV for high fluence irradiations of fusion power reactor (FPR) candidate materials. Material samples located in test modules will be subjected to a radiation load anticipated for a fusion power reactor. The highest neutron flux is expected in the High Flux Test Module, which is considered in the IFMIF design to host around 1000 compactly packed stainless steel samples - the main structure materials of power fusion reactors. Another material subjected to the highest loads in a FPR is a tungsten. It is planned to be used as armour tiles for the divertor or the first wall. It turned out that no specific effort has been undertaken so far to search for a suitable irradiation location in the IFMIF Test Cell which provides a reasonable representation of the irradiation conditions in the divertor of a fusion power reactors. (orig.)

  16. IFMIF target and test cell - design and integration

    International Nuclear Information System (INIS)

    Heinzel, V.

    2007-01-01

    The International Fusion Material Irradiation Facility (IFMIF) aims at the qualification of appropriate materials for a Demonstration Fusion Power Plant (DEMO) to a fluence of up to 150 dpa (displacement per atom) at a DEMO typical neutron spectrum. It comprises two accelerators each providing a deuteron beam with 125 mA and 40 MeV. The deuterons strike a lithium target and create via stripping reactions neutrons. The neutrons are mainly forward directed into the High-Flux-Test-Module (HFTM). The Medium Flux-Test-Modules (MFTM) and the Low-Flux-Test-Modules (LFTM) are arranged in beam direction behind. In the HFTM a damage rate in steel of more than 20 dpa/fpy (displacement per atome per full power year) will be provide in a volume of 0.5 litre. The neutron spectrum is prone to produce helium and tritium in steel like in the first wall of a DEMO reactor. The Medium- Flux-Test-Modules are designed for creep fatigues in situ and tritium release test. The test modules are cooled with helium. The target is a lithium jet with a free surface towards the deuteron beams. The jet follows a concave curved so called back wall. Centrifugal forces increase the static pressure, which prevents lithium boiling at the beam tube pressure and the power release of 10 MW due to the deuteron beams. The target and Test Cell (TTC) houses the target and the test modules as well as the lithium supply tubes and a quench tank into which the lithium splashes after the target. The lithium containing components have a temperature of 250 to 350 C. Nuclear reactions mainly in beam direction contribute to heat releases in TTC components. The TTC is filled with a noble gas with almost atmospheric pressure. Natural convection transfers heat to the walls but also mitigates temperature peaks. The Forschungszentrum Karlsruhe (FZK) has developed or validated tools for: - The extended Monte Carlo Code McDeLicious for calculations of the neutron source term, dpa rates in the material specimens, activation

  17. Application of the IEAF-2001 activation data library to activation analyses of the IFMIF high flux test module

    International Nuclear Information System (INIS)

    Fischer, U.; Wilson, P.P.H.; Leichtle, D.; Simakov, S.P.; Moellendorff, U. von; Konobeev, A.; Korovin, Yu.; Pereslavtsev, P.; Schmuck, I.

    2002-01-01

    A complete activation data library IEAF-2001 (intermediate energy activation file) has been developed in standard ENDF-6 format with neutron-induced activation cross sections for 679 target nuclides from Z=1 (hydrogen) to Z=84 (polonium) and incident neutron energies up to 150 MeV. Using the NJOY processing code, an IEAF-2001 working library has been prepared in a 256 energy group structure for enabling activation analyses of the International Fusion Material Irradiation Facility (IFMIF) D-Li neutron source. This library was applied to the activation analysis of the IFMIF high flux test module using the recent Analytical and Laplacian Adaptive Radioactivity Analysis activation code which is capable of handling the variety of reaction channels open in the energy domain above 20 MeV. The IEAF-2001 activation library was thus shown to be suitable for activation analyses in fusion technology and intermediate energy applications such as the IFMIF D-Li neutron source

  18. IFMIF high flux test module - recent progress in design and manufacturing

    International Nuclear Information System (INIS)

    Leichtle, D.; Arbeiter, F.; Dolensky, B.

    2007-01-01

    The International Fusion Material Irradiation Facility (IFMIF) is an accelerator driven neutron source for irradiation tests of candidate fusion reactor materials. Two 40 MeV deuterium beams with 125 mA each strike a liquid lithium jet target, producing a high intensity neutron flux up to 55 MeV, which penetrates the adjacent test modules. Within the High Flux Test Module (HFTM) a testing volume of 0.5 litres filled by qualified small scale specimens will be irradiated at displacement rates of 20-50 dpa/fpy in structural materials. The HFTM will also provide helium and hydrogen production to dpa ratios that reflect within the uncertainties the values expected in a DEMO fusion reactor The Forschungszentrum Karlsruhe (FZK) has developed a HFTM design which closely follows the design premise of maximising the space available for irradiation specimens in the IFMIF high flux zone and in addition allows keeping the temperature nearly constant in the rigs containing the specimen. Within the entire specimen stack the temperature deviation will be below about 15 K. The main design principles applied are (i) filling the gaps between the specimens with liquid metal, (ii) winding three separately controlled heater sections on the inner capsules and (iii) dividing the test rigs in a hot inner and a cold outer zone, which a separated by a gap filled with stagnant helium that serves as a thermal insulator. Channels between the outer covers (the cold parts) are cooled by helium gas at moderate pressure (3 bars at inlet) and temperature (50 C). 12 identical rigs holding the specimen capsules which are heated by segmented helically wound electrical heaters ensure a flexible loading scheme during IFMIF operation. Complementary analyses on nuclear, thermo-hydraulics and mechanical performance of the HFTM were performed to optimize the design. The present paper highlights the main design characteristics as well as recent progress achieved in this area. This includes the stiffening of

  19. Minutes of the second IFMIF-CDA design integration workshop. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Konishi, Satoshi

    1996-08-01

    The second Design Integration Workshop of IFMIF-CDA was held on May 20-27, 1996 at JAERI/Tokai. The primary objectives were, (1) to review and update the Baseline Design Concept, (2) to review the preliminary schedule and cost estimates, and (3) to establish the R and D needs for the next phase of the activity. This report presents a brief summary of the objective and results of the meeting. Detailed information on the agenda, attendees, and presentation material is included in the Appendix. (author)

  20. Development and validation status of the IFMIF High Flux Test Module

    International Nuclear Information System (INIS)

    Arbeiter, Frederik; Abou-Sena, Ali; Chen Yuming; Dolensky, Bernhard; Heupel, Tobias; Klein, Christine; Scheel, Nicola; Schlindwein, Georg

    2011-01-01

    The development of the IFMIF (International Fusion Material Irradiation Facility) High Flux Test Module in the EVEDA (Engineering Validation and Engineering Design Activities) phase up to 2013 includes conceptual design, engineering analyses, as well as design and engineering validation by building of prototypes and their testing. The High Flux Test Module is the device to facilitate the irradiation of SSTT samples of RAFM steels at temperatures 250-550 deg. C and up to an accumulated irradiation damage of 150 dpa. The requirements, the current design and the performance of the module are discussed, and the development process is outlined.

  1. Development and validation status of the IFMIF High Flux Test Module

    Energy Technology Data Exchange (ETDEWEB)

    Arbeiter, Frederik, E-mail: frederik.arbeiter@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology (KIT-INR), Karlsruhe (Germany); Abou-Sena, Ali; Chen Yuming; Dolensky, Bernhard; Heupel, Tobias; Klein, Christine; Scheel, Nicola; Schlindwein, Georg [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology (KIT-INR), Karlsruhe (Germany)

    2011-10-15

    The development of the IFMIF (International Fusion Material Irradiation Facility) High Flux Test Module in the EVEDA (Engineering Validation and Engineering Design Activities) phase up to 2013 includes conceptual design, engineering analyses, as well as design and engineering validation by building of prototypes and their testing. The High Flux Test Module is the device to facilitate the irradiation of SSTT samples of RAFM steels at temperatures 250-550 deg. C and up to an accumulated irradiation damage of 150 dpa. The requirements, the current design and the performance of the module are discussed, and the development process is outlined.

  2. Overview of results of the first phase of validation activities for the IFMIF High Flux Test Module

    Energy Technology Data Exchange (ETDEWEB)

    Arbeiter, Frederik, E-mail: frederik.arbeiter@kit.edu [Karlsruhe Institute of Technology, Karlsruhe (Germany); Chen Yuming; Dolensky, Bernhard; Freund, Jana; Heupel, Tobias; Klein, Christine; Scheel, Nicola; Schlindwein, Georg [Karlsruhe Institute of Technology, Karlsruhe (Germany)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Validation of computational fluid dynamics (CFD) modeling approach for application in the IFMIF High Flux Test Module. Black-Right-Pointing-Pointer Fabrication of prototypes of the irradiation capsules of the IFMIF High Flux Test Module. - Abstract: The international fusion materials irradiation facility (IFMIF) is projected to create an experimentally validated database of material properties relevant for fusion reactor designs. The IFMIF High Flux Test Module is the dedicated experiment to irradiate alloys in the temperature range 250-550 Degree-Sign C and up to 50 displacements per atom per irradiation cycle. The High Flux Test Module is developed to maximize the specimen payload in the restricted irradiation volume, and to minimize the temperature spread within each specimen bundle. Low pressure helium mini-channel cooling is used to offer a high integration density. Due to the demanding thermo-hydraulic and mechanical conditions, the engineering design process (involving numerical neutronic, thermo-hydraulic and mechanical analyses) is supported by extensive experimental validation activities. This paper reports on the prototype manufacturing, thermo-hydraulic modeling experiments and component tests, as well as on mechanical testing. For the testing of the 1:1 prototype of the High Flux Test Module, a dedicated test facility, the Helium Loop Karlsruhe-Low Pressure (HELOKA-LP) has been taken into service.

  3. Overview of results of the first phase of validation activities for the IFMIF High Flux Test Module

    International Nuclear Information System (INIS)

    Arbeiter, Frederik; Chen Yuming; Dolensky, Bernhard; Freund, Jana; Heupel, Tobias; Klein, Christine; Scheel, Nicola; Schlindwein, Georg

    2012-01-01

    Highlights: ► Validation of computational fluid dynamics (CFD) modeling approach for application in the IFMIF High Flux Test Module. ► Fabrication of prototypes of the irradiation capsules of the IFMIF High Flux Test Module. - Abstract: The international fusion materials irradiation facility (IFMIF) is projected to create an experimentally validated database of material properties relevant for fusion reactor designs. The IFMIF High Flux Test Module is the dedicated experiment to irradiate alloys in the temperature range 250–550 °C and up to 50 displacements per atom per irradiation cycle. The High Flux Test Module is developed to maximize the specimen payload in the restricted irradiation volume, and to minimize the temperature spread within each specimen bundle. Low pressure helium mini-channel cooling is used to offer a high integration density. Due to the demanding thermo-hydraulic and mechanical conditions, the engineering design process (involving numerical neutronic, thermo-hydraulic and mechanical analyses) is supported by extensive experimental validation activities. This paper reports on the prototype manufacturing, thermo-hydraulic modeling experiments and component tests, as well as on mechanical testing. For the testing of the 1:1 prototype of the High Flux Test Module, a dedicated test facility, the Helium Loop Karlsruhe-Low Pressure (HELOKA-LP) has been taken into service.

  4. Neutronics analysis of International Fusion Material Irradiation Facility (IFMIF). Japanese contributions

    International Nuclear Information System (INIS)

    Oyama, Yukio; Noda, Kenji; Kosako, Kazuaki.

    1997-10-01

    In fusion reactor development for demonstration reactor, i.e., DEMO, materials tolerable for D-T neutron irradiation are absolutely required for both mechanical and safety point of views. For this requirement, several kinds of low activation materials were proposed. However, experimental data by actual D-T fusion neutron irradiation have not existed so far because of lack of fusion neutron irradiation facility, except fundamental radiation damage studies at very low neutron fluence. Therefore such a facility has been strongly requested. According to agreement of need for such a facility among the international parties, a conceptual design activity (CDA) of International Fusion Material Irradiation Facility (IFMIF) has been carried out under the frame work of the IEA-Implementing Agreement. In the activity, a neutronics analysis on irradiation field optimization in the IFMIF test cell was performed in three parties, Japan, US and EU. As the Japanese contribution, the present paper describes a neutron source term as well as incident deuteron beam angle optimization of two beam geometry, beam shape (foot print) optimization, and dpa, gas production and heating estimation inside various material loading Module, including a sensitivity analysis of source term uncertainty to the estimated irradiation parameters. (author)

  5. Preliminary study on the possible use of superconducting half-wave resonators in the IFMIF Linac

    International Nuclear Information System (INIS)

    Mosnier, A.; Uriot, D.

    2007-01-01

    The driver of the International Fusion Materials Irradiation Facility (IFMIF) consists of two 125 mA, 40 MeV cw deuteron linacs, providing a total of 10 MW beam power to the liquid lithium target. A superconducting (SC) solution for the 5 to 40 MeV accelerator portion could offer some advantages compared with the copper Alvarez-type Drift Tube Linac reference design: linac length reduction and significant plug power saving. A SC scheme, based on multi-gap CH-structures has been proposed by IAP in Frankfurt. Another SC scheme, using half-wave resonators (HWR), which are in an advanced stage of development at different places, would allow a shorter focusing lattice, resulting in a safe beam transportation with minimal beam loss. In order to investigate the feasibility of the superconducting HWR option, faced with the very high space charge regime of the IFMIF linac, beam dynamics calculations have been performed. This paper presents an optimized linac layout, together with extensive multi-particle simulations including various field and alignment errors. (authors)

  6. IFMIF High Flux Test Module-Recent progress in design and manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Leichtle, D. [Association FZK-EURATOM, Forschungszentrum Karlsruhe, Institut fuer Reaktorsicherheit, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)], E-mail: leichtle@irs.fzk.de; Arbeiter, F.; Dolensky, B.; Fischer, U.; Gordeev, S.; Heinzel, V.; Ihli, T.; Lang, K.-H. [Association FZK-EURATOM, Forschungszentrum Karlsruhe, Institut fuer Reaktorsicherheit, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Moeslang, A. [Association FZK-EURATOM, Forschungszentrum Karlsruhe, Institut fuer Materialforschung, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Simakov, S.P.; Slobodchuk, V.; Stratmanns, E. [Association FZK-EURATOM, Forschungszentrum Karlsruhe, Institut fuer Reaktorsicherheit, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2008-12-15

    The International Fusion Material Irradiation Facility (IFMIF) is an accelerator driven neutron source for irradiation tests of candidate fusion reactor materials. Within the High Flux Test Module (HFTM) a testing volume of 0.5 l filled by qualified small scale specimens will be irradiated at displacement rates of 20-50 dpa/fpy in structural materials. The Forschungszentrum Karlsruhe (FZK) has developed a HFTM design which closely follows the design premise of maximising the space available for irradiation specimens in the IFMIF high flux zone and in addition allows keeping the temperature nearly constant in the rigs containing the specimen. Complementary analyses on nuclear, thermo-hydraulics and mechanical performance of the HFTM were performed to optimize the design. The present paper highlights the main design characteristics as well as recent progress achieved in this area. The contribution also includes (i) recommendations for the use of container, rig and capsule materials, and (ii) a description of the fabrication routes for the entire HFTM including brazing and filling procedures which are currently under development at the Forschungszentrum Karlsruhe.

  7. Design of a beam dump for the IFMIF-EVEDA accelerator

    International Nuclear Information System (INIS)

    Branas, B.; Iglesias, D.; Arranz, F.; Barrera, G.; Casal, N.; Garcia, M.; Gomez, J.; Lopez, D.; Martinez, J.I.; Martin-Fuertes, F.; Ogando, F.; Oliver, C.; Sanz, J.; Sauvan, P.; Ibarra, A.

    2009-01-01

    The IFMIF-EVEDA accelerator will be a 9 MeV, 125 mA cw deuteron accelerator prototype for verifying the validity of the accelerator design for IFMIF. A beam stop will be used for the RFQ and DTL commissioning as well as for the EVEDA accelerator tests. Therefore, this component must be designed to stop 5 MeV and 9 MeV deuteron beams with a maximum power of 1.13 MW. The first step of the design is the beam-facing material selection. The criteria used for this selection are low neutron production, low activation and good thermomechanical behavior. In this paper, the mechanical analysis and radioprotection calculations that have led to the choice of the main beam dump parameters will be described. The present design is based on a conical beam stop (2.5 m length, 30 cm diameter, and 3.5 mm thickness) made of copper plus a cylindrical 0.5 m long beam scraper. The cooling system is based on an axial high velocity flow of water. This design is compliant with the mechanical design rules during full power stationary operation of the accelerator. The radioprotection calculations performed demonstrate that, with an adequate local shielding, doses during beam on/off phases are below the limits.

  8. IFMIF (International Fusion Materials Irradiation Facility) conceptual design activity reduced cost report

    International Nuclear Information System (INIS)

    2000-02-01

    This report describes the results of a preliminary reevaluation of the design and cost of the International Fusion Materials Irradiation Facility (IFMIF) Project in response to the request from the 28th FPCC meeting in January 1999. Two major ideas have been considered: 1) reduction of the total construction cost through elimination of the previously planned facility upgrade and 2) a facility deployment in 3 stages with capabilities for limited experiments in the first stage. As a result, the size and complexity of the facility could be significantly reduced, leading to substantial cost savings. In addition to these two ideas, this study also included a critical review of the original CDA specification with the objective of elimination of nonessential items. For example, the number of lithium targets was reduced from two to one. As a result of these changes in addition to the elimination of the upgrade, the total cost estimate was very substantially reduced from 797.2 MICF to 487.8 MICF, where 1 MICF = 1 Million of the IFMIF Conversion Units (approximately $1M US January, 1996). (author)

  9. Analysis of displacement damage in materials in nuclear fusion facilities (DEMO, IFMIF and TechnoFusion)

    International Nuclear Information System (INIS)

    Mota, F.; Vila, R.; Ortiz, C.; Garcia, A.; Casal, N.; Ibarra, A.; Rapisarda, D.; Queral, V.

    2011-01-01

    Present pathway to fusion reactors includes a rigorous material testing program. To reach this objective, irradiation facilities must produce the displacement damage per atom (dpa), primary knock-on atom (PKA) spectrum and gaseous elements by transmutation reactions (He, H) as closely as possible to the ones expected in the future fusion reactors (as DEMO).The irradiation parameters (PKA spectra and damage function) of some candidate materials for fusion reactors (Al 2 O 3 , SiC and Fe) have been studied and then, the suitability of some proposed experimental facilities, such as IFMIF and TechnoFusion, to perform relevant tests with these materials has been assessed.The following method has been applied: neutron fluxes present in different irradiation modules of IFMIF have been calculated by the neutron transport McDeLicious code. In parallel, the energy differential cross sections of PKA have been calculated by using the NJOY code. After that, the damage generated by the PKA spectra was analyzed using the MARLOWE code (binary collision approximation) and custom analysis codes. Finally, to analyze the ions effects in different irradiation conditions in the TechnoFusion irradiation area, the SRIM and Marlowe codes have been used. The results have been compared with the expected ones for a DEMO HCLL reactor.

  10. Feasibly study of gas-cooled test cell for material testing in IFMIF

    International Nuclear Information System (INIS)

    Yonemoto, Yukihiro; Maki, Eiji; Ebara, Shinji; Yokomine, Takehiko; Shimizu, Akihiko; Korenaga, Tadashi

    2002-01-01

    Temperature control performance of test pieces enclosed in IFMIF capsule by using single phase gas was estimated experimentally. The key issue of this study is to obtain the definite value of dimension of test facility and flow conditions of coolant and to clarify the temperature response of test piece to the beam-off scenario. Firstly, we have examined the cooling performance of the test cell originally proposed in IFMIF-KEP and from results of this calculation performed in three dimensional system by using brand-new turbulence model for flow and thermal fields, it is concluded that the drastical change of design of test cell is needed in order to obtain the unformity of temperature of test piece, to improve the responsibility of temperature measurement of test piece, and to relieve the coolant flow condition, especially for inlet pressure value. Thus, we have proposed new design of test cell and test piece arrangement. A mock-up experimental facility was made based on our design and preliminary experiments for temperature control were performed. As a result, we have verified the cooling performance at the case that corresponds to two beam-off scenario by using mock-up facility

  11. IFMIF High Flux Test Module-Recent progress in design and manufacturing

    International Nuclear Information System (INIS)

    Leichtle, D.; Arbeiter, F.; Dolensky, B.; Fischer, U.; Gordeev, S.; Heinzel, V.; Ihli, T.; Lang, K.-H.; Moeslang, A.; Simakov, S.P.; Slobodchuk, V.; Stratmanns, E.

    2008-01-01

    The International Fusion Material Irradiation Facility (IFMIF) is an accelerator driven neutron source for irradiation tests of candidate fusion reactor materials. Within the High Flux Test Module (HFTM) a testing volume of 0.5 l filled by qualified small scale specimens will be irradiated at displacement rates of 20-50 dpa/fpy in structural materials. The Forschungszentrum Karlsruhe (FZK) has developed a HFTM design which closely follows the design premise of maximising the space available for irradiation specimens in the IFMIF high flux zone and in addition allows keeping the temperature nearly constant in the rigs containing the specimen. Complementary analyses on nuclear, thermo-hydraulics and mechanical performance of the HFTM were performed to optimize the design. The present paper highlights the main design characteristics as well as recent progress achieved in this area. The contribution also includes (i) recommendations for the use of container, rig and capsule materials, and (ii) a description of the fabrication routes for the entire HFTM including brazing and filling procedures which are currently under development at the Forschungszentrum Karlsruhe

  12. Analysis of displacement damage in materials in nuclear fusion facilities (DEMO, IFMIF and TechnoFusion)

    Energy Technology Data Exchange (ETDEWEB)

    Mota, F., E-mail: fernando.mota@ciemat.es [Laboratorio Nacional de Fusion por Confinamiento Magnetico-CIEMAT, 28040 Madrid (Spain); Vila, R.; Ortiz, C.; Garcia, A.; Casal, N.; Ibarra, A.; Rapisarda, D.; Queral, V. [Laboratorio Nacional de Fusion por Confinamiento Magnetico-CIEMAT, 28040 Madrid (Spain)

    2011-10-15

    Present pathway to fusion reactors includes a rigorous material testing program. To reach this objective, irradiation facilities must produce the displacement damage per atom (dpa), primary knock-on atom (PKA) spectrum and gaseous elements by transmutation reactions (He, H) as closely as possible to the ones expected in the future fusion reactors (as DEMO).The irradiation parameters (PKA spectra and damage function) of some candidate materials for fusion reactors (Al{sub 2}O{sub 3}, SiC and Fe) have been studied and then, the suitability of some proposed experimental facilities, such as IFMIF and TechnoFusion, to perform relevant tests with these materials has been assessed.The following method has been applied: neutron fluxes present in different irradiation modules of IFMIF have been calculated by the neutron transport McDeLicious code. In parallel, the energy differential cross sections of PKA have been calculated by using the NJOY code. After that, the damage generated by the PKA spectra was analyzed using the MARLOWE code (binary collision approximation) and custom analysis codes. Finally, to analyze the ions effects in different irradiation conditions in the TechnoFusion irradiation area, the SRIM and Marlowe codes have been used. The results have been compared with the expected ones for a DEMO HCLL reactor.

  13. Study on surface wave characteristics of free surface flow of liquid metal lithium for IFMIF

    International Nuclear Information System (INIS)

    Hoashi, Eiji; Sugiura, Hirokazu; Yoshihashi-Suzuki, Sachiko; Yamaoka, Nobuo; Horiike, Hiroshi; Kanemura, Takuji; Kondo, Hiroo

    2011-01-01

    The international fusion materials irradiation facility (IFMIF) presents an intense neutron source to develop fusion reactor materials. The free surface flow of a liquid metal Lithium (Li) is planned as a target irradiated by two deuteron beams to generate intense neutrons and it is thus important to obtain knowledge of the surface wave characteristic for the safety and the efficiency of system in the IFMIF. We have been studying on surface wave characteristics experimentally using the liquid metal Li circulation facility at Osaka University and numerically using computational fluid dynamics (CFD) code, FLUENT. This paper reports the results of the surface fluctuation, the wave height and the surface velocity in the free surface flow of the liquid metal Li examined experimentally and numerically. In the experiment, an electro-contact probe apparatus was used to obtain the surface fluctuation and the wave height, and a high speed video was used to measure the surface velocity. We resulted in knowledge of the surface wave growth mechanism. On the other hand, a CFD simulation was also conducted to obtain information on the relation of the free surface with the inner flow. In the simulation, the model included from a two-staged contraction nozzle to a flow channel with a free surface flow region and simulation results were compared with the experimental data. (author)

  14. Recent technical progress on BA Program: DEMO activities and IFMIF/EVEDA

    Energy Technology Data Exchange (ETDEWEB)

    Yamanishi, T.; Asakura, N.; Tobita, K.; Ohira, S. [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan); Federici, G. [EFDA Close Support Unit, Garching (Germany); Heidinger, R. [Fusion for Energy, Garching (Germany); Knaster, J. [BA IFMIF/EVEDA Project Team, Rokkasho, Aomori (Japan); Clement, S. [Fusion for Energy, Barcelona (Spain); Nakajima, N. [BA IFERC Project Team, Rokkasho, Aomori (Japan)

    2016-11-01

    The Broader Approach (BA) activities consists of three major projects: the International Fusion Energy Research Center (IFERC) project, the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project, and the Satellite Tokamak Programme (STP, JT-60SA). These projects have been carried out to obtain basic data for the design of DEMO fusion reactor from 2007. For 8-year activities, the above projects could produce a set of fruitful results for the DEMO reactor. DEMO design activity has been conducted to build a set of DEMO design bases in accordance with a series of discussion between EU and JA. In the DEMO R&D activities, five basic R&D subjects for a DEMO blanket system have been selected, and been studies under close collaborations between EU and JA: structure materials (RAFM steels and SiC/SiC composites), functional materials (tritium breeders and neutron multipliers), and tritium technology. Some additional R&D subjects recommended by peer review comments have also been studied successfully in recent years. Regarding the IFMIF/EVEDA project, some main components of the accelerator facility been designed and tested. The validation test using EVEDA Lithium Test Loop (ELTL) was also completed successfully in October 2014.

  15. Measurement of free-surface of liquid metal lithium jet for IFMIF target

    International Nuclear Information System (INIS)

    Hiroo Kondo; Nobuo Yamaoka; Takuji Kanemura; Seiji Miyamoto; Hiroshi Horiike; Mizuho Ida; Hiroo Nakamura; Izuru Matsushita; Takeo Muroga

    2006-01-01

    This reports an experimental study on flow characteristics of a lithium target flow of International Fusion Materials Irradiation Facility (IFMIF). Surface shapes of the target were tried to measure by pattern projection method that is a three dimensional image measurement method. Irregularity of the surface shape caused by surface wakes was successfully measured by the method. IFMIF liquid lithium target is formed a flat plane jet of 25 mm in depth and 260 mm in width, and flows in a flow velocity range of 10 to 20 m/s. Aim of this study is to develop measurement techniques for monitoring of the target when IFMIF is in operation. The lithium target flow is high speed jet and the temperature high is more than 500 K. Also, light is not transmitted into liquid metal lithium. Therefore, almost of all flow measurement techniques developed for water are not used for lithium flow. In this study, pattern projection method was employed to measure the surface irregularity of the target. In the method, stripe patterns are projected onto the flow surface. The projected patterns are deformed according the surface shape. Three-dimensional surface shape is measured by analyzing the deformed patterns recorded using a CCD camera. The method uses the property that lithium dose not transmit visible lights. The experiments were carried out using a lithium loop at Osaka University. In this facility, lithium plane jet of 10 mm in depth and 70 mm width is obtained in the velocity range of less than 15 m/s using a two contractions nozzle. The pattern projection method was used to measure the amplitude of surface irregularity caused by surface wakes. The surface wakes were generated from small damaged at the nozzle edge caused by erosion, and those were successfully measured by the method. The measurement results showed the amplitude of the surface wakes were approximately equal to a size of damage of a nozzle. The amplitude was decreasing with distance to down stream and with decreasing

  16. Deuteron cross section evaluation for safety and radioprotection calculations of IFMIF/EVEDA accelerator prototype

    International Nuclear Information System (INIS)

    Blideanu, Valentin; Garcia, Mauricio; Joyer, Philippe; Lopez, Daniel; Mayoral, Alicia; Ogando, Francisco; Ortiz, Felix; Sanz, Javier; Sauvan, Patrick

    2011-01-01

    In the frame of IFMIF/EVEDA activities, a prototype accelerator delivering a high power deuteron beam is under construction in Japan. Interaction of these deuterons with matter will generate high levels of neutrons and induced activation, whose predicted yields depend strongly on the models used to calculate the different cross sections. A benchmark test was performed to validate these data for deuteron energies up to 20 MeV and to define a reasonable methodology for calculating the cross sections needed for EVEDA. Calculations were performed using the nuclear models included in MCNPX and PHITS, and the dedicated nuclear model code TALYS. Although the results obtained using TALYS (global parameters) or Monte Carlo codes disagree with experimental values, a solution is proposed to compute cross sections that are a good fit to experimental data. A consistent computational procedure is also suggested to improve both transport simulations/prompt dose and activation/residual dose calculations required for EVEDA.

  17. Experimental study of lithium free-surface flow for IFMIF target design

    International Nuclear Information System (INIS)

    Kondo, H.; Fujisato, A.; Yamaoka, N.; Inoue, S.; Miyamoto, S.; Iida, T.; Nakamura, H.; Ida, M.; Matushita, I.; Muroga, T.; Horiike, H.

    2006-01-01

    Lithium free-surface flow experiments to verify the design of IFMIF target have been carried out at Osaka University. The present report summarizes experimental results of surface phenomena, and cavitation characteristics of the loop, so as to try to apply these results to design parameters. Waves on the lithium flow surface is similar to that on water, and can be predicted by a linear stability theory. The wave amplitude is measured by an electro-contact probe. Surface roughness on a target nozzle, caused for example by attached chemical compounds and/or wastages by erosion and corrosion, can lead to a significant loss of target flow stability as well as surface wakes. The need of a polishing manipulator or exchange of the nozzle may be anticipated. Cavitation characteristic of the loop was measured by an accelerometer. From the results, a friction factor could be estimated fort he lithium flow

  18. Nuclear data needs for neutron spectrum tailoring at International Fusion Materials Irradiation Facility (IFMIF)

    International Nuclear Information System (INIS)

    Sugimoto, Masayoshi

    2001-01-01

    International Fusion Materials Irradiation Facility (IFMIF) is a proposal of D-Li intense neutron source to cover all aspects of the fusion materials development in the framework of IEA collaboration. The new activity has been started to qualifying the important technical issues called Key Element technology Phase since 2000. Although the neutron spectrum can be adjusted by changing the incident beam energy, it is favorable to be carried out many irradiation tasks at the same time under the unique beam condition. For designing the tailored neutron spectrum, neutron nuclear data for the moderator-reflector materials up to 50 MeV are required. The data for estimating the induced radioactivity is also required to keep the radiation level low enough at maintenance time. The candidate materials and the required accuracy of nuclear data are summarized. (author)

  19. Nuclear data needs for neutron spectrum tailoring at International Fusion Materials Irradiation Facility (IFMIF)

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    International Fusion Materials Irradiation Facility (IFMIF) is a proposal of D-Li intense neutron source to cover all aspects of the fusion materials development in the framework of IEA collaboration. The new activity has been started to qualifying the important technical issues called Key Element technology Phase since 2000. Although the neutron spectrum can be adjusted by changing the incident beam energy, it is favorable to be carried out many irradiation tasks at the same time under the unique beam condition. For designing the tailored neutron spectrum, neutron nuclear data for the moderator-reflector materials up to 50 MeV are required. The data for estimating the induced radioactivity is also required to keep the radiation level low enough at maintenance time. The candidate materials and the required accuracy of nuclear data are summarized. (author)

  20. Deuteron cross section evaluation for safety and radioprotection calculations of IFMIF/EVEDA accelerator prototype

    Energy Technology Data Exchange (ETDEWEB)

    Blideanu, Valentin [Commissariat a l' energie atomique CEA/IRFU, Centre de Saclay, 91191 Gif sur Yvette cedex (France); Garcia, Mauricio [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Instituto de Fusion Nuclear, UPM, Madrid (Spain); Joyer, Philippe, E-mail: philippe.joyer@cea.fr [Commissariat a l' energie atomique CEA/IRFU, Centre de Saclay, 91191 Gif sur Yvette cedex (France); Lopez, Daniel; Mayoral, Alicia; Ogando, Francisco [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Instituto de Fusion Nuclear, UPM, Madrid (Spain); Ortiz, Felix [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Sanz, Javier; Sauvan, Patrick [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Instituto de Fusion Nuclear, UPM, Madrid (Spain)

    2011-10-01

    In the frame of IFMIF/EVEDA activities, a prototype accelerator delivering a high power deuteron beam is under construction in Japan. Interaction of these deuterons with matter will generate high levels of neutrons and induced activation, whose predicted yields depend strongly on the models used to calculate the different cross sections. A benchmark test was performed to validate these data for deuteron energies up to 20 MeV and to define a reasonable methodology for calculating the cross sections needed for EVEDA. Calculations were performed using the nuclear models included in MCNPX and PHITS, and the dedicated nuclear model code TALYS. Although the results obtained using TALYS (global parameters) or Monte Carlo codes disagree with experimental values, a solution is proposed to compute cross sections that are a good fit to experimental data. A consistent computational procedure is also suggested to improve both transport simulations/prompt dose and activation/residual dose calculations required for EVEDA.

  1. Current status of the technology development on lithium safety handling under IFMIF/EVEDA

    Energy Technology Data Exchange (ETDEWEB)

    Furukawa, Tomohiro, E-mail: furukawa.tomohiro@jaea.go.jp; Hirakawa, Yasushi; Kato, Shoichi; Iijima, Minoru; Ohtaka, Masahiko; Kondo, Hiroo; Kanemura, Takuji; Wakai, Eiichi

    2014-12-15

    Studies for establishing technology for the safe handling of lithium was performed in the Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Material Irradiation Facility (IFMIF). This research comprises four tasks: (a) extinguishing lithium files, (b) chemical reactions of lithium on the event of a leak, (c) lithium removal from the components, and (d) chemical analysis of impurities in lithium. Tasks (a) and (b), related to functions on the event of a lithium leak, involve selection of the material suitable for extinguishing lithium fires and assessment of corrosive effects of leaked lithium on materials at high temperature, respectively. Task (c) involves evaluation of methods for the replacement and/or decommissioning of the lithium components. Task (d) constitutes the development of high-precision techniques for the determination of impurities in lithium, particularly the dominating corrosive impurity—dissolved nitrogen. Experimental results addressing the objectives of each of the tasks are described in this communication.

  2. Certification of contact probe measurement of surface wave of Li jet for IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Okita, Takafumi, E-mail: okita@stu.nucl.eng.osaka-u.ac.jp [Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka (Japan); Hoashi, Eiji; Yoshihashi, Sachiko [Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka (Japan); Kondo, Hiroo; Kanemura, Takuji [Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki (Japan); Yamaoka, Nobuo; Horiike, Hiroshi [Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka (Japan)

    2015-10-15

    Highlights: • We have conducted experiments of liquid lithium free-surface flow for IFMIF. • In the experiment using electro-contact probe apparatus, a droplet of liquid Li on the middle of measurement probe was observed. • Behavior of a droplet and false detections were observed by using HSV camera. • The error of the statistical result was roughly evaluated about 1%. • From results of numerical simulations, we obtained the detailed information about the behavior of a Li droplet. - Abstract: The international fusion material irradiation facility (IFMIF) is a neutron source for developing fusion reactor materials. A liquid lithium (Li) jet with free surface is planned as a target to generate intense neutron field. It is important to obtain information on the surface wave characteristic for safety of the facility and efficient neutron generation. Surface wave characteristics experiment using the liquid Li circulation facility is carried out at Osaka University. In our studies, measurement using an electro-contact probe apparatus is conducted and many data about surface wave height were taken. In this experiment, a liquid Li droplet was observed on the probe. To see effect due to droplets on the probe needle, images near the surface of the Li jet including the Li droplet were taken by HSV camera synchronized with probe contact signals, and correlation between the behavior of the Li droplet and signals was evaluated. From the results, when the droplet on the probe contacts of the droplet with the surface, signals obviously different from the regular signal were observed. The influence on the result of frequency was estimated and is approximately <1%. Accuracy of measurement using probe could be increased by carefully deleting false signals.

  3. Numerical study of the flow conditioner for the IFMIF liquid lithium target

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S., E-mail: sergej.gordeev@kit.edu [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute for Thechnology, Campus North, Hermann v. Helmholtz Platz 1, D76344, Eggenstein-Leopoldshafen (Germany); Gröschel, F. [KIT Fusion Program, Karlsruhe Institute for Thechnology, Campus North, Hermann v. Helmholtz Platz 1, D76344, Eggenstein-Leopoldshafen (Germany); Heinzel, V.; Hering, W.; Stieglitz, R. [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute for Thechnology, Campus North, Hermann v. Helmholtz Platz 1, D76344, Eggenstein-Leopoldshafen (Germany)

    2014-10-15

    Highlights: • A detailed numerical analysis of the flow conditioner efficiency has been performed. • The calculations show that the present design of the flow conditioner cannot suppress swirl motions emerging from the bend. • The transient simulation reveals flow instabilities between the separation zone and the accelerated outer region. • Calculation shows that pitched guide vanes upstream the elbow reduces a generation of backflow areas downstream. - Abstract: IFMIF (International Fusion Materials Irradiation Facility) is an accelerator-based deuteron–lithium (D–Li) neutron source to simulate the neutron irradiation field in a fusion reactor. The target assembly of the IFMIF consists of the flow conditioners and the nozzle, which has to form a stable lithium jet. This work focuses on a numerical study of the flow conditioner efficiency, in which two different types of flow conditioners are compared by means of a detailed numerical analysis with respect to specific hydraulic effects in the pipe elbow and the inflow conditioners. The adequateness of three different turbulence models to simulate a flow through a 90° bend of circular cross section has been examined. The calculations show that a honeycomb-screen combination is not capable to suppress effectively large scale swirl motions emerging from the bend. An increasing number of screens improves the flow uniformity downstream, but increases the pressure drop. In order to detect any transient effects in the separation area a flow straightener configuration consisting of a honeycomb with a subsequent screen has been analyzed by means of a detached eddy simulation (DES). A frequency analysis of the normalized static pressure amplitude conducted by means of a detached eddy simulation (DES) reveals instabilities in the shear layer between the separation zone and the accelerated outer region, which additionally increase the inhomogeneity of the axial velocity distribution. A set of six circumferentially

  4. Effect of activation cross section uncertainties in transmutation analysis of realistic low-activation steels for IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Cabellos, O.; Garcya-Herranz, N.; Sanz, J. [Institute of Nuclear Fusion, UPM, Madrid (Spain); Cabellos, O.; Garcya-Herranz, N.; Fernandez, P.; Fernandez, B. [Dept. of Nuclear Engineering, UPM, Madrid (Spain); Sanz, J. [Dept. of Power Engineering, UNED, Madrid (Spain); Reyes, S. [Safety, Environment and Health Group, ITER Joint Work Site, Cadarache Center (France)

    2008-07-01

    We address uncertainty analysis to draw conclusions on the reliability of the activation calculation in the International Fusion Materials Irradiation Facility (IFMIF) under the potential impact of activation cross section uncertainties. The Monte Carlo methodology implemented in ACAB code gives the uncertainty estimates due to the synergetic/global effect of the complete set of cross section uncertainties. An element-by-element analysis has been demonstrated as a helpful tool to easily analyse the transmutation performance of irradiated materials.The uncertainty analysis results showed that for times over about 24 h the relative error in the contact dose rate can be as large as 23 per cent. We have calculated the effect of cross section uncertainties in the IFMIF activation of all different elements. For EUROFER, uncertainties in H and He elements are 7.3% and 5.6%, respectively. We have found significant uncertainties in the transmutation response for C, P and Nb.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  6. Adjoint sensitivity analysis of dynamic reliability models based on Markov chains - II: Application to IFMIF reliability assessment

    Energy Technology Data Exchange (ETDEWEB)

    Cacuci, D. G. [Commiss Energy Atom, Direct Energy Nucl, Saclay, (France); Cacuci, D. G.; Balan, I. [Univ Karlsruhe, Inst Nucl Technol and Reactor Safetly, Karlsruhe, (Germany); Ionescu-Bujor, M. [Forschungszentrum Karlsruhe, Fus Program, D-76021 Karlsruhe, (Germany)

    2008-07-01

    In Part II of this work, the adjoint sensitivity analysis procedure developed in Part I is applied to perform sensitivity analysis of several dynamic reliability models of systems of increasing complexity, culminating with the consideration of the International Fusion Materials Irradiation Facility (IFMIF) accelerator system. Section II presents the main steps of a procedure for the automated generation of Markov chains for reliability analysis, including the abstraction of the physical system, construction of the Markov chain, and the generation and solution of the ensuing set of differential equations; all of these steps have been implemented in a stand-alone computer code system called QUEFT/MARKOMAG-S/MCADJSEN. This code system has been applied to sensitivity analysis of dynamic reliability measures for a paradigm '2-out-of-3' system comprising five components and also to a comprehensive dynamic reliability analysis of the IFMIF accelerator system facilities for the average availability and, respectively, the system's availability at the final mission time. The QUEFT/MARKOMAG-S/MCADJSEN has been used to efficiently compute sensitivities to 186 failure and repair rates characterizing components and subsystems of the first-level fault tree of the IFMIF accelerator system. (authors)

  7. Adjoint sensitivity analysis of dynamic reliability models based on Markov chains - II: Application to IFMIF reliability assessment

    International Nuclear Information System (INIS)

    Cacuci, D. G.; Cacuci, D. G.; Balan, I.; Ionescu-Bujor, M.

    2008-01-01

    In Part II of this work, the adjoint sensitivity analysis procedure developed in Part I is applied to perform sensitivity analysis of several dynamic reliability models of systems of increasing complexity, culminating with the consideration of the International Fusion Materials Irradiation Facility (IFMIF) accelerator system. Section II presents the main steps of a procedure for the automated generation of Markov chains for reliability analysis, including the abstraction of the physical system, construction of the Markov chain, and the generation and solution of the ensuing set of differential equations; all of these steps have been implemented in a stand-alone computer code system called QUEFT/MARKOMAG-S/MCADJSEN. This code system has been applied to sensitivity analysis of dynamic reliability measures for a paradigm '2-out-of-3' system comprising five components and also to a comprehensive dynamic reliability analysis of the IFMIF accelerator system facilities for the average availability and, respectively, the system's availability at the final mission time. The QUEFT/MARKOMAG-S/MCADJSEN has been used to efficiently compute sensitivities to 186 failure and repair rates characterizing components and subsystems of the first-level fault tree of the IFMIF accelerator system. (authors)

  8. The design status of the liquid lithium target facility of IFMIF at the end of the engineering design activities

    Energy Technology Data Exchange (ETDEWEB)

    Nitti, F.S., E-mail: francesco.nitti@enea.it [IFMIF/EVEDA Project Team, Rokkasho Japan (Japan); Ibarra, A. [CIEMAT, Madrid (Spain); Ida, M. [IHI Corporation, Tokyo (Japan); Favuzza, P. [ENEA Research Center Firenze (Italy); Furukawa, T. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Groeschel, F. [KIT Research Center, Karlsruhe (Germany); Heidinger, R. [F4E Research Center, Garching (Germany); Kanemura, T. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Knaster, J. [IFMIF/EVEDA Project Team, Rokkasho Japan (Japan); Kondo, H. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Micchiche, G. [ENEA Research Center, Brasimone (Italy); Sugimoto, M. [JAEA Research Center, Rokkasho Japan (Japan); Wakai, E. [JAEA Research Center, Tokai-mura, Ibaraki (Japan)

    2015-11-15

    Highlights: • Results of validation and design activity for the Li loop facility of IFMIF. • Demonstration of Li target stability, with surface disturbance <1 mm. • Demonstration of start-up and shut down procedures of Li loop. • Complete design of the heat removal system and C and O purification system. • Conceptual design of N and H isotopes purification systems. - Abstract: The International Fusion Material Irradiation Facility (IFMIF) is an experimental facility conceived for qualifying and characterizing structural materials for nuclear fusion applications. The Engineering Validation and Engineering Design Activity (EVEDA) is a fundamental step towards the final design. It presented two mandates: the Engineering Validation Activities (EVA), still on-going, and the Engineering Design Activities (EDA) accomplished on schedule in June 2013. Five main facilities are identified in IFMIF, among which the Lithium Target Facility constituted a technological challenge overcome thanks to the success of the main validation challenges impacting the design. The design of the liquid Lithium Target Facility at the end of the EDA phase is here detailed.

  9. The design status of the liquid lithium target facility of IFMIF at the end of the engineering design activities

    International Nuclear Information System (INIS)

    Nitti, F.S.; Ibarra, A.; Ida, M.; Favuzza, P.; Furukawa, T.; Groeschel, F.; Heidinger, R.; Kanemura, T.; Knaster, J.; Kondo, H.; Micchiche, G.; Sugimoto, M.; Wakai, E.

    2015-01-01

    Highlights: • Results of validation and design activity for the Li loop facility of IFMIF. • Demonstration of Li target stability, with surface disturbance <1 mm. • Demonstration of start-up and shut down procedures of Li loop. • Complete design of the heat removal system and C and O purification system. • Conceptual design of N and H isotopes purification systems. - Abstract: The International Fusion Material Irradiation Facility (IFMIF) is an experimental facility conceived for qualifying and characterizing structural materials for nuclear fusion applications. The Engineering Validation and Engineering Design Activity (EVEDA) is a fundamental step towards the final design. It presented two mandates: the Engineering Validation Activities (EVA), still on-going, and the Engineering Design Activities (EDA) accomplished on schedule in June 2013. Five main facilities are identified in IFMIF, among which the Lithium Target Facility constituted a technological challenge overcome thanks to the success of the main validation challenges impacting the design. The design of the liquid Lithium Target Facility at the end of the EDA phase is here detailed.

  10. Relationship Between Final Performance and Block Times with the Traditional and the New Starting Platforms with A Back Plate in International Swimming Championship 50-M and 100-M Freestyle Events

    Science.gov (United States)

    Garcia-Hermoso, Antonio; Escalante, Yolanda; Arellano, Raul; Navarro, Fernando; Domínguez, Ana M.; Saavedra, Jose M.

    2013-01-01

    The purpose of this study was to investigate the association between block time and final performance for each sex in 50-m and 100-m individual freestyle, distinguishing between classification (1st to 3rd, 4th to 8th, 9th to 16th) and type of starting platform (old and new) in international competitions. Twenty-six international competitions covering a 13-year period (2000-2012) were analysed retrospectively. The data corresponded to a total of 1657 swimmers’ competition histories. A two-way ANOVA (sex x classification) was performed for each event and starting platform with the Bonferroni post-hoc test, and another two-way ANOVA for sex and starting platform (sex x starting platform). Pearson’s simple correlation coefficient was used to determine correlations between the block time and the final performance. Finally, a simple linear regression analysis was done between the final time and the block time for each sex and platform. The men had shorter starting block times than the women in both events and from both platforms. For 50-m event, medalists had shorter block times than semi- finalists with the old starting platforms. Block times were directly related to performance with the old starting platforms. With the new starting platforms, however, the relationship was inverse, notably in the women’s 50-m event. The block time was related for final performance in the men’s 50- m event with the old starting platform, but with the new platform it was critical only for the women’s 50-m event. Key Points The men had shorter block times than the women in both events and with both platforms. For both distances, the swimmers had shorter block times in their starts from the new starting platform with a back plate than with the old platform. For the 50-m event with the old starting platform, the medalists had shorter block times than the semi-finalists. The new starting platform block time was only determinant in the women’s 50-m event. In order to improve

  11. 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.

  12. Engineering design of the IFMIF EVEDA reference test cell and key components

    Energy Technology Data Exchange (ETDEWEB)

    Tian, Kuo, E-mail: kuo.tian@kit.edu [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology, Karlsruhe (Germany); Arbeiter, Frederik; Chen, Yuming; Heinzel, Volker; Kondo, Keitaro [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology, Karlsruhe (Germany); Mittwollen, Martin [Institute for Material Handling and Logistics, Karlsruhe Institute of Technology, Karlsruhe (Germany)

    2014-10-15

    The latest design updates of the IFMIF-EVEDA reference test cell (TC) are described with emphasis on the following key components: active cooling pipes for concrete biological shielding walls and stainless steel liner, TC gas leak tight boundary, and piping and cabling inside TC and between TC and the access cell (AC). Water cooling is adopted for concrete shielding walls and the liner. Buried pipes are selected for active cooling of the TC surrounding shielding walls; directly welded pipes on the liner are used to remove nuclear heat of the liner. Technical features and layout of the cooling pipes are preliminary defined. The TC vacuum boundary, which includes the TC liner, an independent TC cover plate, a rubber based sealing gasket, and welding seams between interface shielding plugs and TC liner, is described. Engineering design of the piping and cabling plugs as well as the arrangement of pipes and cables under the TC covering plate and the AC floor are updated. Pipes and cable tunnels inside the shielding plugs are arranged with several bends for minimizing neutron streaming from inside to outside of the TC. Pipes, cables, and the corresponding penetrations between the TC and the AC are carefully arranged for convenient access and maintenances.

  13. Designs of contraction nozzle and concave back-wall for IFMIF target

    Energy Technology Data Exchange (ETDEWEB)

    Ida, Mizuho E-mail: ida@ifmif.tokai.jaeri.go.jp; Nakamura, Hideo; Nakamura, Hiroo; Takeuchi, Hiroshi

    2004-02-01

    For the liquid lithium flow target of International Fusion Materials Irradiation Facility (IFMIF), the double reducer (two-step contraction) nozzle with a high-contraction ratio of 10 which generated high-speed uniform jet flows up to 20 m/s was proposed. Multi-dimensional hydraulic analyses were carried out to verify the performance of the proposed nozzle. The analytical results showed that the double reducer nozzle would well generate high-speed uniform flow, while one-step contraction nozzle generated non-uniform flow and resulted in flow thickening at the beam footprint. For the target design, the range of the concave back-wall radius with no lithium boiling due to the centrifugal force and proper component arrangement in the irradiation test cell was determined by the thermal-hydraulic analysis of a free-surface flow. It was verified that the back-wall radius from 0.25 to 10 m was acceptable in the velocity range of 10-20 m/s.

  14. Designs of contraction nozzle and concave back-wall for IFMIF target

    International Nuclear Information System (INIS)

    Ida, Mizuho; Nakamura, Hideo; Nakamura, Hiroo; Takeuchi, Hiroshi

    2004-01-01

    For the liquid lithium flow target of International Fusion Materials Irradiation Facility (IFMIF), the double reducer (two-step contraction) nozzle with a high-contraction ratio of 10 which generated high-speed uniform jet flows up to 20 m/s was proposed. Multi-dimensional hydraulic analyses were carried out to verify the performance of the proposed nozzle. The analytical results showed that the double reducer nozzle would well generate high-speed uniform flow, while one-step contraction nozzle generated non-uniform flow and resulted in flow thickening at the beam footprint. For the target design, the range of the concave back-wall radius with no lithium boiling due to the centrifugal force and proper component arrangement in the irradiation test cell was determined by the thermal-hydraulic analysis of a free-surface flow. It was verified that the back-wall radius from 0.25 to 10 m was acceptable in the velocity range of 10-20 m/s

  15. Reduced cost design of liquid lithium target for international fusion material irradiation facility (IFMIF)

    International Nuclear Information System (INIS)

    Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki

    2001-01-01

    The International Fusion Materials Irradiation Facility (IFMIF) is being jointly planned to provide an accelerator-based D-Li neutron source to produce intense high energy neutrons (2 MW/m 2 ) up to 200 dpa and a sufficient irradiation volume (500 cm 3 ) for testing the candidate materials and components up to about a full lifetime of their anticipated use in ITER and DEMO. To realize such a condition, 40 MeV deuteron beam with a current of 250 mA is injected into high speed liquid lithium flow with a speed of 20 m/s. Following Conceptual Design Activity (1995-1998), a design study with focus on cost reduction without changing its original mission has been done in 1999. The following major changes to the CAD target design have been considered in the study and included in the new design: i) number of the Li target has been changed from 2 to 1, ii) spare of impurity traps of the Li loop was removed although the spare will be stored in a laboratory for quick exchange, iii) building volume was reduced via design changes in lithium loop length. This paper describes the reduced cost design of the lithium target system and recent status of Key Element Technology activities. (author)

  16. Neutron production and dose rate in the IFMIF/EVEDA LIPAc injector beam commissioning

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Keitaro, E-mail: kondo.keitaro@jaea.go.jp [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori (Japan); Narita, Takahiro; Usami, Hiroki; Takahashi, Hiroki; Ochiai, Kentaro; Shinto, Katsuhiro; Kasugai, Atsushi [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori (Japan); Okumura, Yoshikazu [IFMIF/EVEDA Project Team, Rokkasho-mura, Kamikita-gun, Aomori (Japan)

    2016-11-01

    Highlights: • A dedicated neutron production yield monitoring system for LIPAc has been developed. • The biological dose rate during operation of the LIPAc injector was analyzed. • The neutron streaming effect due to penetrations in the shielding wall was investigated. - Abstract: The construction of the Linear IFMIF Prototype Accelerator (LIPAc) is in progress in Rokkasho, Japan, and the deuteron beam commissioning of the injector began in July 2015. Due to the huge beam current of 125 mA, a large amount of d-D neutrons are produced in the commissioning. The neutron streaming effect through pipe penetrations and underground pits may dominate the radiation dose at the outside of the accelerator vault during the injector operation. In the present study the effective dose rate expected during the injector commissioning was analyzed by a Monte Carlo calculation and compared with the measured value. For the comparison it is necessary to know the total neutron production yield in the accelerator vault, thus a dedicated neutron production yield monitoring system was developed. The yield obtained was smaller than that previously reported in a literature by a factor of a few and seems to depend on some beam conditions. From the comparison it was proved that the calculation always provides a conservative estimate and the dose rates in places where occupational works can always access and the controlled area boundary are expected to be far less than the legal criteria throughout the injector commissioning.

  17. Gamma-ray and neutron area monitoring system of linear IFMIF prototype accelerator building

    International Nuclear Information System (INIS)

    Takahashi, Hiroki; Kojima, Toshiyuki; Narita, Takahiro; Tsutsumi, Kazuyoshi; Maebara, Sunao; Sakaki, Hironao; Nishiyama, Koichi

    2013-01-01

    Highlights: • Area monitoring system and control system are needed for LIPAc radiation management. • To secure the radiation safety, these systems are linked with two kinds of data path. • Hardwired data paths are adopted to realize the fast transfer of interlock signals. • Dual LAN and shared memory are adopted to the reliable transfer of monitoring data. • Data transfers without unnecessary load are designed and configured for these systems. -- Abstract: The linear IFMIF prototype accelerator (LIPAc) produces deuteron beam with 1 MW power. Since huge number of neutrons occur from such a high power beam, therefore, it is important for the radiation management to design a high reliability area monitoring system for gamma-rays and neutrons. To obtain the valuable operation data of the high-power deuteron beam at LIPAc, it is important to link and record the beam operation data and the area monitoring data. We realize the reliable data transfer to provide the area monitoring data to the accelerator control system which needs a high reliability using the shared-memory data link method. This paper describes the area monitoring system in the LIPAc building and the data-link between this system and the LIPAc control system

  18. Development status of data acquisition system for IFMIF/EVEDA accelerator

    International Nuclear Information System (INIS)

    Usami, Hiroki; Takahashi, Hiroki; Komukai, Satoshi

    2015-01-01

    EU and JAEA are advancing development of Linear IFMIF Prototype Accelerator (LIPAc) control system jointly, but JAEA keeps developing central control system (CCS) mainly. Data transfer during an equipment control system of CCS and EU is performed through EPICS. JAEA is using PostgreSQL as 1 of development elements in CCS and is advancing development of the system to record the whole EPICS data of LIPAc (the data acquisition system). On the other hand, a data acquisition is performed using BEAUTY (Best Ever Archive Toolset, yet) in an element test of equipment at Europe. Therefore '1 client refers to collected data by more than one server machine' with 'compatibility securement of data with BEAUTY' in case of development of the data acquisition system of CCS, and, it's necessary to consider 'To do a data acquisition and backup work at the same time.' For the moment, former 2 are in progress. And a demonstration of the data acquisition system is being performed simultaneously with commissioning in injector. The data acquisition system is collecting data of injector other ones, and the data reference by a monitor with CSS (Control System Studio) is also possible. We will report on the current state of the development of the data acquisition system by making reference to a result of the test by injector commissioning. (author)

  19. Accessibility evaluation of the IFMIF liquid lithium loop considering activated erosion/corrosion materials deposition

    International Nuclear Information System (INIS)

    Nakamura, H.; Takemura, M.; Yamauchi, M.; Fischer, U.; Ida, M.; Mori, S.; Nishitani, T.; Simakov, S.; Sugimoto, M.

    2005-01-01

    This paper presents an evaluation of accessibility of the Li loop piping considering activated corrosion product. International Fusion Materials Irradiation Facility (IFMIF) is a deuteron-lithium (Li) stripping reaction neutron source for fusion materials testing. Target assembly and back wall are designed as fully remote maintenance component. Accessibility around the Li loop piping will depend on activation level of the deposition materials due to the back wall erosion/corrosion process under liquid Li flow. Activation level of the corrosion products coming from the AISI 316LN back wall is calculated by the ACT-4 of the THIDA-2 code system. The total activities after 1 day, 1 week, 1 month and 1 year cooling are 3.1 x 10 14 , 2.8 x 10 14 , 2.3 x 10 14 and 7.5 x 10 13 Bq/kg, respectively. Radiation dose rate around the Li loop pipe is calculated by QAD-CGGP2R code. Activated area of the back wall is 100 cm 2 . Corrosion rate is assumed 1 μm/year. When 10% of the corrosion material is supposed to be deposited on the inner surface of the pipe, the dose rate is calculated to be less than a permissible level of 10 μSv/h for hands-on maintenance, therefore, the maintenance work is assessed to be possible

  20. Method to reduce damage to backing plate

    Science.gov (United States)

    Perry, Michael D.; Banks, Paul S.; Stuart, Brent C.

    2001-01-01

    The present invention is a method for penetrating a workpiece using an ultra-short pulse laser beam without causing damage to subsequent surfaces facing the laser. Several embodiments are shown which place holes in fuel injectors without damaging the back surface of the sack in which the fuel is ejected. In one embodiment, pulses from an ultra short pulse laser remove about 10 nm to 1000 nm of material per pulse. In one embodiment, a plasma source is attached to the fuel injector and initiated by common methods such as microwave energy. In another embodiment of the invention, the sack void is filled with a solid. In one other embodiment, a high viscosity liquid is placed within the sack. In general, high-viscosity liquids preferably used in this invention should have a high damage threshold and have a diffusing property.

  1. Assessment of plastic flow and fracture properties with small specimens test techniques for IFMIF-designed specimens

    International Nuclear Information System (INIS)

    Spaetig, P.; Campitelli, E.N.; Bonade, R.; Baluc, N.

    2005-01-01

    The primary mission of the International Fusion Material Irradiation Facility (IFMIF) is to generate a material database to be used for the design of various components, for the licensing and for the assessment of the safe operation of a demonstration fusion reactor. IFMIF is an accelerator-based high-energy neutron source whose irradiation volume is quite limited (0.5 l for the high fluence volume). This requires the use of small specimens to measure the irradiation-induced changes on the physical and mechanical properties of materials. In this paper, we developed finite element models to better analyze the results obtained with two different small specimen test techniques applied to the tempered martensitic steel F82H-mod. First, one model was used to reconstruct the load-deflection curves of small ball punch tests, which are usually used to extract standard tensile parameters. It was shown that a reasonable assessment of the overall plastic flow can be done with small ball punch tests. Second, we investigated the stress field sensitivity at a crack tip to the constitutive behavior, for a crack modeled in plane strain, small-scale yielding and fracture mode I conditions. Based upon a local criterion for cleavage, that appears to be the basis to account for the size and geometry effects on fracture toughness, we showed that the details of the constitutive properties play a key role in modeling the irradiation-induced fracture toughness changes. Consequently, we suggest that much more attention and efforts have to be paid in investigating the post-yield behavior of the irradiated specimens and, in order to reach this goal, we recommend the use of not only tensile specimens but also that of compression ones in the IFMIF irradiation matrices. (author)

  2. IFMIF - Layout and arrangement of cells according to requirements of technical logistics, reliability and remote handling

    Energy Technology Data Exchange (ETDEWEB)

    Mittwollen, Martin, E-mail: martin.mittwollen@kit.edu [Karlsruhe Institute of Technology, Institute for Conveying Technology and Logistics, Karlsruhe (Germany); Eilert, Dirk; Kubaschewski, Martin; Madzharov, Vladimir [Karlsruhe Institute of Technology, Institute for Conveying Technology and Logistics, Karlsruhe (Germany); Tian Kuo [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Karlsruhe (Germany)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer In a first approach, layout and arrangement of the cells followed a predetermined plant layout. Black-Right-Pointing-Pointer Disadvantages in technical logistics, reliability and remote handling have been detected. Black-Right-Pointing-Pointer Deliberation with project teams opened space for improvements. Black-Right-Pointing-Pointer Layout and arrangement of cells have been improved by simplification of design. Black-Right-Pointing-Pointer Speed and reliability have been increased significantly. - Abstract: The International Fusion Material Irradiation Facility (IFMIF) is designed to study and qualify structural and functional materials which shall be used in future fusion nuclear power plants. During the current engineering validation and engineering design activities (EVEDA) phase the development of e.g. an optimized layout and arrangement of the cells (Access Cell, Test Cell, and Test Module Handling Cells) is of major interest. After defining different functions for the individual cells like e.g. large scale/fine scale disassembling of test modules a first layout has been developed. This design followed requirements like having a minimum of carrier changes to avoid sources of failures. On the other hand it has had to be a compact arrangement of cells due to restrictions from plant layout. A row of changes of transfer direction, and different crane systems were the consequence. Constructive discussion with project team results in the statement, that for reasons of being reliable and fast, layout and arrangement of cells goes first, plant layout then will follow. The chance for big improvements was taken and the result was a simplified design with strong reduced number of functional elements, and increased reliability and speed.

  3. Optimised design and thermal-hydraulic analysis of the IFMIF/HFTM test section

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S.; Heinzel, V.; Lang, K.H.; Moeslang, A.; Schleisiek, K.; Slobodtchouk, V.; Stratmanns, E.

    2003-10-01

    On the basis of previous concepts, analyses and experiments, the high flux test module (HFTM) for the International Fusion Materials Irradiation Facility (IFMIF) was further optimised. The work focused on the design and the thermal hydraulic analysis of the HFTM section containing the material specimens to be irradiated, the ''test section'', with the main objective to improve the concept with respect to the optimum use of the available irradiation volume and to the temperature of the specimens. Particular emphasis was laid on the application of design principles which assure stable and reproducible thermal conditions. The present work has confirmed the feasibility and suitability of the optimised design of the HFTM test section with chocolate plate like shaped rigs. In particular it has been shown that the envisaged irradiation temperatures can be reached with acceptable temperature differences inside the specimen stack. The latter can be achieved only by additional electrical heating of the axial ends of the capsules. Division of the heater in three sections with separate power supply and control units is necessary. Maintaining of the temperatures during beam-off periods likewise requires electrical heating. The required electrical heaters - mineral isolated wires - are commercially available. The potential of the CFD code STAR-CD for the thermal hydraulic analysis of complex systems like the HFTM was confirmed. Nevertheless, experimental confirmation is desirable. Suitable experiments are under preparation. To verify the assumptions made on the thermal conductivity of the contact faces and layers between the two shells of the rig, dedicated experiments are suggested. The present work must be complemented by a thermal mechanical analysis of the module. Most critical component in this respect seems to be the rig wall. Furthermore, it will be necessary to investigate the response of the HFTM to power transients, and to determine the requirements

  4. Numerical analysis of free surface instabilities in the IFMIF lithium target

    International Nuclear Information System (INIS)

    Gordeev, S.; Heinzel, V.; Moeslang, A.

    2007-01-01

    The International Fusion Materials Facility (IFMIF) facility uses a high speed (10-20 m/s) Lithium (Li) jet flow as a target for two 40 MeV/125 mA deuteron beams. The major function of the Li target is to provide a stable Li jet for the production of an intense neutron flux. For the understanding the lithium jet behaviour and elimination of the free-surface flow instabilities a detailed analysis of the Li jet flow is necessary. Different kinds of instability mechanisms in the liquid jet flow have been evaluated and classified based on analytical and experimental data. Numerical investigations of the target free surface flow have been performed. Previous numerical investigations have shown in principle the suitability of CFD code Star- CD for the simulation of the Li-target flow. The main objective of this study is detailed numerical analysis of instabilities in the Li-jet flow caused by boundary layer relaxation near the nozzle exit, transition to the turbulence flow and back wall curvature. A number of CFD models are developed to investigate the formation of instabilities on the target surface. Turbulence models are validated on the experimental data. Experimental observations have shown that the change of the nozzle geometry at the outlet such as a slight divergence of the nozzle surfaces or nozzle edge defects causes the flow separation and occurrence of longitudinal periodic structures on the free surface with an amplitude up to 5 mm. Target surface fluctuations of this magnitude can lead to the penetration of the deuteron beam in the target structure and cause the local overheating of the back plat. Analysis of large instabilities in the Li-target flow combined with the heat distribution in lithium depending on the free surface shape is performed in this study. (orig.)

  5. IFMIF - International Fusion Materials Irradiation Facility Conceptual Design Activity/Interim Report

    International Nuclear Information System (INIS)

    Rennich, M.J.

    1995-12-01

    Environmental acceptability, safety, and economic viability win ultimately be the keys to the widespread introduction of fusion power. This will entail the development of radiation- resistant and low- activation materials. These low-activation materials must also survive exposure to damage from neutrons having an energy spectrum peaked near 14 MeV with annual radiation doses in the range of 20 displacements per atom (dpa). Testing of candidate materials, therefore, requires a high-flux source of high energy neutrons. The problem is that there is currently no high-flux source of neutrons in the energy range above a few MeV. The goal, is therefore, to provide an irradiation facility for use by fusion material scientists in the search for low-activation and damage-resistant materials. An accellerator-based neutron source has been established through a number of international studies and workshops' as an essential step for materials development and testing. The mission of the International Fusion Materials Irradiation Facility (IFMIF) is to provide an accelerator-based, deuterium-lithium (D-Li) neutron source to produce high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials up to about a full lifetime of anticipated use in fusion energy reactors. would also provide calibration and validation of data from fission reactor and other accelerator-based irradiation tests. It would generate material- specific activation and radiological properties data, and support the analysis of materials for use in safety, maintenance, recycling, decommissioning, and waste disposal systems

  6. Optimised design and thermal-hydraulic analysis of the IFMIF/HFTM test section

    International Nuclear Information System (INIS)

    Gordeev, S.; Heinzel, V.; Lang, K.H.; Moeslang, A.; Schleisiek, K.; Slobodtchouk, V.; Stratmanns, E.

    2003-10-01

    On the basis of previous concepts, analyses and experiments, the high flux test module (HFTM) for the International Fusion Materials Irradiation Facility (IFMIF) was further optimised. The work focused on the design and the thermal hydraulic analysis of the HFTM section containing the material specimens to be irradiated, the ''test section'', with the main objective to improve the concept with respect to the optimum use of the available irradiation volume and to the temperature of the specimens. Particular emphasis was laid on the application of design principles which assure stable and reproducible thermal conditions. The present work has confirmed the feasibility and suitability of the optimised design of the HFTM test section with chocolate plate like shaped rigs. In particular it has been shown that the envisaged irradiation temperatures can be reached with acceptable temperature differences inside the specimen stack. The latter can be achieved only by additional electrical heating of the axial ends of the capsules. Division of the heater in three sections with separate power supply and control units is necessary. Maintaining of the temperatures during beam-off periods likewise requires electrical heating. The required electrical heaters - mineral isolated wires - are commercially available. The potential of the CFD code STAR-CD for the thermal hydraulic analysis of complex systems like the HFTM was confirmed. Nevertheless, experimental confirmation is desirable. Suitable experiments are under preparation. To verify the assumptions made on the thermal conductivity of the contact faces and layers between the two shells of the rig, dedicated experiments are suggested. The present work must be complemented by a thermal mechanical analysis of the module. Most critical component in this respect seems to be the rig wall. Furthermore, it will be necessary to investigate the response of the HFTM to power transients, and to determine the requirements on the electrical

  7. Engineering design of IFMIF/EVEDA lithium test loop. Electro-magnetic pump and pressure drop

    International Nuclear Information System (INIS)

    Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Iuchi, Hiroshi; Kanemura, Takuji; Ida, Mizuho; Watanabe, Kazuyoshi; Wakai, Eiichi; Nakamura, Kazuyuki; Horiike, H.; Yamaoka, N.; Matsushita, I.

    2011-01-01

    The Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) is proceeding as one of the ITER Broader Approach (ITER-BA). A Li circulation loop for testing hydraulic stability of the Li target (high speed free-surface flow of liquid Li as a beam target) and Li purification traps are under construction in the Japan Atomic Energy Agency as a major Japanese activities in the EVEDA. This paper presents specification of an electro-magnetic pump (EMP) for the EVEDA Li Test Loop (ELTL) and evaluation of the pressure drop in the main loop of the ELTL. The EMP circulates the liquid Li at a large flow rate up to 0.05 m 3 /s (3000 l/min) under a vacuum cover gas (Ar) pressure of 10 -3 Pa, thus the evaluation of cavitation generation is a crucial issue. The EMP used in the ELTL consists of two EMPs aligned in series through a U-tube whose size of one EMP is 0.8 m square and 2.6 m in length. The calculation of the pressure drop in the main Li loop to the EMP is approx. 25 kPa at the design maximum flow rate of 0.05 m 3 /s. On the other hand the height from the EMP to a Li tank to supply Li to the EMP is designed to be 9.72 m, and secures a static pressure and the cavitation number of 18 kPa and 3.4 respectively at the maximum flow rate in a vacuum condition. As a result, it is confirmed to prevent cavitation at the inlet of the EMP in this design. (author)

  8. Design description and validation results for the IFMIF High Flux Test Module as outcome of the EVEDA phase

    Directory of Open Access Journals (Sweden)

    F. Arbeiter

    2016-12-01

    Full Text Available During the Engineering Validation and Engineering Design Activities (EVEDA phase (2007-2014 of the International Fusion Materials Irradiation Facility (IFMIF, an advanced engineering design of the High Flux Test Module (HFTM has been developed with the objective to facilitate the controlled irradiation of steel samples in the high flux area directly behind the IFMIF neutron source. The development process addressed included manufacturing techniques, CAD, neutronic, thermal-hydraulic and mechanical analyses complemented by a series of validation activities. Validation included manufacturing of 1:1 parts and mockups, test of prototypes in the FLEX and HELOKA-LP helium loops of KIT for verification of the thermal and mechanical properties, and irradiation of specimen filled capsule prototypes in the BR2 test reactor. The prototyping activities were backed by several R&D studies addressing focused issues like handling of liquid NaK (as filling medium and insertion of Small Specimen Test Technique (SSTT specimens into the irradiation capsules. This paper provides an up-todate design description of the HFTM irradiation device, and reports on the achieved performance criteria related to the requirements. Results of the validation activities are accounted for and the most important issues for further development are identified.

  9. Availability, reliability and logistic support studies of the RF power system design options for the IFMIF accelerator

    International Nuclear Information System (INIS)

    Bargallo, E.; Giralt, A.; Martinez, G.; Weber, M.; Regidor, D.; Arroyo, J.M.; Abal, J.; Dies, J.; Tapia, C.; De Blas, A.; Mendez, P.; Ibarra, A.; Molla, J.

    2013-01-01

    Highlights: ► Current RF system design based on tetrodes chains is evaluated. ► Alternative solid state power amplifiers RF system design is analyzed. ► Both designs are compared in terms of availability, logistics and cost. ► It is concluded that solid state option presents relevant improvements. -- Abstract: The current design of the radio frequency (RF) power system for the International Fusion Materials Irradiation Facility (IFMIF) is based upon tetrodes technology. Due to the improvement in the solid state amplifiers technology, the possibility of using this option for IFMIF RF system is becoming a very competitive alternative presenting from the beginning several advantages in terms of availability, reliability and logistics. The current design based on RF tetrodes chains leads no room for substantial improvements in terms of availability being the requirement for the RF system hard to achieve. The principal goals of this paper are to use RAMI (Reliability, Availability, Maintainability and Inspectionability) analysis in the solid state amplifier design, and to compare the availability, reliability and logistic performances for both alternatives

  10. Availability, reliability and logistic support studies of the RF power system design options for the IFMIF accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Bargallo, E., E-mail: enric.bargallo-font@upc.edu [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Giralt, A.; Martinez, G. [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Weber, M.; Regidor, D.; Arroyo, J.M. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, Madrid (Spain); Abal, J.; Dies, J.; Tapia, C.; De Blas, A. [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia (UPC) Barcelona-Tech, Barcelona (Spain); Mendez, P.; Ibarra, A.; Molla, J. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, Madrid (Spain)

    2013-10-15

    Highlights: ► Current RF system design based on tetrodes chains is evaluated. ► Alternative solid state power amplifiers RF system design is analyzed. ► Both designs are compared in terms of availability, logistics and cost. ► It is concluded that solid state option presents relevant improvements. -- Abstract: The current design of the radio frequency (RF) power system for the International Fusion Materials Irradiation Facility (IFMIF) is based upon tetrodes technology. Due to the improvement in the solid state amplifiers technology, the possibility of using this option for IFMIF RF system is becoming a very competitive alternative presenting from the beginning several advantages in terms of availability, reliability and logistics. The current design based on RF tetrodes chains leads no room for substantial improvements in terms of availability being the requirement for the RF system hard to achieve. The principal goals of this paper are to use RAMI (Reliability, Availability, Maintainability and Inspectionability) analysis in the solid state amplifier design, and to compare the availability, reliability and logistic performances for both alternatives.

  11. Advanced Monte Carlo procedure for the IFMIF d-Li neutron source term based on evaluated cross section data

    International Nuclear Information System (INIS)

    Simakov, S.P.; Fischer, U.; Moellendorff, U. von; Schmuck, I.; Konobeev, A.Yu.; Korovin, Yu.A.; Pereslavtsev, P.

    2002-01-01

    A newly developed computational procedure is presented for the generation of d-Li source neutrons in Monte Carlo transport calculations based on the use of evaluated double-differential d+ 6,7 Li cross section data. A new code M c DeLicious was developed as an extension to MCNP4C to enable neutronics design calculations for the d-Li based IFMIF neutron source making use of the evaluated deuteron data files. The M c DeLicious code was checked against available experimental data and calculation results of M c DeLi and MCNPX, both of which use built-in analytical models for the Li(d, xn) reaction. It is shown that M c DeLicious along with newly evaluated d+ 6,7 Li data is superior in predicting the characteristics of the d-Li neutron source. As this approach makes use of tabulated Li(d, xn) cross sections, the accuracy of the IFMIF d-Li neutron source term can be steadily improved with more advanced and validated data

  12. Advanced Monte Carlo procedure for the IFMIF d-Li neutron source term based on evaluated cross section data

    CERN Document Server

    Simakov, S P; Moellendorff, U V; Schmuck, I; Konobeev, A Y; Korovin, Y A; Pereslavtsev, P

    2002-01-01

    A newly developed computational procedure is presented for the generation of d-Li source neutrons in Monte Carlo transport calculations based on the use of evaluated double-differential d+ sup 6 sup , sup 7 Li cross section data. A new code M sup c DeLicious was developed as an extension to MCNP4C to enable neutronics design calculations for the d-Li based IFMIF neutron source making use of the evaluated deuteron data files. The M sup c DeLicious code was checked against available experimental data and calculation results of M sup c DeLi and MCNPX, both of which use built-in analytical models for the Li(d, xn) reaction. It is shown that M sup c DeLicious along with newly evaluated d+ sup 6 sup , sup 7 Li data is superior in predicting the characteristics of the d-Li neutron source. As this approach makes use of tabulated Li(d, xn) cross sections, the accuracy of the IFMIF d-Li neutron source term can be steadily improved with more advanced and validated data.

  13. Shielding performances analysis for the IFMIF test facility based on high-fidelity Monte Carlo neutronic calculations

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Keitaro, E-mail: kondo.keitaro@jaea.go.jp; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo

    2015-10-15

    Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

  14. Radiation transport analyses for IFMIF design by the Attila software using a Monte-Carlo source model

    International Nuclear Information System (INIS)

    Arter, W.; Loughlin, M.J.

    2009-01-01

    Accurate calculation of the neutron transport through the shielding of the IFMIF test cell, defined by CAD, is a difficult task for several reasons. The ability of the powerful deterministic radiation transport code Attila, to do this rapidly and reliably has been studied. Three models of increasing geometrical complexity were produced from the CAD using the CADfix software. A fourth model was produced to represent transport within the cell. The work also involved the conversion of the Vitenea-IEF database for high energy neutrons into a format usable by Attila, and the conversion of a particle source specified in MCNP wssaformat to a form usable by Attila. The final model encompassed the entire test cell environment, with only minor modifications. On a state-of-the-art PC, Attila took approximately 3 h to perform the calculations, as a consequence of a careful mesh 'layering'. The results strongly suggest that Attila will be a valuable tool for modelling radiation transport in IFMIF, and for similar problems

  15. Shielding performances analysis for the IFMIF test facility based on high-fidelity Monte Carlo neutronic calculations

    International Nuclear Information System (INIS)

    Kondo, Keitaro; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo

    2015-01-01

    Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

  16. The accomplishment of the Engineering Design Activities of IFMIF/EVEDA: The European-Japanese project towards a Li(d,xn) fusion relevant neutron source

    Science.gov (United States)

    Knaster, J.; Ibarra, A.; Abal, J.; Abou-Sena, A.; Arbeiter, F.; Arranz, F.; Arroyo, J. M.; Bargallo, E.; Beauvais, P.-Y.; Bernardi, D.; Casal, N.; Carmona, J. M.; Chauvin, N.; Comunian, M.; Delferriere, O.; Delgado, A.; Diaz-Arocas, P.; Fischer, U.; Frisoni, M.; Garcia, A.; Garin, P.; Gobin, R.; Gouat, P.; Groeschel, F.; Heidinger, R.; Ida, M.; Kondo, K.; Kikuchi, T.; Kubo, T.; Le Tonqueze, Y.; Leysen, W.; Mas, A.; Massaut, V.; Matsumoto, H.; Micciche, G.; Mittwollen, M.; Mora, J. C.; Mota, F.; Nghiem, P. A. P.; Nitti, F.; Nishiyama, K.; Ogando, F.; O'hira, S.; Oliver, C.; Orsini, F.; Perez, D.; Perez, M.; Pinna, T.; Pisent, A.; Podadera, I.; Porfiri, M.; Pruneri, G.; Queral, V.; Rapisarda, D.; Roman, R.; Shingala, M.; Soldaini, M.; Sugimoto, M.; Theile, J.; Tian, K.; Umeno, H.; Uriot, D.; Wakai, E.; Watanabe, K.; Weber, M.; Yamamoto, M.; Yokomine, T.

    2015-08-01

    The International Fusion Materials Irradiation Facility (IFMIF), presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the frame of the Broader Approach Agreement between Europe and Japan, accomplished in summer 2013, on schedule, its EDA phase with the release of the engineering design report of the IFMIF plant, which is here described. Many improvements of the design from former phases are implemented, particularly a reduction of beam losses and operational costs thanks to the superconducting accelerator concept, the re-location of the quench tank outside the test cell (TC) with a reduction of tritium inventory and a simplification on its replacement in case of failure, the separation of the irradiation modules from the shielding block gaining irradiation flexibility and enhancement of the remote handling equipment reliability and cost reduction, and the water cooling of the liner and biological shielding of the TC, enhancing the efficiency and economy of the related sub-systems. In addition, the maintenance strategy has been modified to allow a shorter yearly stop of the irradiation operations and a more careful management of the irradiated samples. The design of the IFMIF plant is intimately linked with the EVA phase carried out since the entry into force of IFMIF/EVEDA in June 2007. These last activities and their on-going accomplishment have been thoroughly described elsewhere (Knaster J et al [19]), which, combined with the present paper, allows a clear understanding of the maturity of the European-Japanese international efforts. This released IFMIF Intermediate Engineering Design Report (IIEDR), which could be complemented if required concurrently with the outcome of the on-going EVA, will allow decision making on its construction and/or serve as the basis for the definition of the next step, aligned with the evolving needs of our fusion community.

  17. IFMIF, the European–Japanese efforts under the Broader Approach agreement towards a Li(d,xn neutron source: Current status and future options

    Directory of Open Access Journals (Sweden)

    J. Knaster

    2016-12-01

    Full Text Available The necessity of a neutron source for fusion materials research was identified already in the 70s. Though neutrons induced degradation present similarities on a mechanistic approach, thresholds energies for crucial transmutations are typically above fission neutrons spectrum. The generation of He via 56Fe (n,α 53Cr in future fusion reactors with around 12 appm/dpa will lead to swelling and structural materials embrittlement. Existing neutron sources, namely fission reactors or spallation sources lead to different degradation, attempts for extrapolation are unsuccessful given the absence of experimental observations in the operational ranges of a fusion reactor. Neutrons with a broad peak at 14MeV can be generated with Li(d,xn reactions; the technological efforts that started with FMIT in the early 80s have finally matured with the success of IFMIF/EVEDA under the Broader Approach Agreement. The status today of five technological challenges, perceived in the past as most critical, are addressed. These are: 1. the feasibility of IFMIF accelerators, 2. the long term stability of lithium flow at IFMIF nominal conditions, 3. the potential instabilities in the lithium screen induced by the 2×5 MW impacting deuteron beam, 4. the uniformity of temperature in the specimens during irradiation, and 5. the validity of data provided with small specimens. Other ideas for fusion material testing have been considered, but they possibly are either not technologically feasible if fixed targets are considered or would require the results of a Li(d,xn facility to be reliably designed. In addition, today we know beyond reasonable doubt that the cost of IFMIF, consistently estimated throughout decades, is marginal compared with the cost of a fusion reactor. The less ambitious DEMO reactor performance being considered correlates with a lower need of fusion neutrons flux; thus IFMIF with its two accelerators is possibly not needed since with only one accelerator as

  18. Production quality controls and geometric characterization of the IFMIF-RFQ modules via the usage of a Coordinate Measuring Machine

    Energy Technology Data Exchange (ETDEWEB)

    Ferrari, Luigi, E-mail: luigi.ferrari@lnl.infn.it [INFN-LNL Laboratori Nazionali di Legnaro, Legnaro (Italy); Palmieri, Antonio [INFN-LNL Laboratori Nazionali di Legnaro, Legnaro (Italy); Pepato, Adriano; Prevedello, Alessandro; Dima, Razvan; Udup, Emil [INFN-Sezione di Padova, Padova (Italy)

    2017-02-15

    Highlights: • The production phases of the IFMIF-RFQ Modules is introduced. • Metrological controls through production are described and some results reported. • Radio-Frequency test is introduced by using geometric considerations. • Results from metrology and RF test are compared. • Acceptance of the modules has been guaranteed from that comparison. - Abstract: The RFQ of the IFMIF/EVEDA project (Pérez et al., 2015) is a 9.8 m long cavity able to accelerate a 125 mA deuteron beam from the input energy of 50 keV/u to the output energy of 2.5 MeV/u. Such RFQ operates at the frequency of 175 MHz and is composed of 18 mechanical modules approximately 0.55 long each (Pepato et al., 2010) . The RFQ realization involves the I.N.F.N. Sections of Padova, Torino and Bologna, as well as the Legnaro National Laboratories (L.N.L.). The metrological measurements via CMM (Coordinate Measuring Machine) provided to be a very effective tool both for quality controls along the RFQ production phases and in the reconstruction of the cavity geometric profile for each RFQ module. The scans in the most sensitive regions with respect to RF frequency, such as modulation, tips, base-vane width and vessel height provided the values of the cavity deviations from nominal geometry to be compared with design physic-driven tolerances and with RF measurements. Moreover, the comparison between mechanical and RF measurements suggests a methodology for the geometric reconstruction of the cavity axis and determines the final machining of the end surfaces of each module in view of the coupling with the adjacent ones. In this paper a detailed description of the metrological procedures and tests and of the RFQ along its production and assembly phases will be given and it will be shown that the adopted procedure allowed the attainment of the tuning range specifications for each RFQ module.

  19. Experimental and analytical studies on high-speed plane jet along concave wall simulating IFMIF Li target flow

    International Nuclear Information System (INIS)

    Nakamura, H.; Ida, M.; Kato, Y.; Maekawa, H.; Katsuta, H.; Itoh, K.; Kukita, Y.

    1998-01-01

    As part of the conceptual design activity (CDA) of the international fusion materials irradiation facility (IFMIF), the characteristics of the high-speed liquid lithium (Li) plane jet target flow have been studied by water experiments and numerical analyses for both heating and non-heating conditions. The simulated prototypal-size water flows were stable over the entire length of ∝130 mm at the average velocity up to 17 m/s. The jet flow had a specific radial velocity profile, close to that of free-vortex flow, because of a static pressure distribution in the jet thickness due to centrifugal force. Detailed velocity measurement revealed that this flow condition is penetrating into the upstream reducer nozzle up to a distance ∼ the jet thickness. The numerical analyses using a two-dimensional Cartesian-coordinate model were successful to predict the velocity profile transient around the nozzle exit, though underestimated the development of the velocity profile and the jet thickness. (orig.)

  20. Relevance of d-D interactions on neutron and tritium production in IFMIF-EVEDA accelerator prototype

    International Nuclear Information System (INIS)

    Mayoral, A.; Sanz, J.; Sauvan, P.; Lopez, D.; Garcia, M.; Ogando, F.

    2011-01-01

    In the IFMIF-EVEDA accelerator prototype, deuterium is implanted in the components due to beam losses and in the beam dump, where the beam is stopped. The interaction of the deuterons with the deuterium previously implanted leads to the production of neutrons and tritium, which are important issues for radioprotection and safety analysis. A methodology to assess these production pathways in more realistic approach has been developed. The new tools and their main achievement are: (i) an 'effective diffusivity coefficient' (deduced from available experimental data) that enables simulation of the diffusion phase, and (ii) the MCUNED code (able to handle deuteron transport libraries) allows to simulate the transport-slowdown of deuteron/tritium (to get the concentration profiles) and the neutron/tritium productions from d-Cu and d-D for up to 9 MeV incident deuteron. The results with/without theses tools are presented and their effect on the relevance of d-D sources versus d-Cu is evaluated.

  1. IFMIF [International Fusion Materials Irradiation Facility], an accelerator-based neutron source for fusion components irradiation testing: Materials testing capabilities

    International Nuclear Information System (INIS)

    Mann, F.M.

    1988-08-01

    The International Fusion Materials Irradiation Facility (IFMIF) is proposed as an advanced accelerator-based neutron source for high-flux irradiation testing of large-sized fusion reactor components. The facility would require only small extensions to existing accelerator and target technology originally developed for the Fusion Materials Irradiation Test (FMIT) facility. At the extended facility, neutrons would be produced by a 0.1-A beam of 35-MeV deuterons incident upon a liquid lithium target. The volume available for high-flux (>10/sup 15/ n/cm/sup 2/-s) testing in IFMITF would be over a liter, a factor of about three larger than in the FMIT facility. This is because the effective beam current of 35-MeV deuterons on target can be increased by a factor of ten to 1A or more. Such an increase can be accomplished by funneling beams of deuterium ions from the radio-frequency quadruple into a linear accelerator and by taking advantage of recent developments in accelerator technology. Multiple beams and large total current allow great variety in available testing. For example, multiple simultaneous experiments, and great flexibility in tailoring spatial distributions of flux and spectra can be achieved. 5 refs., 2 figs., 1 tab

  2. Development of lithium target system in engineering validation and engineering design activity of the international fusion materials irradiation facility (IFMIF/EVEDA)

    International Nuclear Information System (INIS)

    Wakai, Eiichi; Kondo, Hiroo; Sugimoto, Masayoshi; Ida, Mizuho; Kanemura, Takuji; Watanabe, Kazuyoshi; Fujishiro, Kouji; Edao, Yuuki; Niitsuma, Shigeto; Kimura, Haruyuki; Fukada, Satoshi; Hiromoto, Tetsushi; Shigeharu, Satoshi; Yagi, Jyuro; Furukawa, Tomohiro; Hirakawa, Yasushi; Suzuki, Akihiro; Terai, Takayuki; Horiike, Hiroshi; Hoashi, Eiji; Suzuki, Sachiko; Yamaoka, Nobuo; Serizawa, Hisashi; Kawahito, Yosuke; Tsuji, Yoshiyuki; Furuya, Kazuyuki; Takeo, Fumio

    2012-01-01

    Engineering validation and engineering design activity (EVEDA) for the international fusion materials irradiation facility (IFMIF) has been conducted since 2007. Research and development of the Lithium target facility is an important part of this activity. We constructed a world largest liquid Lithium test loop with a capacity of 5000 L in 2010 and successfully completed the first stage validation tests (functional tests of components and Lithium flow test (flow velocity 15 m/s at the target). In the present article, recent results of the EVEDA activity for the Lithium target facility and related technologies on liquid Lithium are reviewed. (author)

  3. IFMIF – Layout and arrangement of cells according to requirements of technical logistics, reliability and remote handling

    International Nuclear Information System (INIS)

    Mittwollen, Martin; Eilert, Dirk; Kubaschewski, Martin; Madzharov, Vladimir; Tian Kuo

    2012-01-01

    Highlights: ► In a first approach, layout and arrangement of the cells followed a predetermined plant layout. ► Disadvantages in technical logistics, reliability and remote handling have been detected. ► Deliberation with project teams opened space for improvements. ► Layout and arrangement of cells have been improved by simplification of design. ► Speed and reliability have been increased significantly. - Abstract: The International Fusion Material Irradiation Facility (IFMIF) is designed to study and qualify structural and functional materials which shall be used in future fusion nuclear power plants. During the current engineering validation and engineering design activities (EVEDA) phase the development of e.g. an optimized layout and arrangement of the cells (Access Cell, Test Cell, and Test Module Handling Cells) is of major interest. After defining different functions for the individual cells like e.g. large scale/fine scale disassembling of test modules a first layout has been developed. This design followed requirements like having a minimum of carrier changes to avoid sources of failures. On the other hand it has had to be a compact arrangement of cells due to restrictions from plant layout. A row of changes of transfer direction, and different crane systems were the consequence. Constructive discussion with project team results in the statement, that for reasons of being reliable and fast, layout and arrangement of cells goes first, plant layout then will follow. The chance for big improvements was taken and the result was a simplified design with strong reduced number of functional elements, and increased reliability and speed.

  4. Mini-channel flow experiments and CFD validation analyses with the IFMIF Thermo- Hydraulic Experimental facility (ITHEX)

    International Nuclear Information System (INIS)

    Arbeiter, F.; Heinzel, V.; Leichtle, D.; Stratmanns, E.; Gordeev, S.

    2006-01-01

    The design of the IFMIF High Flux Test Module (HFTM) is based on the predictions for the heat transfer in narrow channels conducting helium flow of 50 o C inlet temperature at 0.3 MPa. The emerging helium flow conditions are in the transition regime of laminar to turbulent flow. The rectangular cooling channels are too short for the full development of the coolant flow. Relaminarization along the cooling passage is expected. At the shorter sides of the channels secondary flow occurs, which may have an impact on the temperature field inside the irradiation specimen's stack. As those conditions are not covered by available experimental data, the dedicated gas loop ITHEX has been constructed to operate up to a pressure of 0.42 MPa and temperatures of 200 o C. It's objective is to conduct experiments for the validation of the STAR-CD CFD code used for the design of the HFTM. As a first stage, two annular test-sections with hydraulic diameter of 1.2 mm have been used, where the experiments have been varied with respect to gas species (N 2 , He), inlet pressure, dimensionless heating span and Reynolds number encompassing the range of operational parameters of the HFTM. Local friction factors and Nusselt numbers have been obtained giving evidence that the transition regime will extend to Reynolds 10,000. For heating rates comparable to the HFTM filled with RAFM steels, local heat transfer coefficients are in consistence with the measured friction data. To validate local velocity profiles the ITHEX facility was further equipped with a flat rectangular test-section and a Laser Doppler Anemometry (LDA) system. An appropriate optical system has been developed and tested for the tiny observation volume of 40 μm diameter. Velocity profiles as induced by the transition of a wide inlet plenum to the flat mini-channels have been measured. Whereas the CFD models were able to reproduce the patterns far away from the nozzle, they show some disagreement for the conditions at the

  5. Adjoint sensitivity analysis procedure of Markov chains with applications on reliability of IFMIF accelerator-system facilities

    Energy Technology Data Exchange (ETDEWEB)

    Balan, I.

    2005-05-01

    This work presents the implementation of the Adjoint Sensitivity Analysis Procedure (ASAP) for the Continuous Time, Discrete Space Markov chains (CTMC), as an alternative to the other computational expensive methods. In order to develop this procedure as an end product in reliability studies, the reliability of the physical systems is analyzed using a coupled Fault-Tree - Markov chain technique, i.e. the abstraction of the physical system is performed using as the high level interface the Fault-Tree and afterwards this one is automatically converted into a Markov chain. The resulting differential equations based on the Markov chain model are solved in order to evaluate the system reliability. Further sensitivity analyses using ASAP applied to CTMC equations are performed to study the influence of uncertainties in input data to the reliability measures and to get the confidence in the final reliability results. The methods to generate the Markov chain and the ASAP for the Markov chain equations have been implemented into the new computer code system QUEFT/MARKOMAGS/MCADJSEN for reliability and sensitivity analysis of physical systems. The validation of this code system has been carried out by using simple problems for which analytical solutions can be obtained. Typical sensitivity results show that the numerical solution using ASAP is robust, stable and accurate. The method and the code system developed during this work can be used further as an efficient and flexible tool to evaluate the sensitivities of reliability measures for any physical system analyzed using the Markov chain. Reliability and sensitivity analyses using these methods have been performed during this work for the IFMIF Accelerator System Facilities. The reliability studies using Markov chain have been concentrated around the availability of the main subsystems of this complex physical system for a typical mission time. The sensitivity studies for two typical responses using ASAP have been

  6. Effect of activation cross section uncertainties in the assessment of primary damage for MFE/IFE low-activation steels irradiated in IFMIF

    International Nuclear Information System (INIS)

    Cabellos, O.; Sanz, J.; Garcia-Herranz, N.; Otero, B.

    2009-01-01

    The present study is mainly aimed to provide the primary damage (displacements per atom, generation of solid transmutants and gas production rates) of structural materials irradiated in the high and medium flux test modules of the International Fusion Materials Irradiation Facility (IFMIF). We have investigated if the change of the composition during the irradiation time has effect on the prediction of the atomic displacements. The effect of the activation cross section uncertainties in the assessment of both solid transmutants and hydrogen and helium production is also analyzed. The results are provided element-by-element, so that the primary damage of any material irradiated in such neutron environments can be easily assessed; in this paper, we have predicted the primary damage of the low activation steel Eurofer.

  7. Effect of activation cross section uncertainties in the assessment of primary damage for MFE/IFE low-activation steels irradiated in IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Cabellos, O. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (UPM), C/Jose Gutierrez Abascal, n2, 28006 Madrid (Spain); Dept. de Ingenieria Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain)], E-mail: cabellos@din.upm.es; Sanz, J. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (UPM), C/Jose Gutierrez Abascal, n2, 28006 Madrid (Spain); Dept. de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, 28045 Madrid (Spain); Garcia-Herranz, N. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (UPM), C/Jose Gutierrez Abascal, n2, 28006 Madrid (Spain); Dept. de Ingenieria Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain); Otero, B. [Dept. de Ingenieria Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain)

    2009-04-30

    The present study is mainly aimed to provide the primary damage (displacements per atom, generation of solid transmutants and gas production rates) of structural materials irradiated in the high and medium flux test modules of the International Fusion Materials Irradiation Facility (IFMIF). We have investigated if the change of the composition during the irradiation time has effect on the prediction of the atomic displacements. The effect of the activation cross section uncertainties in the assessment of both solid transmutants and hydrogen and helium production is also analyzed. The results are provided element-by-element, so that the primary damage of any material irradiated in such neutron environments can be easily assessed; in this paper, we have predicted the primary damage of the low activation steel Eurofer.

  8. Assessment of the structural integrity of a prototypical instrumented IFMIF high flux test module rig by fully 3D X-ray microtomography

    Energy Technology Data Exchange (ETDEWEB)

    Tiseanu, Ion [National Institute for Laser, Plasma and Radiation Physics, Plasma Physics and Nuclear Fusion Laboratory NILPRP, P.O. Box MG-36, R-77125 Bucharest-Magurele (Romania)], E-mail: tiseanu@infim.ro; Simon, Martin [Hans Waelischmiller GmbH (HWM), Schiessstattweg 16, D-88677 Markdorf (Germany); Craciunescu, Teddy; Mandache, Bogdan N. [National Institute for Laser, Plasma and Radiation Physics, Plasma Physics and Nuclear Fusion Laboratory NILPRP, P.O. Box MG-36, R-77125 Bucharest-Magurele (Romania); Heinzel, Volker; Stratmanns, Erwin; Simakov, Stanislaw P.; Leichtle, Dieter [Forschungszentrum Karlsruhe (FZK), Institut fuer Reaktorsicherheit IRS, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2007-10-15

    An inspection procedure to assess the mechanical integrity of the International Fusion Materials Irradiation Facility (IFMIF) capsules and rigs during the irradiation campaign is necessary. Due to its penetration ability and contrast mechanism, the X-ray microtomography is the only known tool that could meet these requirements. In the high flux test module (HFTM) of IFMIF miniaturized specimens are densely packed in capsules. The capsules, which wear electric heaters and thermocouples, are housed in rigs. To assure a well-defined thermal contact the heater wires have to be attached to the capsules by brazing them into grooves. The examination of the quality of the braze material layer is of crucial interest in order to assure the best heat coupling of the heater wires to the capsule. A high density of the heaters is necessary to maintain the required temperature and, in addition NaK filling of narrow channels is employed for improving the 3D-heat transfer between the irradiation specimens and the capsule wall. Fully 3D tomographic inspections of a prototypical HFTM instrumented capsule, developed and manufactured at FZK, were conducted. In order to identify the optimum irradiation parameters and scanning configuration we carried out a comparative NDT analysis on two microtomography facilities: a compact, high magnification installation at NILPRP and a high-end industrial tomography facility with higher X-ray energy and intensity at HWM. At optimum inspection parameters of a directional microfocus X-ray source (U = 220 kV and I = 300 {mu}A) the geometry resolution was about 30 microns for characteristic dimension of the sample of 50 mm. Voids of 30 microns diameter and cracks of about 20 microns width can be detected. The absolute error of geometrical measurements is sufficient for the assessment of the structural integrity of the irradiation capsule and for the geometry description within the thermal-hydraulic modeling. The space resolution and the overall

  9. Assessment of the structural integrity of a prototypical instrumented IFMIF high flux test module rig by fully 3D X-ray microtomography

    International Nuclear Information System (INIS)

    Tiseanu, Ion; Simon, Martin; Craciunescu, Teddy; Mandache, Bogdan N.; Heinzel, Volker; Stratmanns, Erwin; Simakov, Stanislaw P.; Leichtle, Dieter

    2007-01-01

    An inspection procedure to assess the mechanical integrity of the International Fusion Materials Irradiation Facility (IFMIF) capsules and rigs during the irradiation campaign is necessary. Due to its penetration ability and contrast mechanism, the X-ray microtomography is the only known tool that could meet these requirements. In the high flux test module (HFTM) of IFMIF miniaturized specimens are densely packed in capsules. The capsules, which wear electric heaters and thermocouples, are housed in rigs. To assure a well-defined thermal contact the heater wires have to be attached to the capsules by brazing them into grooves. The examination of the quality of the braze material layer is of crucial interest in order to assure the best heat coupling of the heater wires to the capsule. A high density of the heaters is necessary to maintain the required temperature and, in addition NaK filling of narrow channels is employed for improving the 3D-heat transfer between the irradiation specimens and the capsule wall. Fully 3D tomographic inspections of a prototypical HFTM instrumented capsule, developed and manufactured at FZK, were conducted. In order to identify the optimum irradiation parameters and scanning configuration we carried out a comparative NDT analysis on two microtomography facilities: a compact, high magnification installation at NILPRP and a high-end industrial tomography facility with higher X-ray energy and intensity at HWM. At optimum inspection parameters of a directional microfocus X-ray source (U = 220 kV and I = 300 μA) the geometry resolution was about 30 microns for characteristic dimension of the sample of 50 mm. Voids of 30 microns diameter and cracks of about 20 microns width can be detected. The absolute error of geometrical measurements is sufficient for the assessment of the structural integrity of the irradiation capsule and for the geometry description within the thermal-hydraulic modeling. The space resolution and the overall

  10. Assessment of the Structural Integrity of a Prototypical Instrumented IFMIF High Flux Test Module Rig by Fully 3D X-Ray Microtomography

    International Nuclear Information System (INIS)

    Tiseanu, I.; Craciunescu, T.; Mandache, B.N.; Simon, M.; Heinzel, V.; Stratmanns, E.; Simakov, S.P.; Leichtle, D.

    2006-01-01

    An inspection procedure to asses the mechanical integrity of IFMIF (International Fusion Materials Irradiation Facility) capsules and rigs during the irradiation campaign is necessary. Due to its penetration ability and contrast mechanism, the X-ray micro-tomography is the only known tool that could meet these requirements. In the High Flux Test Module (HFTM) of IFMIF miniaturized specimens are densely packed in capsules. The capsules which wear electric heaters and thermocouples are housed in rigs. To assure a well defined thermal contact the heater wires have to be attached to the capsules by brazing them into grooves. The examination of the quality of the braze material layer is of crucial interest in order to assure the best heat coupling of the heater wires to the capsule. A high density of the heaters is necessary to maintain the required temperature and, in addition NaK filling of narrow channels is employed for improving the 3D-heat transfer between the irradiation specimens and the capsule wall. Fully 3D tomographic inspections of a prototypical HFTM instrumented capsule, developed and manufactures at FZK, were conducted. In order to identify the optimum irradiation parameters and scanning configuration we carried out a comparative NDT analysis on two micro-tomography facilities, our compact, high magnification installation at NILPRP and two high-end industrial tomography facilities with higher X-ray energy and intensity at HWM. At optimum inspection parameters of a microfocus X-ray source (U=220 kV and I=300 μA) the geometry resolution was about 30-50 microns for characteristic dimension of the sample of 50 mm. Voids of 30 microns diameter and cracks of about 20 microns width can be detected. The absolute error of geometrical measurements should be sufficient for the assessment of the structural integrity of the irradiation capsule and for the geometry description within the thermal-hydraulic modeling. Space resolution could be further improved if one

  11. Large-Eddy Simulation of Turbulent Flow and Heat Transfer in a Mildly Expanded Channel of IFMIF High Flux Test Module

    International Nuclear Information System (INIS)

    Shinji Ebara; Takehiko Yokomine; Akihiko Shimizu

    2006-01-01

    During irradiation test periods in the International Fusion Material Irradiation Facility (IFMIF), irradiated materials must be maintained at constant temperatures because irradiation characteristics of materials have a large dependency on temperature. In the high flux test module of the IFMIF, required performances for temperature control using gas-cooling and heater-heating are especially stringent because available space for temperature control is remarkably restricted due to very small irradiation volume of about 0.5 l. We proposed an alternative design of the test module with advantages of temperature monitoring and temperature uniformity in specimens. This design employs a rectangular duct as the vessel to pack capsules housing specimens compactly into the small irradiation volume. In the vessel the coolant flows between the capsules and vessel wall. In the basic design, both thickness of a vessel wall and a width of cooling channel are considered as 1.0 mm. Since inside the vessel gaseous helium of several atmospheric pressure flows as a coolant and a low vacuum environment is kept outside the vessel for safety requirements and thermal stress is foreseen to appear due to nuclear heating of the vessel itself, the vessel wall is considered to deform readily and this leads expansion of the cooling channels. It is also considered that a slight expansion of the vessel can have severe influence on the cooling performance due to the initial narrow channel width of 1.0 mm. Therefore, it is necessary to estimate cooling performances for the coolant flowing in the deformed channel. We conduct a finite element analysis of turbulent heat transfer in a mildly expanded channel using large-eddy simulation in this study. In a numerical system, fluid is enclosed by three-dimensionally expanded vessel wall and flat capsule wall, and flows into the system with a fully developed velocity profile. In this study, we focus not only on the cooling performances but also on change in

  12. Structural Response of Submerged Air-Backed Plates by Experimental and Numerical Analyses

    Directory of Open Access Journals (Sweden)

    Lloyd Hammond

    2000-01-01

    Full Text Available This paper presents the results of a series of small-scale underwater shock experiments that measured the structural responses of submerged, fully clamped, air-backed, steel plates to a range of high explosive charge sizes. The experimental results were subsequently used to validate a series of simulations using the coupled LS-DYNA/USA finite element/boundary element codes. The modelling exercise was complicated by a significant amount of local cavitation occurring in the fluid adjacent to the plate and difficulties in modelling the boundary conditions of the test plates. The finite element model results satisfactorily predicted the displacement-time history of the plate over a range of shock loadings although a less satisfactory correlation was achieved for the peak velocities. It is expected that the predictive capability of the finite element model will be significantly improved once hydrostatic initialisation can be fully utilised with the LS-DYNA/USA software.

  13. Activation of Eurofer in an IFMIF-like neutron field

    Czech Academy of Sciences Publication Activity Database

    Bém, Pavel; Burjan, Václav; Götz, Miloslav; Honusek, Milan; Fischer, V.; Kroha, Václav; von Mollendorff, U.; Novák, Jan; Simakov, SP.; Šimečková, Eva

    2005-01-01

    Roč. 75, č. 9 (2005), s. 829-833 ISSN 0920-3796 R&D Projects: GA AV ČR KSK1048102 Grant - others:Evropská unie(BE) EFDA TW3-TTMN-002/D5a Institutional research plan: CEZ:AV0Z1048901 Keywords : Eurofer-97 * steel * radio-nuclides Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 0.981, year: 2005

  14. Relationship Between Final Performance and Block Times with the Traditional and the New Starting Platforms with A Back Plate in International Swimming Championship 50-M and 100-M Freestyle Events

    Directory of Open Access Journals (Sweden)

    Antonio Garcia-Hermoso

    2013-12-01

    Full Text Available The purpose of this study was to investigate the association between block time and final performance for each sex in 50-m and 100-m individual freestyle, distinguishing between classification (1st to 3rd, 4th to 8th, 9th to 16th and type of starting platform (old and new in international competitions. Twenty-six international competitions covering a 13-year period (2000-2012 were analysed retrospectively. The data corresponded to a total of 1657 swimmers’ competition histories. A two-way ANOVA (sex x classification was performed for each event and starting platform with the Bonferroni post-hoc test, and another two-way ANOVA for sex and starting platform (sex x starting platform. Pearson’s simple correlation coefficient was used to determine correlations between the block time and the final performance. Finally, a simple linear regression analysis was done between the final time and the block time for each sex and platform. The men had shorter starting block times than the women in both events and from both platforms. For 50-m event, medalists had shorter block times than semi- finalists with the old starting platforms. Block times were directly related to performance with the old starting platforms. With the new starting platforms, however, the relationship was inverse, notably in the women’s 50-m event. The block time was related for final performance in the men’s 50- m event with the old starting platform, but with the new platform it was critical only for the women’s 50-m event.

  15. Neutron-induced damage evolution under Beam Raster Scanner conditions for IFMIF

    International Nuclear Information System (INIS)

    Mota, Fernando; Ortiz, Christophe J.; Ibarra, Angel; Vila, Rafael

    2011-01-01

    The formation and evolution of defects in materials irradiated with a homogeneous neutron source and with the Beam Raster Scanner (BRS) solution was investigated. The intensity neutron source fluctuations inherent to the BRS system were determined using the neutron transport McDeLicious code. Defects generated during irradiation were calculated using the binary collision approximation MARLOWE code, using the primary knock-on atom (PKA) energy spectrum resulting from neutron interactions with the material. In order to predict the evolution of defects during irradiation, a Rate Theory model based on ab initio parameters was developed. Our model accounts for the migration of mobile defects, the formation of clusters and their recombination. As an example, we investigated defect evolution in Fe irradiated at room temperature in both beam configurations. Simulation results clearly indicate that the defect evolution expected in the BRS configuration is nearly the same as the one expected in a homogeneous irradiation system.

  16. The Control Architecture of Large Scientific Facilities: ITER and LHC lessons for IFMIF

    CERN Document Server

    Marqueta Barbero, A; Nishiyama, K; Ibarra, A; Vergara-Fernandez, A; Wallander, A; Zerlauth, M

    2014-01-01

    The development of an intense source of neutrons with the spectrum of deuterium-tritium (DT) fusion reactions is indispensable to qualify suitable materials for the blanket of the nuclear vessel in fusion power plants.

  17. Preliminary assessment of the activation of the IFMIF accelerator structure by deuterons and neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Itacil C. [Argonne National Lab., IL (United States); Bruhwiler, David L. [Northrop Grumman Corp., Princeton, NJ (United States). Advanced Systems and Technology

    1997-12-01

    This paper presents a preliminary analysis of the IFMF (International Fusion Materials Irradiation Facility) accelerator structure activation by deuterons and neutrons. The main objective of this study is to identify the source terms and to quantify the radioactivity levels at different positions in the accelerator vault. The MCNP code is used to perform radiation transport analysis, the RACC activation code is used for neutron activation analysis, and the cross section library of the LAHET code is used to generate the cross section for the deuteron interaction with the inside surfaces of the accelerator. (author). 10 refs., 5 figs.

  18. Close to Kill: Vestigial Technologies and Combat Arms

    Science.gov (United States)

    2012-06-01

    Armor (Philadelphia, PA: Chilton Books , 1962), 77. 35 For detailed accounts on individuals, to include James D. Bulloch and Lieutenant James H. North...third-order effects when the transportation routes were needed later on. plants were secondhand ...to access the firearms in timely manner. See Martin J. Brayley, Bayonets: An Illustrated History (Iola, WI: KP Books , 2004). Evans, The Bayonet

  19. Measurement of deuteron-induced activation cross section for IFMIF accelerator structural materials in 22-40 MeV region

    International Nuclear Information System (INIS)

    Nakao, Makoto; Hori, Jun-ichi; Ochiai, Kentaro; Sato, Satoshi; Yamauchi, Michinori; Nishitani, Takeo; Ishioka, Noriko S.

    2004-01-01

    The activation cross-sections for the deuteron-induced reactions have been obtained for Al, Cu and W in 22-40 MeV regions and compared with previous experimental ones and the data in ACSELAM library. For 27 Mg, ACSELAM were smaller than the present result by a factor of 1.3-2.0. For 24 Na, ACSELAM resembled experimental values in shape but were lower than these by about 1 order. For 61 Cu, 64 Cu, 62 Zn and 63 Zn, the present results resembled other experimental data and ACSELAM in shape. In the case of 61 Cu and 62 Zn, ACSELAM became higher than the present results by a factor of 2-4. In the case of 64 Cu and 63 Zn, ACSELAM and the present results were in agreement within 40%. For 181 Re, 182m+g Re and 183 Re, the present data and the data in ACSELAM were about same shapes and in agreement within 30%. For 184m+g Re, 186 Re and 187 W, the data in ACSELAM were different from the present data about 1.5-7 times. (author)

  20. International fusion materials irradiation facility and neutronic calculations for its test modules

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.

    1997-01-01

    The International Fusion Material Irradiation Facility (IFMIF) is a projected high intensity neutron source for material testing. Neutron transport calculations for the IFMIF project are performed for variety of here explained reasons. The results of MCNP neutronic calculations for IFMIF test modules with NaK and He cooled high flux test cells are presented in this paper. (author). 3 refs., 2 figs., 3 tabs

  1. IS BAJONETTE UITGEDIEN?

    OpenAIRE

    W. Otto

    2012-01-01

    It is sometimes difficult for people to break with outmoded traditions, and this is particularly noticeable in the military world, where today the most modern and technological equipment is carried into battle by soldiers also equipped with a weapon of quite historic origin, the bayonet. During the period of the single-shot infantry weapon, the bayonet formed an essential part of the soldier's equipment, being useful in mounting or repelling assaults at close-quarters when it was impossible t...

  2. British Intelligence Operations as They Relate to Britain’s Defeat at Yorktown, 1781

    Science.gov (United States)

    2010-06-11

    121Matthew H. Spring, With Zeal and With Bayonets Only: The British Army on Campaign in North America, 1775-1783 ( Norman , OK: University of Oklahoma...humorously complained to Lieutenant Colonel Francis Lord Rawdon, “All my accounts about [Major General] Smallwood agree with yours, but mine are: ‘I...Matthew H. With Zeal and With Bayonets Only: The British Army on Campaign in North America, 1775-1783. Norman , OK: University of Oklahoma Press, 2008

  3. Vegetation Resources of Rocky Mountain Arsenal, Adams County, Colorado

    Science.gov (United States)

    1989-10-01

    SUCCULENTS Opuntia polyacantha Plains Prickly Pear Cactaceae Table 4 .(cont’d.) Scientific Name Common Name Family Name Yucca glauca Spanish Bayonet Agavaceae...Lycium halimitolium Matrim~ony Bush Solanceae Salix exigua Coyote Willow Salicaceae CACTI AND SUCCULENTS Coryphantha vivipara Ball Cactus Cactaceae ...Qpuntia compressa Prickly Pear Cactus Cactaceae Opuntia polyacantha Plains Prickly Pear Cactaceae Yucca glauca Spanish Bayonet Agavaceae cq Nl Nl Co~ V c0

  4. Results from the CDE phase activity on neutron dosimetry for the international fusion materials irradiation facility test cell

    CERN Document Server

    Esposito, B; Maruccia, G; Petrizzi, L; Bignon, G; Blandin, C; Chauffriat, S; Lebrun, A; Recroix, H; Trapp, J P; Kaschuck, Y

    2000-01-01

    The international fusion materials irradiation facility (IFMIF) project deals with the study of an accelerator-based, deuterium-lithium source, producing high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials for fusion energy reactors. IFMIF would also provide calibration and validation of data from fission reactor and other accelerator based irradiation tests. This paper describes the activity on neutron/gamma dosimetry (necessary for the characterization of the specimens' irradiation) performed in the frame of the IFMIF conceptual design evaluation (CDE) neutronics tasks. During the previous phase (conceptual design activity (CDA)) the multifoil activation method was proposed for the measurement of the neutron fluence and spectrum and a set of suitable foils was defined. The cross section variances and covariances of this set of foils have now been used for tests on the sensitivity of the IFMIF neutron spectrum determination to cross section uncertainties...

  5. CMC vane assembly apparatus and method

    Science.gov (United States)

    Schiavo, Anthony L; Gonzalez, Malberto F; Huang, Kuangwei; Radonovich, David C

    2012-10-23

    A metal vane core or strut (64) is formed integrally with an outer backing plate (40). An inner backing plate (38) is formed separately. A spring (74) with holes (75) is installed in a peripheral spring chamber (76) on the strut. Inner and outer CMC shroud covers (46, 48) are formed, cured, then attached to facing surfaces of the inner and outer backing plates (38, 40). A CMC vane airfoil (22) is formed, cured, and slid over the strut (64). The spring (74) urges continuous contact between the strut (64) and airfoil (66), eliminating vibrations while allowing differential expansion. The inner end (88) of the strut is fastened to the inner backing plate (38). A cooling channel (68) in the strut is connected by holes (69) along the leading edge of the strut to peripheral cooling paths (70, 71) around the strut. Coolant flows through and around the strut, including through the spring holes.

  6. Stability of fluid flow through deformable tubes and channels: An ...

    Indian Academy of Sciences (India)

    This was immediately followed by the theoretical studies of ... both phenomenological spring-backed plate models and continuum linear viscoelas .... the stability of flow at high Reynolds number in section 4, where the inviscid instability mecha ...

  7. Designing an Innovative Composite Armor System for Affordable Ballistic Protection

    National Research Council Canada - National Science Library

    Ma, Zheng-Dong; Wang, Hui; Cui, Yushun; Rose, Douglas; Socks, Adria; Ostberg, Donald

    2006-01-01

    .... This paper focuses on the frontal armor plate and back plate design problems with demonstration examples, including both results of the virtual prototyping and ballistic testing for proof-of-concept...

  8. Integrated boiler, superheater, and decomposer for sulfuric acid decomposition

    Science.gov (United States)

    Moore, Robert [Edgewood, NM; Pickard, Paul S [Albuquerque, NM; Parma, Jr., Edward J.; Vernon, Milton E [Albuquerque, NM; Gelbard, Fred [Albuquerque, NM; Lenard, Roger X [Edgewood, NM

    2010-01-12

    A method and apparatus, constructed of ceramics and other corrosion resistant materials, for decomposing sulfuric acid into sulfur dioxide, oxygen and water using an integrated boiler, superheater, and decomposer unit comprising a bayonet-type, dual-tube, counter-flow heat exchanger with a catalytic insert and a central baffle to increase recuperation efficiency.

  9. Evaluation and Testing of the Suitability of a Coal-Based Jet Fuel

    Science.gov (United States)

    2008-06-01

    with a total wattage of 7980 watts. Each oven section has two K type thermocouples per zone with Inconel sheathed spring loaded bayonet type mounts...also exceeded the thermal stability goals (525°F bulk and 625 °F WWT) for the JP-8+225 fuel program. Tests were conducted on a JP-8 fuel to compare

  10. USA sõdurid tulid Eestisse langevarjuhüppega / fotod: Dmitri Kotjuh

    Index Scriptorium Estoniae

    2016-01-01

    Sada Ameerika Ühendriikide 173. õhudessantbrigaadi õhudessantväelast sooritasid langevarjuhüppe Järvamaa Nurmsi murulennuväljale. Ameerika Ühendriikide sõjaväelased teevad kaasa õppuses Bayonet Strike, mis viiakse läbi kolmes Balti riigis ja Poolas

  11. The Department of Defense Press Pool: Did in Work in Panama?

    Science.gov (United States)

    1990-01-01

    who would physically attack a paper and its printer that did not reflect their opinion (Tebbel, 5). Politicians rarely brought the papers to task...Rico single-handedly and James Creelman , a reporter for Hearst, leading a bayonet charge outside Santiago. Military and press relations were good. The

  12. Closure for spent-fuel transport and storage containers

    International Nuclear Information System (INIS)

    Ahner, S.; Knackstedt, H.G.; Srostlik, P.

    1980-01-01

    The container has a transport closure and a shielding closure. This shielding closure consists of two pieces (double closure system), which can be fartened to one another like a bayonet fixing. A central motion of rotation is enough to open the closure. It can be done remote-controlled as well as manually. (DG) [de

  13. Staged deployment of the International Fusion Materials Irradiation Facility

    International Nuclear Information System (INIS)

    Takeuchi, H.; Sugimoto, M.; Nakamura, H.

    2001-01-01

    The International Fusion Materials Irradiation Facility (IFMIF) employs an accelerator based D-Li intense neutron source as defined in the 1995-96 Conceptual Design Activity (CDA) study. In 1999, IEA mandated a review of the CDA IFMIF design for cost reduction without change to its original mission. This objective was accomplished by eliminating the previously assumed possibility of potential upgrade of IFMIF beyond the user requirements. The total estimated cost was reduced from $797.2 M to $487.8 M. An option of deployment in 3 stages was also examined to reduce the initial investment and annual expenditures during construction. In this scenario, full performance is achieved gradually with each interim stage as follows. 1st Stage: 20% operation for material selection for ITER breeding blanket, 2nd Stage: 50% operation to demonstrate materials performance of a reference alloy for DEMO, 3rd Stage: full performance operation ( 2MW/m 2 at 500cm 3 ) to obtain engineering data for potential DEMO materials under irradiation up to 100-200 dpa. In summary, the new, reduced cost IFMIF design and staged deployment still satisfies the original mission. The estimated cost of the 1st Stage facility is only $303.6 M making it financially much more attractive. Currently, IFMIF Key Element Technology Phase (KEP) is underway to reduce the key technology risk factors. (author)

  14. How to improve the irradiation conditions for the International Fusion Materials Irradiation Facility

    CERN Document Server

    Daum, E

    2000-01-01

    The accelerator-based intense D-Li neutron source International Fusion Materials Irradiation Facility (IFMIF) provides very suitable irradiation conditions for fusion materials development with the attractive option of accelerated irradiations. Investigations show that a neutron moderator made of tungsten and placed in the IFMIF test cell can further improve the irradiation conditions. The moderator softens the IFMIF neutron spectrum by enhancing the fraction of low energy neutrons. For displacement damage, the ratio of point defects to cascades is more DEMO relevant and for tritium production in Li-based breeding ceramic materials it leads to a preferred production via the sup 6 Li(n,t) sup 4 He channel as it occurs in a DEMO breeding blanket.

  15. Results from the CDE phase activity on neutron dosimetry for the international fusion materials irradiation facility test cell

    Energy Technology Data Exchange (ETDEWEB)

    Esposito, B. E-mail: esposito@frascati.enea.it; Bertalot, L.; Maruccia, G.; Petrizzi, L.; Bignan, G.; Blandin, C.; Chauffriat, S.; Lebrun, A.; Recroix, H.; Trapp, J.P.; Kaschuck, Y

    2000-11-01

    The international fusion materials irradiation facility (IFMIF) project deals with the study of an accelerator-based, deuterium-lithium source, producing high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials for fusion energy reactors. IFMIF would also provide calibration and validation of data from fission reactor and other accelerator based irradiation tests. This paper describes the activity on neutron/gamma dosimetry (necessary for the characterization of the specimens' irradiation) performed in the frame of the IFMIF conceptual design evaluation (CDE) neutronics tasks. During the previous phase (conceptual design activity (CDA)) the multifoil activation method was proposed for the measurement of the neutron fluence and spectrum and a set of suitable foils was defined. The cross section variances and covariances of this set of foils have now been used for tests on the sensitivity of the IFMIF neutron spectrum determination to cross section uncertainties. The analysis has been carried out using the LSL-M2 code, which optimizes the neutron spectrum by means of a least-squares technique taking into account the variance and covariance files. In the second part of the activity, the possibility of extending to IFMIF the use of existing on-line in-core neutron/gamma monitors (to be located at several positions inside the IFMIF test cell for beam control, safety and diagnostic purposes) has been studied. A feasibility analysis of the modifications required to adapt sub-miniature fission chambers (recently developed by CEA-Cadarache) to the high flux test module of the test cell has been carried out. The verification of this application pertinence and a gross definition of the in-core detector characteristics are described. The option of using self-powered neutron detectors (SPNDs) is also discussed.

  16. Advanced computational tools and methods for nuclear analyses of fusion technology systems

    International Nuclear Information System (INIS)

    Fischer, U.; Chen, Y.; Pereslavtsev, P.; Simakov, S.P.; Tsige-Tamirat, H.; Loughlin, M.; Perel, R.L.; Petrizzi, L.; Tautges, T.J.; Wilson, P.P.H.

    2005-01-01

    An overview is presented of advanced computational tools and methods developed recently for nuclear analyses of Fusion Technology systems such as the experimental device ITER ('International Thermonuclear Experimental Reactor') and the intense neutron source IFMIF ('International Fusion Material Irradiation Facility'). These include Monte Carlo based computational schemes for the calculation of three-dimensional shut-down dose rate distributions, methods, codes and interfaces for the use of CAD geometry models in Monte Carlo transport calculations, algorithms for Monte Carlo based sensitivity/uncertainty calculations, as well as computational techniques and data for IFMIF neutronics and activation calculations. (author)

  17. Comparison of material irradiation conditions for fusion, spallation, stripping and fission neutron sources

    International Nuclear Information System (INIS)

    Vladimirov, P.; Moeslang, A.

    2004-01-01

    Selection and development of materials capable of sustaining irradiation conditions expected for a future fusion power reactor remain a big challenge for material scientists. Design of other nuclear facilities either in support of the fusion materials testing program or for other scientific purposes presents a similar problem of irradiation resistant material development. The present study is devoted to an evaluation of the irradiation conditions for IFMIF, ESS, XADS, DEMO and typical fission reactors to provide a basis for comparison of the data obtained for different material investigation programs. The results obtained confirm that no facility, except IFMIF, could fit all user requirements imposed for a facility for simulation of the fusion irradiation conditions

  18. Intense neutron irradiation facility for fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji; Oyama, Yukio; Kato, Yoshio; Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Technical R and D of d-Li stripping type neutron irradiation facilities for development of fusion reactor materials was carried out in Fusion Materials Irradiation Test Facility (FMIT) project and Energy Selective Neutron Irradiation Test Facility (ESNIT) program. Conceptual design activity (CDA) of International Fusion Materials Irradiation Facility (IFMIF), of which concept is an advanced version of FMIT and ESNIT concepts, are being performed. Progress of users` requirements and characteristics of irradiation fields in such neutron irradiation facilities, and outline of baseline conceptual design of IFMIF were described. (author)

  19. The Development of a High-Power, Low-Frequency Underwater Acoustic Source for Use in a Deep-Towed Geophysical Array Section

    Science.gov (United States)

    1982-11-15

    Metca) for each ring 6 Aluminum orirlce end ring 15 Rubber bumpers 7 Back plate inner boot ring 16 Castor oil rill-fluid I ’)-ring %eals I ’ Nikel ...electrode .ab% with hook-up wire -ig. 6 - Sectional v-icw orf ISRI) typc (662 tran.dhcer 9 YOUNG, TIMS. AND HENRIQUFZ 0.42 m. Approximately 85% of the...electrical leads for each ring are routed to individual bulkhead connectors on the trans- ducer back plate with high-voltage silicone-jacketed hookup wire

  20. On the Prediction of Hot Tearing in Al-to-Steel Welding by Friction Melt Bonding

    Science.gov (United States)

    Jimenez-Mena, N.; Jacques, P. J.; Drezet, J. M.; Simar, A.

    2018-04-01

    Aluminum alloy AA6061 was welded to dual-phase steel 980 (DP980) by the friction melt bonding (FMB) process. Hot tears have been suppressed by controlling the thermomechanical cycle. In particular, the welding speed and the thermal conductivity of the backing plate have been optimized. A finite-element thermomechanical model coupled with the Rappaz-Drezet-Gremaud (RDG) criterion has been used to explain these experimental observations. The hot tear susceptibility has been reduced with large thermal gradients and with the formation of a cellular microstructure. Both effects are favored by a backing plate made of a material with high thermal conductivity, such as copper.

  1. Synthesis, Characterization and Application of Functional Carbon Nano Materials

    Science.gov (United States)

    2014-05-05

    backing plate and this bond must be able to withstand a temperature of 220℃. Fig. 2.7c shows one example of a ZnO target bonded to a copper backing plate...44: 2155 (2006) [15] D.B. Hibbert, Introduction to Electrochemistry , Mackays of Chatham PLC Chatham, Kent (1993) [16] S.M. Sze, Semiconductor...Publications during Ph.D study 1. J. Chu, X.Y. Peng, K. Dasari, R. Palai, P. Feng, “The shift of optical band gap in W-doped ZnO with oxygen pressure and

  2. Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

    Science.gov (United States)

    Narcisi, V.; Giannetti, F.; Del Nevo, A.; Tarantino, M.; Caruso, G.

    2017-11-01

    In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

  3. The Pendulum of War and Politics

    Science.gov (United States)

    2013-05-16

    Americanization produced an increase in the magnitude of effort. As additional resources increased so did the media coverage, debates amongst the...Tactical Zone’s (CTZ) Pleiku Province.43 The battle damages assessment ( BDA ) of SILVER BAYONET, commonly known as the Battle of Ia Drang, included...objective by air assault.47 Operation CEDAR FALLS’ BDA : 750 enemy casualties, and 72 U.S. causalities, a 10 – 1 kill ratio. With such unbalanced BDA

  4. In Defense of Japan

    Science.gov (United States)

    2018-01-03

    potential addition of F-35B STOVL aircraft to their Izumo class helicopter destroyers is a good example. Japan’s conduct during World War II remain locked ...antiwar and antinuclear identity . But after sixty years, Japan’s neighbors still see bayoneted babies. Relations between Japan and its former victims...and the world order Robert Cooper defines Japan as the lone post-modern country surrounded by states firmly locked into an earlier age and that if

  5. Training Capability Data for Dismounted Soldier Training System

    Science.gov (United States)

    2015-06-01

    Virtual Squad Training System ( VSTS ). Like some of its predecessors, VSTS included a combination of man-wearable, tethered, and desktop interfaces...Simulator Bayonet w/Omni Directional Treadmill TRAC-WSMR Soldier Station Soldier Visualization Station V-IMTS SVS2-DI, DAGGERS, ASWETS VSTS Dismounted...Simulation, VSTS – Virtual Squad Training System 4 and microphone. The VSMM utilizes radio frequency identification (RFID) tags and hand sensors to

  6. Solid Propellant Subscale Burning Rate Analysis Methods for US and Selected NATO Facilities

    Science.gov (United States)

    2002-01-01

    impossibility of the center of a particle lying closer than its radius from a solid boundary, * Due to surface tension and sedimentation (tends to level...34 effect (for bottom cast or bayonet cast grains) may consist of sedimentation of larger particles against the walls during casting flow, with the...February 2000. 91 Ratti A., "Metodi di Riduzione Dati Balistici per i Boosters a Propellente Solido di Ariane-4 e di Ariane-5," M.Sc. Thesis in Aerospace

  7. The Activities of the European Consortium on Nuclear Data Development and Analysis for Fusion

    International Nuclear Information System (INIS)

    Fischer, U.; Avrigeanu, M.; Avrigeanu, V.; Cabellos, O.; Kodeli, I.; Koning, A.; Konobeyev, A.Yu.; Leeb, H.; Rochman, D.; Pereslavtsev, P.; Sauvan, P.; Sublet, J.-C.; Trkov, A.; Dupont, E.; Leichtle, D.; Izquierdo, J.

    2014-01-01

    This paper presents an overview of the activities of the European Consortium on Nuclear Data Development and Analysis for Fusion. The Consortium combines available European expertise to provide services for the generation, maintenance, and validation of nuclear data evaluations and data files relevant for ITER, IFMIF and DEMO, as well as codes and software tools required for related nuclear calculations

  8. The Activities of the European Consortium on Nuclear Data Development and Analysis for Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physic and Reactor Technology, 76344 Eggenstein-Leopoldshafen (Germany); Avrigeanu, M.; Avrigeanu, V. [Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), RO-077125 Magurele (Romania); Cabellos, O. [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain); Kodeli, I. [Jozef Stefan Institute (JSI), Jamova 39, 1000 Ljubljana (Slovenia); Koning, A. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 LE Petten (Netherlands); Konobeyev, A.Yu. [Karlsruhe Institute of Technology, Institute for Neutron Physic and Reactor Technology, 76344 Eggenstein-Leopoldshafen (Germany); Leeb, H. [Technische Universitaet Wien, Atominstitut, Wiedner Hauptstrasse 8–10, 1040 Wien (Austria); Rochman, D. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 LE Petten (Netherlands); Pereslavtsev, P. [Karlsruhe Institute of Technology, Institute for Neutron Physic and Reactor Technology, 76344 Eggenstein-Leopoldshafen (Germany); Sauvan, P. [Universidad Nacional de Educacion a Distancia, C. Juan del Rosal, 12, 28040 Madrid (Spain); Sublet, J.-C. [Euratom/CCFE Fusion Association, Culham Science Centre, OX14 3DB (United Kingdom); Trkov, A. [Jozef Stefan Institute (JSI), Jamova 39, 1000 Ljubljana (Slovenia); Dupont, E. [OECD Nuclear Energy Agency, Paris (France); Leichtle, D.; Izquierdo, J. [Fusion for Energy, Barcelona (Spain)

    2014-06-15

    This paper presents an overview of the activities of the European Consortium on Nuclear Data Development and Analysis for Fusion. The Consortium combines available European expertise to provide services for the generation, maintenance, and validation of nuclear data evaluations and data files relevant for ITER, IFMIF and DEMO, as well as codes and software tools required for related nuclear calculations.

  9. International Fusion Materials Irradiation Facility conceptual design activity. Present status and perspective

    International Nuclear Information System (INIS)

    Kondo, Tatsuo; Noda, Kenji; Oyama, Yukio

    1998-01-01

    For developing the materials for nuclear fusion reactors, it is indispensable to study on the neutron irradiation behavior under fusion reactor conditions, but there is not any high energy neutron irradiation facility that can simulate fusion reactor conditions at present. Therefore, the investigation of the IFMIF was begun jointly by Japan, USA, Europe and Russia following the initiative of IEA. The conceptual design activities were completed in 1997. As to the background and the course, the present status of the research on heavy irradiation and the testing means for fusion materials, the requirement and the technical basis of high energy neutron irradiation, and the international joint design activities are reported. The materials for fusion reactors are exposed to the neutron irradiation with the energy spectra up to 14 MeV. The requirements from the users that the IFMIF should satisfy, the demand of the tests for the materials of prototype and demonstration fusion reactors and the evaluation of the neutron field characteristics of the IFMIF are discussed. As to the conceptual design of the IFMIF, the whole constitution, the operational mode, accelerator system and target system are described. (K.I.)

  10. Measurement of Li target thickness in the EVEDA Li 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; Yoshihashi, Sachiko; Horiike, Hiroshi [Osaka University, 2-1 Yamada-oka, Suita, Osaka 565-0871 (Japan); Wakai, Eiichi [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan)

    2015-10-15

    Highlights: • The objective is to validate stability of the IFMIF liquid Li target flowing at 15 m/s. • Design requirement of target thickness fluctuation is ±1 mm. • Mean and maximum wave amplitude are 0.26 and 1.46 mm, respectively. • Average thickness can be well predicted with developed analytical model. • Li target was adequately stable and satisfied design requirement. - Abstract: A high-speed (nominal: 15 m/s, range: 10–16 m/s) liquid lithium wall jet is planned to serve as the target for two 40 MeV and 125 mA deuteron beams in the International Fusion Materials Irradiation Facility (IFMIF). The design requirement of target thickness stability is 25 ± 1 mm under a vacuum of 10{sup −3} Pa. This paper presents the results of the target thickness measurement conducted in the EVEDA Li Test Loop under a wide range of conditions including the IFMIF condition (target speed of 10, 15, and 20 m/s; vacuum pressure of 10{sup −3} Pa; and Li temperature of 250 °C). For measurement, we use a laser probe method that we developed in advance; this method generates statistical measurements method using a laser distance meter. The measurement results obtained under the IFMIF nominal condition (15 m/s, 10{sup −3} Pa, 250 °C) at the IFMIF beam center are as follows: average target thickness = 26.08 ± 0.09 mm (2σ), mean wave amplitude = 0.26 ± 0.01 mm (2σ), and maximum wave amplitude = 1.46 ± 0.25 mm (2σ). Of the total wave components, 99.7% are within the design requirement. The analytically predicted target thickness is in excellent agreement with the experimental data, resulting in successful characterization of the Li target thickness.

  11. 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.)

  12. 49 CFR 215.103 - Defective wheel.

    Science.gov (United States)

    2010-10-01

    ... of the rim; or, (i) A wheel on the car has been welded unless the car is being moved for repair in... on the car shows evidence of being loose such as oil seepage on the back hub or back plate; (h) A...

  13. Encapsulation of polymer photovoltaic prototypes

    DEFF Research Database (Denmark)

    Krebs, Frederik C

    2006-01-01

    A simple and efficient method for the encapsulation of polymer and organic photovoltaic prototypes is presented. The method employs device preparation on glass substrates with subsequent sealing using glass fiber reinforced thermosetting epoxy (prepreg) against a back plate. The method allows...

  14. Comparison of Cooling Different Parts in a High Pressure Ratio Centrifugal Compressor

    Directory of Open Access Journals (Sweden)

    S. Mostafa Moosania

    2016-12-01

    Full Text Available Cooling in a centrifugal compressor can improve the performance and reduce the impeller temperature. In a centrifugal compressor, external walls can be cool down, which is known as the shell cooling. This method avoids undesirable effects induced by other cooling methods. Cooling can be applied on different external walls, such as the shroud, diffuser or the back plate. This paper focuses on seeking the most effective cooling place to increase the performance and reduce the impeller temperature. It is found that shroud cooling improves the compressor performance the most. Shroud cooling with 2400 W of cooling power increases the pressure ratio by 4.6% and efficiency by 1.49%. Each 500 W increase in the shroud cooling power, increases the efficiency by 0.3%. Diffuser cooling and back plate cooling have an identical effect on the polytropic efficiency. However, back plate cooling increases the pressure ratio more than diffuser cooling. Furthermore, only back plate cooling reduces the impeller temperature, and with 2400 W of cooling power, the impeller temperature reduces by 45 K.

  15. The activation system Easy 2007

    International Nuclear Information System (INIS)

    Forrest, R.A.; Kopecky, J.

    2007-01-01

    Full text of publication follows: Safety and waste management of materials for ITER, IFMIF and future power plants require detailed knowledge of the activation caused by irradiation with neutrons, or in the case of IFMIF, deuterons. The European Activation System (EASY) has been developed for such calculations and a new version (EASY-2007) was released earlier this year. This contains a large amount of nuclear data in the European Activation File (EAF-2007) covering neutron-, deuteron- and proton-induced cross sections (about 200,000 reactions have data extending up to 60 MeV), decay data (2,231 nuclides) and subsidiary data on e.g. biological hazards. These data are input to the FISPACT inventory code used to calculate the activation. Recent work has concentrated on the validation of EASY-2007 using integral and differential measurements; these studies are summarised showing examples of reactions agreeing with the experimental results and cases where the library data require further improvement. Integral data above 20 MeV are especially important in improving the library for IFMIF calculations. Using a previous version of EASY a study of the activation of all the elements enabled the identification of the reactions important in producing activation below 20 MeV. The list of 1,340 neutron-reactions producing the dominant radio-nuclides enables further studies to be focused on the important data. This study made extensive use of importance diagrams. This work has been extended to cover the energy region up to 60 MeV, and the new important radionuclides and reactions in this energy range are reported. Although the data above 20 MeV are important for IFMIF and are of interest because of their novelty, the traditional energy region below 20 MeV remains of great importance for most fusion applications. The testing of such large data libraries for reactions with no experimental data is necessary and results from the use of the recently developed method of Statistical

  16. Conceptual Study of the LB/TS (Large Blast/Thermal Simulator) Instrumentation, Data Acquisition and Facility Controls System.

    Science.gov (United States)

    1984-09-12

    length cabanos on Nrm size -Auoo-ore 𔃻ra"Selctin dgoo: Atcm~tcI i aoanc atar ~ nuberof Lens mounr Ftuggedzea bayonet mount for Nexon F standard ens...Bibliographic Input Report No.: ARBRL-MR-03207; SBI-AD-F300 118 Nov 82 92 p Languages: English NTIS Prices: PC A05/MF A01 Journal Announcement: GRA18307 Country...Proving Ground, MD, Ballistic Research Lab A -I Corp. Source Codes: 054817004; 393471 Report No.: ARBRL-TR-02367 September 81 91p Languages: English

  17. Handling device for nuclear fuel assemblies and assembly appropriate for such a device

    International Nuclear Information System (INIS)

    Cransac, J.P.; Jaquelin, R.; Renaux, C.

    1985-01-01

    The handling device comprises a guide tube of which axis is vertical, in which a grab moves, hanging from a chain, under the action of a back-geared motor. The grab being stopped in its rotation in the guide tube, an assembly can be gripped with a bayonet system while controlling the rotation of the grab - guide tube system a back-geared motor. The device can be hanged from the small or large rotating plug of a fast neutron reactor. It can be used in a handling flask [fr

  18. Cold Steel, Weak Flesh: Mechanism, Masculinity and the Anxieties of Late Victorian Empire.

    Science.gov (United States)

    Brown, Michael

    2017-03-15

    This article considers the reception and representation of advanced military technology in late nineteenth- and early twentieth-century Britain. It argues that technologies such as the breech-loading rifle and the machine gun existed in an ambiguous relationship with contemporary ideas about martial masculinities and in many cases served to fuel anxieties about the physical prowess of the British soldier. In turn, these anxieties encouraged a preoccupation in both military and popular domains with that most visceral of weapons, the bayonet, an obsession which was to have profound consequences for British military thinking at the dawn of the First World War.

  19. The European fusion program and the role of the research reactors

    International Nuclear Information System (INIS)

    Laesser, R.; Andreani, R.; Diegele, E.

    2005-01-01

    The main objectives of the European long-term Fusion Technology Program are i) investigation of DEMO breeding blankets options, ii) development of low activation materials resistant to high neutron fluence, iii) construction of IFMIF for validation of DEMO materials, and iv) promotion of modelling efforts for the understanding of radiation damage. A large effort is required for the development and performance verification of the materials subjected to the intense neutron irradiation encountered in fusion reactors. In the absence of a strong 14.1 MeV neutron source fission materials research reactors are used. Elaborate in-pile and post-irradiation examinations are performed. In addition, the modelling effort is increased to predict the damage by a 'true' fusion spectrum in the future. Even assuming that a positive decision for IFMIF construction can be reached, the operation of a limited number of materials test reactors is needed to perform irradiation studies on large samples and for screening. (author)

  20. Outline of application plans of accelerator beams in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Yasuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Japan Atomic Energy Research Institute (JAERI) has various application plans of accelerators such as; Neutron Science Research Complex (NSRC), Positron Factory, International Fusion Material Irradiation Facility (IFMIF), and Spring-8 Project. Each application plan has its own research program and its own core accelerator. The NSRC is a multi-purpose research complex composed of seven research facilities: slow neutron scattering facility for material science, the nuclear energy research facility like nuclear transmutation and so on. The Positron Factory will be applied to the research of precise analysis of material structure by novel method of positron probing. The IFMIF aims at simulating the wall loading of a demo fusion reactor by producing high intense neutron flux. The SPring-8 is the largest synchrotron radiation source in the world. More than 60 X-ray beam lines will be equipped for the various researches. (author)

  1. EPICS based low-level radio frequency control system in LIPAc

    Energy Technology Data Exchange (ETDEWEB)

    Calvo, Julio, E-mail: julio.calvo@ciemat.es [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, Ciemat (Spain); Rivers, Mark L. [Department of Geophysical Sciences and Center for Advanced Radiation Sources, The University of Chicago (United States); Patricio, Miguel A. [Departamento de Informatica, Universidad Carlos III de Madrid (Spain); Ibarra, Angel [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, Ciemat (Spain)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer The system proposed can control amplitude and phase of each cavity. Black-Right-Pointing-Pointer Rapid diagnostics are refreshed in milliseconds. Black-Right-Pointing-Pointer Increasing control parameters will not increase consumed time neither complexity. Black-Right-Pointing-Pointer IQ demodulation can be achieved thanks to the transformed values at driver level. - Abstract: The IFMIF-EVEDA (International Fusion Materials Irradiation Facility - Engineering Validation and Engineering Design Activity) linear accelerator, known as Linear IFMIF Prototype Accelerator (LIPAc), will be a 9 MeV, 125 mA continuous wave (CW) deuteron accelerator prototype to validate the technical options of the accelerator design for IFMIF. The primary mission of such facility is to test and verify materials performance when subjected to extensive neutron irradiation of the type encountered in a fusion reactor to prepare for the design, construction, licensing and safe operation of a fusion demonstration reactor (DEMO). The radio frequency (RF) power system of IFMIF-EVEDA consists of 18 RF chains working at 175 MHz with three amplification stages each. The low-level radio frequency (LLRF) controls the amplitude and phase of the signal to be synchronized with the beam and it also controls the resonance frequency of the cavities. The system is based on a commercial compact peripheral component interconnect (cPCI) field programmable gate array (FPGA) board, provided by Lyrtech and controlled by a Windows host PC. For this purpose, it is mandatory to communicate the cPCI FPGA board from EPICS Channel Access [1]. A software architecture on EPICS framework in order to control and monitor the LLRF system is presented.

  2. Water experiment of high-speed, free-surface, plane jet along concave wall

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Ida, Mizuho; Kato, Yoshio; Maekawa, Hiroshi; Itoh, Kazuhiro; Kukita, Yutaka

    1997-01-01

    In the International Fusion Materials Irradiation Facility (IFMIF), an intense 14 MeV neutron beam will be generated in the high-speed liquid lithium (Li) plane jet target flowing along concave wall in vacuum. As part of the conceptual design activity (CDA) of the IFMIF, the stability of the plane liquid jet flow was studied experimentally with water in a well-defined channel geometry for non-heating condition. A two-dimensional double-reducer nozzle being newly proposed for the IFMIF target successfully provided a high-speed (≤ 17 m/s) stable water jet with uniform velocity distribution at the nozzle exit without flow separation in the nozzle. The free surface of the jet was covered by two-dimensional and/or three-dimensional waves, the size of which did not change much over the tested jet length of ∼130 mm. The jet velocity profile changed around the nozzle exit from uniform to that of free-vortex flow where the product of the radius of stream line and local velocity is constant in the jet thickness. The jet thickness increased immediately after exiting the nozzle because of the velocity profile change. The predicted jet thickness by a modified one-dimensional momentum model agreed with the data well. (author)

  3. Safety assessment of a lithium target

    International Nuclear Information System (INIS)

    Burgazzi, Luciano; Roberta, Ferri; Barbara, Giannone

    2006-01-01

    This paper addresses the safety assessment of the lithium target of the International Fusion Materials Irradiation Facility (IFMIF) through evaluating the most important risk factors related to system operation and verifying the fulfillment of the safety criteria. The hazard assessment is based on using a well-structured Failure Mode and Effect Analysis (FMEA) procedure by detailing on a component-by-component basis all the possible failure modes and identifying their effects on the plant. Additionally, a systems analysis, applying the fault tree technique, is performed in order to evaluate, from a probabilistic standpoint, all the relevant and possible failures of each component required for safe system operation and assessing the unavailability of the lithium target system. The last task includes the thermal-hydraulic transient analysis of the target lithium loop, including operational and accident transients. A lithium target loop model is developed, using the RELAP5/Mod3.2 thermal-hydraulic code, which has been modified to include specific features of IFMIF itself. The main conclusions are that target safety is fulfilled, the hazards associated with lithium operation are confined within the IFMIF security boundaries, the environmental impact is negligible, and the plant responds to the simulated transients by being able to reach steady conditions in a safety situation

  4. Reducing risk and accelerating delivery of a neutron source for fusion materials research

    Energy Technology Data Exchange (ETDEWEB)

    Surrey, E., E-mail: Elizabeth.Surrey@ccfe.ac.uk [EURATOM/CCFE, Abingdon OX14 3DB (United Kingdom); Porton, M. [EURATOM/CCFE, Abingdon OX14 3DB (United Kingdom); Davenne, T.; Findlay, D.; Letchford, A.; Thomason, J. [STFC Rutherford Appleton Laboratory, Harwell OX11 0QX (United Kingdom); Roberts, S.G.; Marrow, J.; Seryi, A. [University of Oxford, Oxford OX1 3DP (United Kingdom); Connolly, B. [University of Birmingham, Edgbaston B15 2TT (United Kingdom); Owen, H. [University of Manchester, Manchester M13 9PL (United Kingdom)

    2014-04-15

    Highlights: • Proposed neutron source for fusion materials – FAFNIR – n(d,C) stripping source. • Near term technology, reduces risk compared with IFMIF, timely data production. • Technical, economic and programme needs assessed, compatible with EU Roadmap proposals. • Safety case impacts regulatory role for source, now mainly stakeholder insurance. - Abstract: The materials engineering database relevant to fusion irradiation is poorly populated and it has long been recognized that a fusion spectrum neutron source will be required, the facility IFMIF being the present proposal. Re-evaluation of the regulatory approach for the EU proposed DEMO device shows that the purpose of the source can be changed from lifetime equivalent irradiation exposure to data generation at lower levels of exposure by adopting a defence in depth strategy and regular component surveillance. This reduces the specification of the source with respect to IFMIF allowing lower risk technology solutions to be considered. A description of such a source, the Facility for Fusion Neutron Irradiation Research, FAFNIR, is presented here along with project timescales and costs.

  5. Magnetic light cloaking control in the marine planktonic copepod Sapphirina

    Science.gov (United States)

    Kashiwagi, H.; Mizukawa, Y.; Iwasaka, M.; Ohtsuka, S.

    2017-05-01

    We investigated the light cloaking behavior of the marine planktonic copepod Sapphirina under a magnetic field. Optical interferences in the multi-laminated guanine crystal layer beneath the dorsal body surface create a brilliant structural color, which can be almost entirely removed by changing the light reflection. In the investigation, we immersed segments of Sapphirina in seawater contained in an optical chamber. When the derived Sapphirina segments were attached to the container surface, they were inert to magnetic fields up to 300 mT. However, when the back plate segments were attached to the substrate at a point, with most of the plate floating in the seawater, the plate rotated oppositely to the applied magnetic field. In addition, the brilliant parts of the Sapphirina back plate rotated backward and forward by changing the magnetic field directions. Our experiment suggests a new model of an optical micro-electro-mechanical system that is controllable by magnetic fields.

  6. Investigations on 3-dimensional temperature distribution in a FLATCON-type CPV module

    Science.gov (United States)

    Wiesenfarth, Maike; Gamisch, Sebastian; Kraus, Harald; Bett, Andreas W.

    2013-09-01

    The thermal flow in a FLATCON®-type CPV module is investigated theoretically and experimentally. For the simulation a model in the computational fluid dynamics (CFD) software SolidWorks Flow Simulation was established. In order to verify the simulation results the calculated and measured temperatures were compared assuming the same operating conditions (wind speed and direction, direct normal irradiance (DNI) and ambient temperature). Therefore, an experimental module was manufactured and equipped with temperature sensors at defined positions. In addition, the temperature distribution on the back plate of the module was displayed by infrared images. The simulated absolute temperature and the distribution compare well with an average deviation of only 3.3 K to the sensor measurements. Finally, the validated model was used to investigate the influence of the back plate material on the temperature distribution by replacing the glass material by aluminum. The simulation showed that it is important to consider heat dissipation by radiation when designing a CPV module.

  7. Microstructure Evolution and Composition Control during the Processing of Thin-gage Metallic Foil (Preprint)

    Science.gov (United States)

    2012-02-01

    applications requiring characteristics such as light weight, high structural stiffness, or low thermal conductivity. Ductile, low temperature metals such as...was EDM’ed from the billet/ingot, stress relieved, finish ground, brazed onto an oxygen-free high -conductivity copper backing plate, and attached to...of each alloying element and hence the composition of the deposit. The substrates were a high - temperature alloy steel. They were heated to a

  8. Analytical and experimental investigation of the elastic and plastic behavior of plates on foundations subjected to dynamic punch loading

    International Nuclear Information System (INIS)

    Duffey, T.A.; Sutherland, S.H.; Cheresh, M.

    1980-01-01

    Analytical solutions and experimental results are presented for the response of foundation-backed plates to static and dynamic punch loading. Tests were performed on polyurethane foam-backed and unbacked plates; plates were centrally loaded over a range-in plastic deformations up to complete failure. This is part of an attempt to understand the puncture resistance of the sidewalls of containers used to ship hazardous wastes

  9. Encapsulation of polymer photovoltaic prototypes

    Energy Technology Data Exchange (ETDEWEB)

    Krebs, Frederik C. [The Danish Polymer Centre, RISOE National Laboratory, P.O. Box 49, DK-4000 Roskilde (Denmark)

    2006-12-15

    A simple and efficient method for the encapsulation of polymer and organic photovoltaic prototypes is presented. The method employs device preparation on glass substrates with subsequent sealing using glass fiber reinforced thermosetting epoxy (prepreg) against a back plate. The method allows for transporting oxygen and water sensitive devices outside a glove box environment after sealing and enables sharing of devices between research groups such that efficiency and stability can be evaluated in different laboratories. (author)

  10. A revised design approach of the attachment system for the ITER EU-HCPB-TBM based on a central cylindrical connection element

    International Nuclear Information System (INIS)

    Zeile, Christian; Neuberger, Heiko

    2012-01-01

    Highlights: ► Design of an attachment system based on a cylinder to connect TBM and shield. ► Attachment system has to cope with high EM loads and different thermal expansions. ► Stiff design and central position fulfill these requirements. ► Transient thermal-structural analyses confirm compliance of design with design codes. - Abstract: The EU-Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM), which is located inside an equatorial port plug, is attached to the shield by an attachment system. The design of the attachment system has to fulfill two conflicting requirements. On the one hand, it has to transfer the high electromagnetic forces acting on the TBM to the shield and on the other hand, it has to compensate the different thermal expansions between the shield and the back plate of the TBM. The recent design approach of the attachment system consists of a hollow cylinder located at the center of the back plate. This design combines two advantages: a simple geometry and correspondingly low fabrication effort and the central location where the differential strain between back plate and shield is minimal. Static and transient thermal-structural analyses of the most demanding load cases, a fast vertical displacement event type II and the operation state tritium outgassing, have been performed to evaluate the design and confirm the compliance with the relevant design codes. A welded connection of the attachment system to the TBM back plate and a bolted connection in combination with a splined shaft is proposed for the shield side because of the dissimilar materials.

  11. Picosecond Laser Pulse Interactions with Metallic and Semiconductor Surfaces.

    Science.gov (United States)

    1984-11-01

    thermometric determination of plasma relaxation is by far more sensitive than direct optical measurements. The solid line in Fig. 4 shows the calculated...passively mode-locked Nd:yttrium aluminum garnet in Si, several researchers have used high picosecond or fem- laser was used to produce single 30-ps, 1.06...these targets to an aluminum backing plate with a silver-epoxy conducting glue (Ablestik). The conductivity of the targets was high enough to make

  12. Distillation, destructive: gas retorts

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, G; Buchanan, J

    1872-06-12

    Retorts used in the distillation of coal, cannel, shale, or other minerals are discharged by means of a joint metal bar or a chain inserted prior to or with the charge, and fitted with a plate or series of plates preferably with the back plate the largest. The rod or chain is formed with a hook, loop, or coupling for withdrawal, preferably by a steam windlass running on rails.

  13. Defect reduction in seeded aluminum nitride crystal growth

    Science.gov (United States)

    Bondokov, Robert T.; Schowalter, Leo J.; Morgan, Kenneth; Slack, Glen A; Rao, Shailaja P.; Gibb, Shawn Robert

    2017-09-26

    Bulk single crystal of aluminum nitride (AlN) having an areal planar defect density.ltoreq.100 cm.sup.-2. Methods for growing single crystal aluminum nitride include melting an aluminum foil to uniformly wet a foundation with a layer of aluminum, the foundation forming a portion of an AlN seed holder, for an AlN seed to be used for the AlN growth. The holder may consist essentially of a substantially impervious backing plate.

  14. Research and Development for Robotic Transportable Waste to Energy System (TWES)

    Science.gov (United States)

    2012-01-01

    Feedmiser package - 100-gal duplex Feedmiser package, including two hp Grundfos pumps rated to 15 gpm at 602 ft of head and a 100-gal tank The...observed. This was possibly from a bad weld between the back plate and a fire tube, or a leak from a tube itself. This problem warrants further...for damages (to be inserted where spool piece resides on 2” MPS header) Completed • Inspect welds on pipe supports and boiler Completed TWES

  15. TFTR movable limiter instrumentation and controls

    International Nuclear Information System (INIS)

    Frankenberg, J.; Collins, D.; Kaufmann, D.; Mamoun, A.

    1983-01-01

    The TFTR movable limiter is a single poloidal limiter located within one 18 /SUP o/ segment of the vacuum vessel. It consists of three (3) interconnected inconel backing plates covered with titanium carbide coated graphite tiles. The backing plates are positioned by three independent screw drive actuators. Cooling water is fed through the horizontal port cover to tubes brazed onto the backs of the backing plates. Thermocouples monitor the limiter temperature. (1) and more fully described in refs. (1) and (2). The positioning actuators are driven by independently controlled DC servo motors, controlled either locally or from CICADA. Drive motor shaft position is monitored by chain driven encoders and potentiometers. Limiter blade position can be varied to suit any plasma within the operating range. CICADA is programmed to keep the limiter stroke within safe operating limits. A microprocessor duplicates the CICADA protective function allowing limiter operation without CICADA. The potentiometer signal is sent to an analog computer, which safeguards the limiter against failure of the encoders or the micro-processor. Cooling water flows through the limiter in 3 separate paths, one for each blade. The flow rate and temperature rise through each loop are measured accurately to allow CICADA to calculate the heat into each blade. The water system is also interlocked and alarmed to prevent dumping of water into the vacuum vessel

  16. 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)

  17. 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)

  18. Design of TFTR movable limiter blades for ohmic and neutral-beam-heated plasmas

    International Nuclear Information System (INIS)

    Doll, D.W.; Ulrickson, M.A.; Cecchi, J.L.; Citrolo, J.C.; Weissenburger, D.; Bialek, J.

    1981-10-01

    A new set of movable limiter blades has been designed for TFTR that will meet both the requirements of the 4 MW ohmic heated and the 33 MW neutral beam heated plasmas. This is accomplished with three limiter blades each having and elliptical shape along the toroidal direction. Heat flux levels are acceptable for both ohmic heated and pre-strong compression plasmas. The construction consists of graphite tiles attached to cooled backing plates. The tiles have an average thickness of approx. 4.7 cm and are drawn against the backing plate with spring loaded fasteners that are keyed into the graphite. The cooled backing plate provides the structure for resisting disruption and fault induced loads. A set of rollers attached to the top and bottom blades allow them to be expanded and closed in order to vary the plasma surface for scaling experiments. Water cooling lines penetrate only the mid-plane port cover/support plate in such a way as to avoid bolted water connections inside the vacuum boundary and at the same time allow blade movement. Both the upper and lower blades are attached to the mid-plane limiter blade through pivots. Pivot connections are protected against arcing with an alumina coating and a shunt bar strap. Remote handling is considered throughout the design

  19. High power beam dump project for the accelerator prototype LIPAc: cooling design and analysis

    International Nuclear Information System (INIS)

    Parro Albeniz, M.

    2015-01-01

    In the nuclear fusion field running in parallel to ITER (International Thermonuclear Experimental Reactor) as one of the complementary activities headed towards solving the technological barriers, IFMIF (International Fusion Material Irradiation Facility) project aims to provide an irradiation facility to qualify advanced materials resistant to extreme conditions like the ones expected in future fusion reactors like DEMO (DEMOnstration Power Plant). IFMIF consists of two constant wave deuteron accelerators delivering a 125 mA and 40 MeV beam each that will collide on a lithium target producing an intense neutron fluence (1017 neutrons/s) with a similar spectra to that of fusion neutrons [1], [2]. This neutron flux is employed to irradiate the different material candidates to be employed in the future fusion reactors, and the samples examined after irradiation at the so called post-irradiative facilities. As a first step in such an ambitious project, an engineering validation and engineering design activity phase called IFMIF-EVEDA (Engineering Validation and Engineering Design Activities) is presently going on. One of the activities consists on the construction and operation of an accelerator prototype named LIPAc (Linear IFMIF Prototype Accelerator). It is a high intensity deuteron accelerator identical to the low energy part of the IFMIF accelerators. The LIPAc components, which will be installed in Japan, are delivered by different european countries. The accelerator supplies a 9 MeV constant wave beam of deuterons with a power of 1.125 MW, which after being characterized by different instruments has to be stopped safely. For such task a beam dump to absorb the beam energy and take it to a heat sink is needed. Spain has the compromise of delivering such device and CIEMAT (Centro de Investigaciones Energéticas Medioambientales y Tecnológicas) is responsible for such task. The central piece of the beam dump, where the ion beam is stopped, is a copper cone with

  20. Out pile test of a disassembly tool for the intermediate examination of nuclear fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Joung, Chang-Young; Ahn, Sung-Ho; Yang, Tae-Ho; Jang, Seo-Yoon; Park, Seung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The two fuel rod assemblies are assembled with a bayonet coupler, and the non-instrumented fuel rod assembly can be disassembled for intermediate examination. A tool to disassemble the non-instrumented fuel rod assembly from the test rig was developed, and steel wires are connected to the tool to operate release function. In this study, an assembly plug with a quick plug typed bayonet coupler and the accompanying disassembly tool was designed to prevent the interference problem. A test rig mockup was fabricated, and performance test was carried out in the laboratory. And, the out pile test was also carried out in the single channel test loop established in the KAERI. In this study, a modified coupler design to disassemble the non-instrumented fuel rod assembly from the test rig for the intermediate examination was suggested to solve interference problem of previous design. The performance of the modified design was verified by test mockup fabricated with the modified coupler design and accompanied disassembly tool design. Finally, out pile test was carried out in the single channel test loop in the KAERI, and the test rig and the disassembly tool showed good performance and reliability. The developed technique will be useful to the periodic intermediate examination of nuclear fuel rods.

  1. Out pile test of a disassembly tool for the intermediate examination of nuclear fuel rods

    International Nuclear Information System (INIS)

    Hong, Jintae; Joung, Chang-Young; Ahn, Sung-Ho; Yang, Tae-Ho; Jang, Seo-Yoon; Park, Seung-Jae

    2016-01-01

    The two fuel rod assemblies are assembled with a bayonet coupler, and the non-instrumented fuel rod assembly can be disassembled for intermediate examination. A tool to disassemble the non-instrumented fuel rod assembly from the test rig was developed, and steel wires are connected to the tool to operate release function. In this study, an assembly plug with a quick plug typed bayonet coupler and the accompanying disassembly tool was designed to prevent the interference problem. A test rig mockup was fabricated, and performance test was carried out in the laboratory. And, the out pile test was also carried out in the single channel test loop established in the KAERI. In this study, a modified coupler design to disassemble the non-instrumented fuel rod assembly from the test rig for the intermediate examination was suggested to solve interference problem of previous design. The performance of the modified design was verified by test mockup fabricated with the modified coupler design and accompanied disassembly tool design. Finally, out pile test was carried out in the single channel test loop in the KAERI, and the test rig and the disassembly tool showed good performance and reliability. The developed technique will be useful to the periodic intermediate examination of nuclear fuel rods

  2. HYBRID SULFUR CYCLE FLOWSHEETS FOR HYDROGEN PRODUCTION USING HIGH-TEMPERATURE GAS-COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.

    2011-07-06

    Two hybrid sulfur (HyS) cycle process flowsheets intended for use with high-temperature gas-cooled reactors (HTGRs) are presented. The flowsheets were developed for the Next Generation Nuclear Plant (NGNP) program, and couple a proton exchange membrane (PEM) electrolyzer for the SO2-depolarized electrolysis step with a silicon carbide bayonet reactor for the high-temperature decomposition step. One presumes an HTGR reactor outlet temperature (ROT) of 950 C, the other 750 C. Performance was improved (over earlier flowsheets) by assuming that use of a more acid-tolerant PEM, like acid-doped poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI), instead of Nafion{reg_sign}, would allow higher anolyte acid concentrations. Lower ROT was accommodated by adding a direct contact exchange/quench column upstream from the bayonet reactor and dropping the decomposition pressure. Aspen Plus was used to develop material and energy balances. A net thermal efficiency of 44.0% to 47.6%, higher heating value basis is projected for the 950 C case, dropping to 39.9% for the 750 C case.

  3. Dynamic Management of Carrying Capacity in Mountain Heritage%山岳型遗产地环境容量动态管理研究

    Institute of Scientific and Technical Information of China (English)

    王德刚; 赵建峰; 黄潇婷

    2015-01-01

    管理的安全可控。%Assessment and management of heritage carrying capacity involve many disciplines. Traditional static measurements obtain the optimal value and saturation value for a certain space, which is the main technical method of traditional carrying capacity study. However, tourists, the subjects who affect carrying capacity of environment are dynamic. Especially, as to Mountain heritage sites, its environmental capacity are relatively limited for their topography characteristics. Massive and concentrated visits will not only increase the pressure of environmental protection, but also influence the quality of tourist experience, and even sometimes threaten tourist physical security. Traditional researches on tourism environmental capacity mainly depend on static measurement. By introducing time dimension and analyzing the rule of spatiotemporal distribution of tourist flow, this paper proposed a new method of dynamic management of carrying capacity in mountain heritage through putting forward the concept of “the spatiotemporal bayonet of tourist flow”, and recognizes it with time geography method. The results of data analysis of Mountain Tai show that there are seven spatial bayonets, which are Red Gate, Stone Valley, Half Way Gate to Heaven, South Gate to Heaven, Tianjie, Jade Emperor Peak and Riguan peak, while the temporal bayonets are decentralization relatively. All those temporal-spatial bayonets of the middle route in Mountain Tai only account for 5. 52% of the whole area, which attract almost 50. 04% tourists. Hence one can see that, recognition and management of Space-Time Bayonet of tourists flow is quite important to dynamic management of carrying capacity in mountain heritage. Besides, this paper discussed the academic contribution of the spatiotemporal bayonet of tourist flow and its potential application practically through regulating tourist ’s behavior, controlling tourism environmental capacity dynamically and building intelligent early-warning system three aspect

  4. Status and possible prospects of an international fusion materials irradiation facility

    International Nuclear Information System (INIS)

    Cozzani, F.

    1999-01-01

    Structural materials for future DT fusion power reactors will have to operate under intense neutron fields with energies up to 14 MeV and fluences in the order of 2 MW/m 2 per year. As environmental acceptability, safety considerations and economic viability will be ultimately the keys to the widespread introduction of fusion power, the development of radiation-resistant and low activation materials would contribute significantly to fusion development. For this purpose, testing of materials under irradiation conditions close to those expected in a fusion power station would require the availability, in an appropriate time framework, of an intense, high-energy neutron source. Recent advances in linear accelerator technology, in small specimens testing technology, and in the comprehension of damage phenomena, lead to the conclusion that an accelerator-based D-Li neutron source, with beam energy variability, would provide the most realistic option for a fusion materials testing facility. Under the auspices of the IEA, an international effort (EU, Japan, US, RF) to carry out the conceptual design activities (CDA) of an international fusion materials irradiation facility (IFMIF), based on the D-Li concept, have been carried out successfully. A final conceptual design report was produced at the end of 1996. A phase of conceptual design evaluation (CDE), presently underway, is extending and further refining some of the conceptual design details of IFMIF. The results indicate that an IFMIF-class installation would be technically feasible and could meet its mission objectives. However, a suitable phase of Engineering Validation, to carry out some complementary R and D and prototyping, would still be needed to resolve a few key technical uncertainties before the possibility to proceed toward detailed design and construction could be explored. (orig.)

  5. Integral activation experiment of fusion reactor materials with d-Li neutrons up to 55 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Moellendorff, Ulrich von [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Wada, Masayuki [Business Automation Co., Ltd., Tokyo (Japan)

    2000-03-01

    An integral activation experiment of fusion reactor materials with a deuteron-lithium neutron source was performed. Since the maximum energy of neutrons produced was 55 MeV, the experiment with associated analysis was one of the first attempts for extending the energy range beyond 20 MeV. The following keywords represent the present study: d-Li neutrons, 55 MeV, dosimetry, SAND-II, spectrum adjustment, LA-150, MCNP, McDeLi, IFMIF, fusion reactor materials, integral activation experiment, low-activation, F82H, vanadium-alloy, IEAF, ALARA, and sequential charged particle reaction. (author)

  6. Computational methods, tools and data for nuclear analyses of fusion technology systems

    International Nuclear Information System (INIS)

    Fischer, U.

    2006-01-01

    An overview is presented of the Research and Development work conducted at Forschungszentrum Karlsruhe in co-operation with other associations in the framework of the European Fusion Technology Programme on the development and qualification of computational tools and data for nuclear analyses of Fusion Technology systems. The focus is on the development of advanced methods and tools based on the Monte Carlo technique for particle transport simulations, and the evaluation and qualification of dedicated nuclear data to satisfy the needs of the ITER and the IFMIF projects. (author)

  7. Protective clothing

    International Nuclear Information System (INIS)

    Winnett, G.F.

    1979-01-01

    A protective suit used for isolating its wearer from radioactively contaminated areas is described in three parts. The first part includes the covering for the wearer's head, arms and upper body and at the waist is releasably fitted around an opening into the contaminated area. The second part includes the legs of the suit and is releasably connectible to the first part of the suit to enclose the wearer who is then supplied with air through an umbilical pipe. A further part surrounds the second part and is releasably connectible to it, enclosing a space between the parts. This further part is also releasably connectible to the opening at the waist to prevent egress from the contaminated area. The releasable connections between the parts may be bayonet type fittings or may be rotating T-shaped projections which engage in T-shaped grooves. (author)

  8. Adaption of the LHC cold mass cooling system to the requirements of the Future Circular Collider (FCC)

    Science.gov (United States)

    Kotnig, C.; Tavian, L.; Brenn, G.

    2017-12-01

    The cooling of the superconducting magnet cold masses with superfluid helium (He II) is a well-established concept successfully in operation for years in the LHC. Consequently, its application for the cooling of FCC magnets is an obvious option. The 12-kW heat loads distributed over 10-km long sectors not only require an adaption of the magnet bayonet heat exchangers but also present new challenges to the cryogenic plants, the distribution system and the control strategy. This paper recalls the basic LHC cooling concept with superfluid helium and defines the main parameters for the adaption to the FCC requirements. Pressure drop and hydrostatic head are developed in the distribution and pumping systems; their impact on the magnet temperature profile and the corresponding cooling efficiency is presented and compared for different distribution and pumping schemes.

  9. Liquid metal steam generator

    International Nuclear Information System (INIS)

    Wolowodiuk, W.

    1975-01-01

    A liquid metal heated steam generator is described which in the event of a tube failure quickly exhausts out of the steam generator the products of the reaction between the water and the liquid metal. The steam is generated in a plurality of bayonet tubes which are heated by liquid metal flowing over them between an inner cylinder and an outer cylinder. The inner cylinder extends above the level of liquid metal but below the main tube sheet. A central pipe extends down into the inner cylinder with a centrifugal separator between it and the inner cylinder at its lower end and an involute deflector plate above the separator so that the products of a reaction between the liquid metal and the water will be deflected downwardly by the deflector plate and through the separator so that the liquid metal will flow outwardly and away from the central pipe through which the steam and gaseous reaction products are exhausted. (U.S.)

  10. HIGHTEX: a computer program for the steady-state simulation of steam-methane reformers used in a nuclear process heat plant

    International Nuclear Information System (INIS)

    Tadokoro, Yoshihiro; Seya, Toko

    1977-08-01

    This report describes a computational model and the input procedure of HIGHTEX, a computer program for steady-state simulation of the steam-methane reformers used in a nuclear process heat plant. The HIGHTEX program simulates rapidly a single reformer tube, and treats the reactant single-phase in the two-dimensional catalyst bed. Output of the computer program is radial distributions of temperature and reaction products in the catalyst-packed bed, pressure loss of the packed bed, stress in the reformer tube, hydrogen permeation rate through the reformer tube, heat rate of reaction, and heat-transfer rate between helium and process gas. The running time (cpu) for a 9m-long bayonet type reformer tube is 12 min with FACOM-230/75. (auth.)

  11. Booster cryogenics

    International Nuclear Information System (INIS)

    Storm, D.W.; Weitkamp, W.G.; Will, D.I.

    1984-01-01

    During this past year the authors have ordered a helium refrigerator, developed cryostat specifications and come to understand better some of the potential problems to avoid in helium distribution systems. The helium refrigerator consists of a Koch Process Systems 2800HR with three type RS screw compressors. The 2800HR has two dry expansion engines, each with two 3'' diameter pistons, and one wet expansion engine with a single 2'' diameter piston. It has guaranteed capacities at 4.5 0 K of 440 W without liquid nitrogen precool and of 510 W with liquid nitrogen precool which compare favorably with the estimated need of 300 W. At present the authors have nearly completed material, technique and performance specifications for their cryostats, pending a decision on bayonet design, and the authors are beginning preliminary specifications for their liquid helium distribution manifold and transfer siphons

  12. Air box shock absorber for a nuclear reactor

    International Nuclear Information System (INIS)

    Pradhan, A.V.; George, J.A.

    1977-01-01

    Disclosed is an air box type shock absorber primarily for use in an ice condenser compartment of a nuclear reactor. The shock absorber includes a back plate member and sheet metal top, bottom, and front members. The front member is prefolded, and controlled clearances are provided among the members for predetermined escape of air under impact and compression. Prefolded internal sheet metal stiffeners also absorb a portion of the kinetic energy imparted to the shock absorber, and limit rebound. An external restraining rod guided by restraining straps insures that the sheet metal front member compresses inward upon impact. 6 claims, 11 figures

  13. Monte Carlo analysis of helium production in the ITER shielding blanket module

    International Nuclear Information System (INIS)

    Sato, S.

    1999-01-01

    In order to examine the shielding performances of the inboard blanket module in the international thermonuclear experimental reactor (ITER), shielding calculations have been carried out using a three-dimensional Monte Carlo method. The impact of radiation streaming through the front access holes and gaps between adjacent blanket modules on the helium gas production in the branch pipe weld locations and back plate have been estimated. The three-dimensional model represents an 18 sector of the overall torus region and includes the vacuum vessel, inboard blanket and back plate, plasma region, and outboard reflecting medium. And it includes the 1 m high inboard mid-plane module and the 20 mm wide gaps between adjacent modules. From the calculated results for the reference design, it has been found that the helium production at the plug of the branch pipe is four to five times higher than the design goal of 1 appm for a neutron fluence of 0.9 MW a m -2 at the inboard mid-plane first wall. Also, it has been found that the helium production at the back plate behind the horizontal gap is about three times higher than the design goal. In the reference design, the stainless steel (SS):H 2 O composition in the blanket module is 80:20%. Shielding calculations also have been carried out for the SS:H 2 O composition of 70:30, 60:40, 50:50 and 40:60%. From the evaluated results for their design, it has been found that the dependence of helium production on the SS:H 2 170 mm will reduce helium production to satisfy the design goal and not have a significant impact on weight limitations imposed by remote maintenance handling limitations. Also based on the calculated results, about 200 mm thick shields such as a key structure in the vertical gap are suggested to be installed in the horizontal gap as well to reduce the helium production at the back plate and to satisfy the design goal. (orig.)

  14. Acoustic Panel Liner for an Engine Nacelle

    Science.gov (United States)

    Jones, Michael G. (Inventor); Nark, Douglas M. (Inventor); Ayle, Earl (Inventor); Ichihashi, Fumitaka (Inventor)

    2016-01-01

    An acoustic panel liner includes a face sheet, back plate, and liner core positioned there-between, which may be used in an engine nacelle. Elongated chambers contain variable amounts of septa at a calibrated depth or depths. The septa may have varying DC flow resistance. The chambers may have a hexagonal or other polygonal cross sections. The septa, such as mesh caps, may be bonded to an inner wall of a corresponding chamber. The insertion depths may be the same or different. If different, the pattern of distribution of the depths may be randomized.

  15. TFTR bumper limiter and final protective plate engineering, fabrication and assembly

    International Nuclear Information System (INIS)

    Helmich, R.C.; Snook, P.G.; Loesser, G.D.; Reilly, T.B.; Trachsel, C.A.

    1986-01-01

    The inner vacuum vessel wall of the Tokamak Fusion Test Reactor (TFTR) is protected from plasma impingement by a bumper limiter and from neutral beam bombardment by protective plates. Engineering problems and solutions relating to Inconel 718, such as welding, machining in the annealed or age-hardened condition, selection of feeds, speeds and the need for rigid tooling are discussed. Vacuum furnace brazing of the 5/16'' Inconel 600 cooling tubing to the backing plates in both horizontal and vertical sections are shown. A detailed description of the plate and tile fabrication and assembly, with manufacturing and management techniques is outlined in this paper

  16. Experimental investigation on streaming due to a gap between blanket modules in ITER

    International Nuclear Information System (INIS)

    Konno, Chikara; Maekawa, Fujio; Oyama, Yukio; Uno, Yoshitomo; Kasugai, Yoshimi; Maekawa, Hiroshi; Ikeda, Yujiro; Wada, Masayuki

    2000-01-01

    A gap streaming experiment was performed by using a D-T neutron source at FNS/JAERI as the ITER/EDA R and D Task T-218, in order to examine the streaming effects due to gap between shield blanket modules in ITER. The experiment had three phases. The first one defined neutron source characteristics (Source Characterization Experiment), the second (Experiment-l ) aimed at shield for welding part between shield blanket and back plate and the third (Experiment-2) focused on the influence that the gap between shield blanket modules would have on superconducting magnet. The effects of gap streaming were examined in detail experimentally. (author)

  17. Effect of guideway discontinuities on magnetic levitation and drag forces

    International Nuclear Information System (INIS)

    Rossing, T.D.; Korte, R.; Hull, J.R.

    1991-01-01

    Transients in the lift and drag forces on a NdFeB permanent magnet were observed as the magnet passed over various discontinuities in a rotating aluminum disk at velocities of 4 to 25 m/s. For full cuts in the disk, the amplitude of the lift and drag transients and the wave form of the drag transient depend on the width, and the amplitudes are much larger than for partial cuts. The use of a backing plate to join two cut segments is ineffective

  18. JAEA’s R&D on the Thermochemical Hydrogen Production IS Process

    International Nuclear Information System (INIS)

    Kasahara, Seiji; Tanaka, Nobuyuki; Noguchi, Hiroki; Iwatsuki, Jin; Takegami, Hiroaki; Yan, Xing L.; Kubo, Shinji

    2014-01-01

    Japan Atomic Energy Agency (JAEA) has studied iodine-sulfur (IS) process, a thermochemical cycle to produce hydrogen by water splitting. This process is a candidate application of high temperature heat from high temperature gas-cooled reactors. This paper outlines the IS process study in JAEA, in particular recent situation of the R&D. Reactor components and a total process facility are tested to evaluate their integrity. A Bunsen reactor, a H_2SO_4 decomposer and a HI decomposer made of industrial materials such as SiC ceramic, fluoroplastic and lining materials have been examined separately as reactor components. A semibatch test and a thermal cycle test were operated in the Bunsen reactor. H_2SO_4 decomposition test is in a bayonet type reactor and HI decomposition test in an adiabatic radial flow type reactor are now under way. On the basis of a demonstration of continuous hydrogen production of 31 NL/h by a glass apparatus, an experimental apparatus of the total IS process has just been constructed to verify integrity of process components of industrial materials, H_2 production scale of which is 200 NL/h. Electro-electrodialysis (EED) cells to concentrate HI before distillation and a SiC-made bayonet type H_2SO_4 decomposer are applied in the facility. Process data of EED cells has been collected aiming to improve H_2 production thermal efficiency. Influence of temperature, composition in solution and existence of impurities on the cell properties has been investigated. Reduction of heat input to a HI separation step by applying the results of the study was shown. (author)

  19. Advanced light ion source extraction system for a new electron cyclotron resonance ion source geometry at Saclay

    Energy Technology Data Exchange (ETDEWEB)

    Delferriere, O.; Gobin, R.; Harrault, F.; Nyckees, S.; Sauce, Y.; Tuske, O. [Commissariat a l' Energie Atomique, CEA/Saclay, DSM/IRFU, 91191 Gif/Yvette (France)

    2012-02-15

    One of the main goal of intense light ion injector projects such as IPHI, IFMIF, or SPIRAL2, is to produce high current beams while keeping transverse emittance as low as possible. To prevent emittance growth induced in a dual solenoid low energy transfer line, its length has to be minimized. This can be performed with the advanced light ion source extraction system concept that we are developing: a new ECR 2.45 GHz type ion source based on the use of an additional low energy beam transport (LEBT) short length solenoid close to the extraction aperture to create the resonance in the plasma chamber. The geometry of the source has been considerably modified to allow easy maintenance of each component and to save space in front of the extraction. The source aims to be very flexible and to be able to extract high current ion beams at energy up to 100 kV. A specific experimental setup for this source is under installation on the BETSI test bench, to compare its performances with sources developed up to now in the laboratory, such as SILHI, IFMIF, or SPIRAL2 ECR sources. This original extraction source concept is presented, as well as electromagnetic simulations with OPERA-2D code. Ion beam extraction in space charge compensation regime with AXCEL, and beam dynamics simulation with SOLMAXP codes show the beam quality improvement at the end of the LEBT.

  20. Numerical analysis of high-speed liquid lithium free-surface flow

    International Nuclear Information System (INIS)

    Gordeev, Sergej; Heinzel, Volker; Stieglitz, Robert

    2012-01-01

    Highlights: ► The free surface behavior of a high speed lithium jet is investigated by means of a CFD LES analysis. ► The study is aiming to validate adequate LES technique. ► The Osaka University experiments with liquid lithium jet have been simulated. ► Four cases with jet flow velocities of 4, 9, 13 and 15 m/s are analyzed. ► Calculation results show a good qualitative and a quantitative agreement with the experimental data. - Abstract: The free-surface stability of the target of the International Fusion Material Irradiation Facility (IFMIF) is one of the crucial issues, since the spatio-temporal behavior of the free-surface determines the neutron flux to be generated. This article investigates the relation between the evolution of a wall boundary layer in a convergent nozzle and the free surface shape of a high speed lithium jet by means of a CFD LES analysis using the Osaka University experiments. The study is aiming to validate adequate LES technique to analyze the individual flow phenomena observed. Four cases with jet flow velocities of 4, 9, 13 and 15 m/s are analyzed. First analyses of calculation results show that the simulation exhibits a good qualitative and a quantitative agreement with the experimental data, which allows in the future a more realistic prediction of the IFMIF target behavior.

  1. Proceedings of the IEA-technical workshop on the test cell system for an international fusion materials irradiation facility, Karlsruhe, Germany, July 3-6, 1995. IEA-implementing agreement for a programme of research and development on fusion materials

    International Nuclear Information System (INIS)

    Moeslang, A.; Lindau, R.

    1995-09-01

    After a Conceptual Design Activity (CDA) study on an International Fusion Material Irradiation Facility (IFMIF) has been launched under the auspices of the IEA, working groups and relevant tasks have been defined and agreed in an IEA-workshop that was held September 26-29 1994 at Karlsruhe. For the Test Cell System 11 tasks were identified which can be grouped into the three major fields neutronics, test matrix/users and test cell engineering. In order to discuss recently achieved results and to coordinate necessary activities for an effective design integration, a technical workshop on the Test Cell System was initiated. This workshop was organized on July 3-6 1995 by the Institute for Materials Research I at the Forschungszentrum Karlsruhe and attended by 20 specialists working in the fields neutronics, fusion materials R and D and test cell engineering in the European Union, Japan, and the United States of America. The presentations and discussions during this workshop have shown together with the elaborated lists of action items, that has been achieved in all three fields, and that from the future IFMIF experimental program for a number of materials a database covering widerspread loading conditions up to DEMO-reactor relevant end-of-life damage levels can be expected. (orig.)

  2. Neutronics experiments, radiation detectors and nuclear techniques development in the EU in support of the TBM design for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Angelone, M., E-mail: maurizio.angelone@enea.it [ENEA UT-FUS C.R. Frascati, via E. Fermi, 45-00044 Frascati (Italy); Fischer, U. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Flammini, D. [ENEA UT-FUS C.R. Frascati, via E. Fermi, 45-00044 Frascati (Italy); Jodlowski, P. [AGH University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow (Poland); Klix, A. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Kodeli, I. [Jožef Stefan Institute, Ljubljana (Slovenia); Kuc, T. [AGH University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow (Poland); Leichtle, D. [Fusion for Energy, C/Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Lilley, S. [Culham Centre for Fusion Energy, Culham, OX14 3DB (United Kingdom); Majerle, M.; Novák, J. [Nuclear Physics Institute of the ASCR, Řež 130, 250 68 Řež (Czech Republic); Ostachowicz, B. [AGH University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow (Poland); Packer, L.W. [Culham Centre for Fusion Energy, Culham, OX14 3DB (United Kingdom); Pillon, M. [ENEA UT-FUS C.R. Frascati, via E. Fermi, 45-00044 Frascati (Italy); Pohorecki, W. [AGH University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow (Poland); Radulović, V. [Jožef Stefan Institute, Ljubljana (Slovenia); Šimečková, E. [Nuclear Physics Institute of the ASCR, Řež 130, 250 68 Řež (Czech Republic); and others

    2015-10-15

    Highlights: • A number of experiments and tests are ongoing to develop detectors and methods for HCLL and HCPM ITER-TBM. • Experiments for measuring gas production relevant to IFMIF are also performed using a cyclotron. • A benchmark experiment with a Cu block is performed to validate copper cross sections. • Experimental techniques to measure tritium in TBM are presented. • Experimental verification of activation cross sections for a Neutron Activation System for TBM is addressed. - Abstract: The development of high quality nuclear data, radiation detectors and instrumentation techniques for fusion technology applications in Europe is supported by Fusion for Energy (F4E) and conducted in a joint and collaborative effort by several European research associations (ENEA, KIT, JSI, NPI, AGH, and CCFE) joined to form the “Consortium on Nuclear Data Studies/Experiments in Support of TBM Activities”. This paper presents the neutronics activities carried out by the Consortium. A selection of available results are presented. Among then a benchmark experiment on a pure copper block to study the Cu cross sections at neutron energies relevant to fusion, the fabrication of prototype neutron detectors able to withstand harsh environment and temperature >200 °C (artificial diamond and self-powered detectors) developed for operating in ITER-TBM as well as measurement of relevant activation and integral gas production cross-sections. The latter measured at neutron energies relevant to IFMIF (>14 MeV) and the development of innovative experimental techniques for tritium measurement in TBM.

  3. International Fusion Materials Irradiation Facility injector acceptance tests at CEA/Saclay: 140 mA/100 keV deuteron beam characterization

    International Nuclear Information System (INIS)

    Gobin, R.; Bogard, D.; Chauvin, N.; Chel, S.; Delferrière, O.; Harrault, F.; Mattei, P.; Senée, F.; Cara, P.; Mosnier, A.; Shidara, H.; Okumura, Y.

    2014-01-01

    In the framework of the ITER broader approach, the International Fusion Materials Irradiation Facility (IFMIF) deuteron accelerator (2 × 125 mA at 40 MeV) is an irradiation tool dedicated to high neutron flux production for future nuclear plant material studies. During the validation phase, the Linear IFMIF Prototype Accelerator (LIPAc) machine will be tested on the Rokkasho site in Japan. This demonstrator aims to produce 125 mA/9 MeV deuteron beam. Involved in the LIPAc project for several years, specialists from CEA/Saclay designed the injector based on a SILHI type ECR source operating at 2.45 GHz and a 2 solenoid low energy beam line to produce such high intensity beam. The whole injector, equipped with its dedicated diagnostics, has been then installed and tested on the Saclay site. Before shipment from Europe to Japan, acceptance tests have been performed in November 2012 with 100 keV deuteron beam and intensity as high as 140 mA in continuous and pulsed mode. In this paper, the emittance measurements done for different duty cycles and different beam intensities will be presented as well as beam species fraction analysis. Then the reinstallation in Japan and commissioning plan on site will be reported

  4. Report of the second joint Research Committee for Fusion Reactor and Materials. July 12, 2002, Tokyo, Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    Joint research committees in purpose of the discussion on DEMO blanket in view point of the both of reactor technology and materials were held by the Research Committee for Fusion Reactor and Fusion Materials. The joint research committee was held in Tokyo on July 12, 2002. In the committee, the present status of development of solid and liquid breeding blanket, the present status of development of reduced activation structure materials, and IFMIF (International Fusion Materials Irradiation Facility) program were discussed based on the discussions of the development programs of the blanket and materials at the first joint research committee. As a result, it was confirmed that high electric efficiency with 41% would be obtained in the solid breeding blanket system, that neutron radiation data of reduced activation ferritic steel was obtained by HFIR collaboration, and that KEP (key element technology phase) of IFMIF would be finished at the end of 2002 and the data base for the next step, i.e. EVEDA (engineering validation/engineering design activity) was obtained. In addition, the present status of ITER CTA, which was a transient phase for the construction, and the outline of ITER Fast Track, which was an accelerated plan for the performance of the power plants, were reported. This report consists of the summary of the discussion and the viewgraphs which were used at the second joint research committee, and these are very useful for the researchers of the fusion area in Japan. (author)

  5. Technological and engineering challenges of fusion

    International Nuclear Information System (INIS)

    Maisonnier, David; Hayward, Jim

    2008-01-01

    The current fusion development scenario in Europe assumes the sequential achievement of key milestones. Firstly, the qualification of the DEMO/reactor physics basis in ITER, secondly, the qualification of materials for in-vessel components in IFMIF and, thirdly, the qualification of components and processes in DEMO. Although this scenario is constrained by budgetary considerations, it assumes the resolution of many challenges in physics, technology and engineering. In the first part of the paper, the technological and engineering challenges to be met in order to satisfy the current development scenario will be highlighted. These challenges will be met by an appropriate share of the work between ITER, IFMIF, DEMO and the necessary accompanying programme, which will have to include a number of dedicated facilities (e.g. for the development of H and CD systems). In the second part of the paper, the consequences of a considerable acceleration of the fusion development programme will be discussed. Although most of the technological and engineering challenges identified above will have to be met within a shorter timescale, it is possible to limit the requirements and expectation for a first fusion power plant with respect to those adopted for the current fusion development scenario. However, it must be recognised that such a strategy will inevitably result in increased risk and a reduction in the economy of the plant. (author)

  6. ANITA-IEAF activation code package - updating of the decay and cross section data libraries and validation on the experimental data from the Karlsruhe Isochronous Cyclotron

    Science.gov (United States)

    Frisoni, Manuela

    2017-09-01

    ANITA-IEAF is an activation package (code and libraries) developed in the past in ENEA-Bologna in order to assess the activation of materials exposed to neutrons with energies greater than 20 MeV. An updated version of the ANITA-IEAF activation code package has been developed. It is suitable to be applied to the study of the irradiation effects on materials in facilities like the International Fusion Materials Irradiation Facility (IFMIF) and the DEMO Oriented Neutron Source (DONES), in which a considerable amount of neutrons with energies above 20 MeV is produced. The present paper summarizes the main characteristics of the updated version of ANITA-IEAF, able to use decay and cross section data based on more recent evaluated nuclear data libraries, i.e. the JEFF-3.1.1 Radioactive Decay Data Library and the EAF-2010 neutron activation cross section library. In this paper the validation effort related to the comparison between the code predictions and the activity measurements obtained from the Karlsruhe Isochronous Cyclotron is presented. In this integral experiment samples of two different steels, SS-316 and F82H, pure vanadium and a vanadium alloy, structural materials of interest in fusion technology, were activated in a neutron spectrum similar to the IFMIF neutron field.

  7. Detailed mechanical design of the LIPAc beam dump radiological shielding

    Energy Technology Data Exchange (ETDEWEB)

    Nomen, Oriol, E-mail: onomen@irec.cat [IREC, Barcelona, Catalonia (Spain); CDEI-UPC, Barcelona, Catalonia (Spain); Martínez, José I.; Arranz, Fernando; Iglesias, Daniel; Barrera, Germán; Brañas, Beatriz [CIEMAT, Madrid (Spain); Ogando, Francisco [UNED, Madrid (Spain); Molla, Joaquín [CIEMAT, Madrid (Spain); Sanmartí, Manel [IREC, Barcelona, Catalonia (Spain)

    2013-10-15

    Highlights: ► Mechanical design of the IFMIF LIPAc beam dump shielding has been performed. ► Lead shutter design performed to shield radiation from beam dump when LIPAc is off. ► External loads, working and dismantling conditions, included as design constraints. -- Abstract: The LIPAc is a 9 MeV, D{sup +} linear prototype accelerator for the validation of the IFMIF accelerator design. The high intensity, 125 mA CW beam is stopped in a copper cone involving a high production of neutrons and gamma radiation and activation of its surface. The beam stopper is surrounded by a shielding to attenuate the resulting radiation so that dose rate values comply with the limits at the different zones of the installation. The shielding includes for that purpose polyethylene rings, water tanks and gray cast iron rings. A lead shutter has also been designed to shield the gamma radiation that comes through the beam tube when the linear accelerator is not in operation, in order to allow access inside the building for maintenance tasks. The present work summarizes the detailed mechanical design of the beam dump shielding and the lead shutter taking into account the design constraints, such as working conditions and other external loads, as well as including provisions for dismantling.

  8. Deuteron and neutron induced activation in the Eveda accelerator materials: implications for the accelerator maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, M.; Sanz, J.; Garcia, N.; Cabellos, O. [Madrid Univ. Politecnica, C/ Jose Gutierrez Abascal, lnstituto de Fusion Nuclear (Spain); Sauvan, R. [Universidad Nacional de Educacion a Distancia (UNED), Madrid (Spain); Moreno, C.; Sedano, L.A. [CIEMAT-Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, Association Euratom-CIEMAT, Madrid (Spain)

    2007-07-01

    Full text of publication follows: The IFMIF (International Fusion Materials Irradiation Facility) is an accelerator-based DLi neutron source designed to test fusion reactor candidate materials for high fluence neutrons. Before deciding IFMIF construction, an engineering design and associated experimental data acquisition, defined as EVEDA, has been proposed. Along the EVEDA accelerator, deuteron beam losses collide with the accelerator materials, producing activation and consequent radiations responsible of dose. Calculation of the dose rates in the EVEDA accelerator room is necessary in order to analyze the feasibility for manual maintenance. Dose rates due to the activation produced by the deuteron beam losses interaction with the accelerator materials, will be calculated with the ACAB activation code, using EAF2007 library for deuteron activation cross-sections. Also, dose rates from the activation induced by the neutron source produced by the interaction of deuteron beam losses with the accelerator materials and the deuterium implanted in the structural lattice, will be calculated with the SRIM2006, TMAP7, DROSG2000/NEUYIE, MCNPX and ACAB codes. All calculations will be done for the EVEDA accelerator with the room temperature DTL structure, which is based on copper cavities for the DTL. Some calculations will be done for the superconducting DTL structure, based on niobium cavities for the DTL working at cryogenic temperature. Final analysis will show the dominant mechanisms and major radionuclides contributing to the surface dose rates. (authors)

  9. Conceptual design of the liquid metal laboratory of the TECHNOFUSION facility

    International Nuclear Information System (INIS)

    Abánades, A.; García, A.; Casal, N.; Perlado, J.M.; Ibarra, A.

    2012-01-01

    Highlights: ► Conceptual design of a liquid Li facility. ► Components and cost estimation. ► Liquid metal laboratory into TEHNOFUSION proposal. - Abstract: The application of liquid metal technology in fusion devices requires R and D related to many phenomena: interaction between liquid metals and structural material as corrosion, erosion and passivation techniques; magneto-hydrodynamics; free surface fluid-dynamics and any other physical aspect that will be needed for their safe reliable operation. In particular, there is a significant shortage of experimental facilities dedicated to the development of the lithium technology. In the framework of the TECHNOFUSION project, an experimental laboratory devoted to the lithium technology development is proposed, in order to shed some light in the path to IFMIF and the design of chamber's first wall and divertors. The conceptual design foresee a development in two stages, the first one consisting on a material testing loop. The second stage proposes the construction of a mock-up of the IFMIF target that will allow to assess the behaviour of a free-surface lithium target under vacuum conditions. In this paper, such conceptual design is addressed.

  10. Linac design for intense hadron beams

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chuan

    2009-12-14

    Based on the RFQ and H-type DTL structures, this dissertation is dedicated to study the beam dynamics in the presence of significantly strong space-charge effects while accelerating intense hadron beams in the low- and medium-{beta} region. Besides the 5 mA/30 mA, 17 MeV proton injector (RFQ+DTL) and the 125 mA, 40 MeV deuteron DTL of the EUROTRANS and IFMIF facilities, a 200 mA, 700 keV proton RFQ has been also intensively studied for a small-scale but ultra-intense neutron source FRANZ planned at Frankfurt University. The most remarkable properties of the FRANZ RFQ and the IFMIF DTL are the design beam intensities, 200 mA and 125 mA. A new design approach, which can provide a balanced and accelerated beam bunching at low energy, has been developed for intense beams. To design the IFMIF DTL and the injector DTL part of the EUROTRANS driver linac, which have been foreseen as the first real applications of the novel superconducting CH-DTL structure, intensive attempts have been made to fulfill the design goals under the new conditions. For the IFMIF DTL, the preliminary IAP design has been considerably improved with respect to the linac layout as well as the beam dynamics. By reserving sufficient drift spaces for the cryosystem, diagnostic devices, tuner and steerer, introducing SC solenoid lenses and adjusting the accelerating gradients and accordingly other configurations of the cavities, a more realistic, reliable and efficient linac system has been designed. On the other hand, the specifications and positions of the transverse focusing elements as well as the phase- and energy-differences between the bunch-center particle and the synchronous particle at the beginning of the {phi}{sub s}=0 sections have been totally redesigned. For the EUROTRANS injector DTL, in addition to the above-mentioned procedures, extra optimization concepts to coordinate the beam dynamics between two intensities have been applied. In the beam transport simulations for both DTL designs

  11. Linac design for intense hadron beams

    International Nuclear Information System (INIS)

    Zhang, Chuan

    2009-01-01

    Based on the RFQ and H-type DTL structures, this dissertation is dedicated to study the beam dynamics in the presence of significantly strong space-charge effects while accelerating intense hadron beams in the low- and medium-β region. Besides the 5 mA/30 mA, 17 MeV proton injector (RFQ+DTL) and the 125 mA, 40 MeV deuteron DTL of the EUROTRANS and IFMIF facilities, a 200 mA, 700 keV proton RFQ has been also intensively studied for a small-scale but ultra-intense neutron source FRANZ planned at Frankfurt University. The most remarkable properties of the FRANZ RFQ and the IFMIF DTL are the design beam intensities, 200 mA and 125 mA. A new design approach, which can provide a balanced and accelerated beam bunching at low energy, has been developed for intense beams. To design the IFMIF DTL and the injector DTL part of the EUROTRANS driver linac, which have been foreseen as the first real applications of the novel superconducting CH-DTL structure, intensive attempts have been made to fulfill the design goals under the new conditions. For the IFMIF DTL, the preliminary IAP design has been considerably improved with respect to the linac layout as well as the beam dynamics. By reserving sufficient drift spaces for the cryosystem, diagnostic devices, tuner and steerer, introducing SC solenoid lenses and adjusting the accelerating gradients and accordingly other configurations of the cavities, a more realistic, reliable and efficient linac system has been designed. On the other hand, the specifications and positions of the transverse focusing elements as well as the phase- and energy-differences between the bunch-center particle and the synchronous particle at the beginning of the φ s =0 sections have been totally redesigned. For the EUROTRANS injector DTL, in addition to the above-mentioned procedures, extra optimization concepts to coordinate the beam dynamics between two intensities have been applied. In the beam transport simulations for both DTL designs, no beam

  12. Design and analysis of the vacuum vessel for RTO/RC-ITER

    International Nuclear Information System (INIS)

    Onozuka, M.; Ioki, K.; Johnson, G.; Kodama, T.; Sannazzaro, G.; Utin, Y.

    2000-01-01

    Recent progress in design and analysis of the vacuum vessel (VV) for the reduced technical objectives/reduced cost International Thermonuclear Experimental Reactor (RTO/RC-ITER) is presented. The basic VV design is similar to the previous ITER VV. However, because the back plate for the blanket modules could be eliminated, its previous functions could be transferred to the VV. For this option, the blanket modules are supported directly by the VV and the blanket coolant channels are structurally part of the VV double wall structure. In addition, a 'tight fitting' configuration is required to correctly position the modules' first wall. Although such modifications of the VV complicate its structure and increase its fabrication cost, the design of the VV is considered to be still feasible. The structural analyses of the VV have been conducted using several FE models of the VV, including global and local models. Although further assessment is required, based on the analyses performed to date, the structural aspects of the VV for the case without the back plate appear feasible

  13. Engineering and thermal-hydraulics design of PFC cooling for SST-1 Tokamak

    International Nuclear Information System (INIS)

    Chaudhuri, Paritosh; Reddy, D. Chenna; Santra, P.; Khiwadkar, S.; Prakash, N. Rabi; Ramash, G.; Dubey, Santosh; Prakash, Arun; Saxena, Y. C.

    2003-01-01

    The main consideration in the design of the PFC cooling for SST-1 tokamak is the steady state heat removal of upto 1MW/m2. The PFC also has been design to withstand the peak heat fluxes without significant erosion such that frequent replacement is not necessary. Proper brazing of cooling tube on the copper back plate is necessary for the efficient heat transfer from the tube to the back plate. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during baking, and cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, and tile fitting mechanism. Finite Element (FE) models using ANSYS have been developed to conduct the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The temperature distribution results for different PFC obtained by FE results were assessed by comparison with 2-D Finite Difference code. The results of the calculation led to a good understanding of the coolant flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. The contact at the brazed joint of the tube to the backplate is critical for the above application. The manufactured modules need to be evaluated for the quality of brazed joint. Using an infra-red-camera, spatial and temporal evaluation of the temperature profile is studied under various flow parameters. These results of this study will be presented in details in this paper

  14. Structural analysis of vacuum vessel and blanket support system for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Koizumi, Kouichi; Takatsu, Hideyuki; Tada, Eisuke; Shimane, Hideo.

    1996-11-01

    Structural analyses of vacuum vessel and blanket support system have been performed to examine their integrated structural behavior under the design loads and to assess their structural feasibility, with two kinds of three-dimensional (3-D) FEM models; a detailed model with 18deg sector region to investigate the detailed mechanical behaviors of the blanket and vessel components under the several symmetric loads, and a 180deg torus model with relatively coarser meshes to assess the structural responses under the asymmetric VDE load. The analytical results obtained by both models were also compared for the several symmetric loads to check the equivalent mechanical stiffness of the 180deg torus model. As the results, most of the vessel and blanket components have sufficient mechanical integrities with the stress level below the allowable limit of the materials, while the lower parts of inboard/outboard back plate need to be reinforced by increasing the thickness and/or mounting a toroidal ring support at the lower edge of the back plate. Two types of eigenvalue analyses were also conducted with the 180deg torus model to investigate natural frequencies of the vessel torus support system and to assess the mechanical integrity of the elastic stability under the asymmetric VDE load. Analytical results show that the mechanical stiffness of the vessel gravity support should be higher in the view point of a seismic response, and that those of the blanket support structures should also be increased for the buckling strength against the VDE vertical force. (author)

  15. Investigation of factors affecting the calibration of strain gage based transducers (''Goodzeit gages'') for SSC magnets

    International Nuclear Information System (INIS)

    Davidson, M.; Gilbertson, A.; Dougherty, M.

    1991-03-01

    These transducers are designed to measure stresses on SSC collared coils. They are individually calibrated with a bonded ten-stack of SSC inner coil cable by applying a known load and reading corresponding output from the gages. The transducer is supported by a notched ''backing plate'' that allows for bending of the gage beam during calibration or in use with an actual coil. Several factors affecting the calibration and use of the transducers are: the number of times a ''backing plate'' is used, the similarities or difficulties between bonded ten-stacks, and the differences between the ten-stacks and the coil they represent. The latter is probably the most important because a calibration curve is a model of how a transducer should react within a coil. If the model is wrong, the calibration curve is wrong. Information will be presented regarding differences in calibrations between Brookhaven National Labs (also calibrating these transducers) and Fermilab -- what caused these differences, the investigation into the differences between coils and ten-stacks and how they relate to transducer calibration, and some suggestions for future calibrations

  16. Design and analysis of the vacuum vessel for RTO/RC-ITER

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. E-mail: onozukm@itereu.de; Ioki, K.; Johnson, G.; Kodama, T.; Sannazzaro, G.; Utin, Y

    2000-11-01

    Recent progress in design and analysis of the vacuum vessel (VV) for the reduced technical objectives/reduced cost International Thermonuclear Experimental Reactor (RTO/RC-ITER) is presented. The basic VV design is similar to the previous ITER VV. However, because the back plate for the blanket modules could be eliminated, its previous functions could be transferred to the VV. For this option, the blanket modules are supported directly by the VV and the blanket coolant channels are structurally part of the VV double wall structure. In addition, a 'tight fitting' configuration is required to correctly position the modules' first wall. Although such modifications of the VV complicate its structure and increase its fabrication cost, the design of the VV is considered to be still feasible. The structural analyses of the VV have been conducted using several FE models of the VV, including global and local models. Although further assessment is required, based on the analyses performed to date, the structural aspects of the VV for the case without the back plate appear feasible.

  17. Experimental and numerical results from hybrid retrofitted photovoltaic panels

    International Nuclear Information System (INIS)

    Rossi, Cecilia; Tagliafico, Luca A.; Scarpa, Federico; Bianco, Vincenzo

    2013-01-01

    Highlights: • The experimental study focuses on the feasibility of hybrid PV/T panels retrofitting. • The critical role of a thin layer of air between PV panel and back plate is evidenced. • The benefit of the addition of a conductive paste layer is analyzed via FEM simulations. • The use of wood ribs to stick the back plate represents a cheap effective solution. - Abstract: The aim of present study is to investigate different methodologies to achieve a better contact between a photovoltaic panel and a thermal plate, in order to cool the PV panel by means of water in the perspective of coupling it with a heat pump. It is believed that this kind of system allows to obtain a higher energy efficiency. The analysis is developed both experimentally and numerically, testing different kinds of configurations in different operating conditions. Simulations are employed to analyze the effect of the variations of the contact resistance between the panel and the thermal plates, demonstrating that the use of a conductive paste increases the overall performance of the panel. Results show interesting possibilities in terms of retrofitting of existing photovoltaic panels by employing very simple solutions, such as to fix the thermal plate on the rear of the panel by means of wood ribs

  18. Electron backstream to the source plasma region in an ion source

    International Nuclear Information System (INIS)

    Ohara, Y.; Akiba, M.; Arakawa, Y.; Okumura, Y.; Sakuraba, J.

    1980-01-01

    The flux of backstream electrons to the source plasma region increases significantly with the acceleration voltage of an ion beam, so that the back plate in the arc chamber should be broken for quasi-dc operation. The flux of backstream electrons is estimated at the acceleration voltage of 50--100 kV for a proton beam with the aid of ion beam simulation code. The power flux of backstream electrons is up to about 7% of the total beam output at the acceleration voltage of 75 kV. It is pointed out that the conventional ion sources such as the duoPIGatron or the bucket source which use a magnetic field for source plasma production are not suitable for quasi-dc and high-energy ion sources, because the surface heat flux of the back plate is increased by the focusing of backstream electrons and the removal of it is quite difficult. A new ion source which has an electron beam dump in the arc chamber is proposed

  19. Engineering and thermal-hydraulic design of water cooled PFC for SST-1 tokamak

    International Nuclear Information System (INIS)

    Paritosh Chaudhuri; Santra, P.; Rabi Prakash, N.; Khirwadkar, S.; Arun Prakash, A.; Ramash, G.; Dubey, S.; Chenna Reddy, D.; Saxena, Y.C.

    2005-01-01

    Full text of publication follows: Steady state Superconducting Tokamak (SST-1) is a medium size tokamak with superconducting magnetic field coils. It is a large aspect ratio tokamak with a major radius of 1.1 m and minor radius of 0.20 m. SST-1 is designed for plasma discharge duration of ∼1000 seconds to obtain fully steady state plasma with total input power up to 1.0 MW. First Wall or Plasma Facing Components (PFC) is one or the major sub-systems of SST-1 tokamak consisting of divertors, passive stabilizers, baffles, and poloidal limiters are designed to be compatible for steady state operation. All the PFC has the same basic configuration: graphite tiles are mechanically attached to a back plate made of high strength copper alloy, and SS tubes are embedded in the groove made in the back plate. Same tube will be used for cooling during plasma operation and baking during wall conditioning. The main consideration in the design of the PFC is the steady state heat removal of up to 1 MW/m 2 . In addition to remove high heat fluxes, the PFC are also designed to be compatible for high temperature baking at 350 deg. C. Water was chosen as the coolant because of its appropriate thermal properties, and while baking, hot nitrogen gas would flow through these tubes to bake the PFC at high temperature. Extensive studies, involving different flow parameters and various cooling layouts, has been done to select the final cooling parameters and layout, compatible for cooling and baking. During steady state operation, divertor and passive stabilizer heat loads are expected to be 0.6 and 0.25 MW/m 2 . The PFC also has been design to withstand the peak heat fluxes without significant erosion such that frequent replacement is not necessary. Since the tile must be mechanically attached to the back plate (heat sink), the fitting technique must provide the highest mechanical stress so that thermal transfer efficiency is maximized. Proper brazing of cooling tube on the copper back

  20. A new BETSI test bench at CEA/Saclay

    International Nuclear Information System (INIS)

    Nyckees, S.; Adroit, G.; Delferriere, O.; Duperrier, R.; Gauthier, Y.; Gobin, R.; Harrault, F.; Mateo, C.M.; Napoly, O.; Pottin, B.; Sauce, Y.; Senee, F.; Tuske, O.; Vacher, T.

    2012-01-01

    In the nineties, CEA has undertaken to develop the production of high intensity light ion beams from plasma generated by electron cyclotron resonance (ECR). Important results were obtained with the SILHI source in pulsed or continuous mode. Presently, CEA/Saclay is now involved in the construction of different injectors dedicated to large infrastructures like IFMIF or SPIRAL2. Other installations are also interested by high intensity ion sources like ESS or FAIR. To improve and test new sources, a new test bench named BETSI (Banc d'Etudes et de Tests des Sources d'Ions) has been operating for several years. Low energy beam line diagnostics consist of a Faraday cup, cameras and a species analyzer. The SILHI emittance scanner can also be installed on the beam line. On this test bench, different permanent magnet source configurations are tested. The paper is followed by the associated poster. (authors)

  1. Fusion neutronics plan in the development of fusion reactor. With the aim of realizing electric power

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Hiroo; Morimoto, Yuichi; Ochiai, Kentarou; Sugimoto, Masayoshi; Nishitani, Takeo; Takeuchi, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    On June 1992, Atomic Energy Commission in Japan has settled Third Phase Program of Fusion Research and Development to achieve self-ignition condition, to realize long pulse burning plasma and to establish basis of fusion engineering for demonstration reactor. This report describes research plan of Fusion Neutron Laboratory in JAERI toward a development of fusion reactor with an aim of realizing electric power. The fusion neutron laboratory has a fusion neutronics facility (FNS), intense fusion neutron source. The plan includes research items in the FNS; characteristics of shielding and breeding materials, nuclear characteristics of materials, fundamental irradiation process of insulator, diagnostics materials and structural materials, and development of in-vessel diagnostic technology. Upgrade of the FNS is also described. Also, the International Fusion Material Irradiation Facility (IFMIF) for intense neutron source to develop fusion materials is described. (author)

  2. Uncertainty and sensitivity analysis on probabilistic safety assessment of an experimental facility

    International Nuclear Information System (INIS)

    Burgazzi, L.

    2000-01-01

    The aim of this work is to perform an uncertainty and sensitivity analysis on the probabilistic safety assessment of the International Fusion Materials Irradiation Facility (IFMIF), in order to assess the effect on the final risk values of the uncertainties associated with the generic data used for the initiating events and component reliability and to identify the key quantities contributing to this uncertainty. The analysis is conducted on the expected frequency calculated for the accident sequences, defined through the event tree (ET) modeling. This is in order to increment credit to the ET model quantification, to calculate frequency distributions for the occurrence of events and, consequently, to assess if sequences have been correctly selected on the probability standpoint and finally to verify the fulfillment of the safety conditions. Uncertainty and sensitivity analysis are performed using respectively Monte Carlo sampling and an importance parameter technique. (author)

  3. Thermonuclear controlled fusion: international cooperation

    International Nuclear Information System (INIS)

    Conscience, J.-F.

    2001-01-01

    This report summarizes the current worldwide status of research in the field of thermonuclear controlled fusion as well as the international research programme planed for the next decades. The two main projects will be the ITER facility (International Thermonuclear Experimental Reactor) that should produce 10 times more energy than the energy injected, and the IFMIF (International Fusion Materials Irradiation Facility) designed to study the reactions of materials under intense neutron fluxes. The future of the pioneering JET facility (Joint European Torus) is also discussed. The engagement of the various countries (USA, Japan, Germany, Russian Federation and Canada) and international organisations (EURATOM and IEA) in terms of investment and research is described. Switzerland is involved in this program through an agreement with EURATOM and is mainly dedicated to experimental studies with the TCV machine in Lausanne and numerical studies of plasma configurations. It will participate to the development of the microwave plasma heating system for the ITER machine

  4. Recent progress in reduced activation ferritic steels R and D in Japan

    International Nuclear Information System (INIS)

    Kimura, A.; Kohyama, A.; Sawai, T.; Shiba, K.; Hishinuma, A.; Jitsukawa, S.; Ukai, S.

    2003-01-01

    The Japanese RAFSs R and D road map toward DEMO is shown. Important steps include high-dose irradiation by fission reactors, such as HFIR in ORNL, irradiation tests by 14 MeV neutrons in IFMIF and application to ITER test blanket modules to provide an adequate database of RAFS for the design of DEMO. Current status of RAFS development is also introduced. The major properties of concern are well within our knowledge and process technologies are mostly ready for fusion application. The RAFSs are now certainly ready to proceed to the next stage. Material database is already in hand and further progress is anticipated for the design of ITER test blanket. Oxide Dispersion strengthening (ODS) steels are quite promising for high temperature operation of the blanket system with potential improvements in radiation resistance of mechanical performances and of corrosion. (author)

  5. Tools for simulation of high beam intensity ion accelerators; Simulationswerkzeuge fuer die Berechnung hochintensiver Ionenbeschleuniger

    Energy Technology Data Exchange (ETDEWEB)

    Tiede, Rudolf

    2009-07-09

    A new particle-in-cell space charge routine based on a fast Fourier transform was developed and implemented to the LORASR code. It provides the ability to perform up to several 100 batch run simulations with up to 1 million macroparticles each within reasonable computation time. The new space charge routine was successfully validated in the framework of the European ''High Intensity Pulsed Proton Injectors'' (HIPPI) collaboration: Several static Poisson solver benchmarking comparisons were performed, as well as particle tracking comparisons along the GSI UNILAC Alvarez section. Moreover machine error setting routines and data analysis tools were developed and applied on error studies for the ''Heidelberg Cacer Therapy'' (HICAT) IH-type drift tube linear accelerator (linac), the FAIR Facility Proton Linac and the proposal of a linac for the ''International Fusion Materials Irradiation Facility'' (IFMIF) based on superconducting CH-type structures. (orig.)

  6. Neutronics comparisons of d-Li and t-H2O neutron sources

    International Nuclear Information System (INIS)

    Doran, D.G.; Cierjacks, S.; Mann, F.M.; Greenwood, L.R.; Daum, E.

    1995-01-01

    Calculations were performed to compare the neutronics of two neutron source concepts which are candidates for an international fusion materials irradiation facility (IFMIF). One concept, d-Li, produces neutrons by stopping 35 MeV deuterons in a flowing lithium target. Criticism of this concept because of the high energy tail above 14 MeV gave rise to the t-H 2 O concept proposed by Cierjacks. It would generate neutrons below 14.6 MeV ( 2 O. Test volumes that met certain damage parameter criteria were estimated. Because of the softer spectra and somewhat lower yields for t-H 2 O, the d-Li concept was found to have a test volume advantage of a factor of 2 or more, depending on the material to be irradiated. ((orig.))

  7. Determination of neutron spectra formed by 40-MeV deuteron bombardment of a lithium target with multi-foil activation technique

    CERN Document Server

    Maekawa, F; Wada, M; Wilson, P P H; Ikeda, Y

    2000-01-01

    Neutron flux spectra at an irradiation field produced by a 40-MeV deuteron bombardment on a thick lithium-target at Forschungszentrum Karlsruhe, Germany, have been determined by the multi-foil activation technique. Twenty-seven dosimetry reactions having a wide energy range of threshold energies up to 38 MeV were employed as detectors for the neutron flux spectra extending to 55 MeV. The spectra were adjusted with the SAND-II code with the experimental reaction rates based on an iterative method. The adjusted spectra validated quantitatively the Monte Carlo deuteron-lithium (d-Li) neutron source model code (M sup C DeLi) which was used to calculate initial guess spectra and also has been used for IFMIF nuclear designs. Accuracy of the adjusted spectra was approx 10% that was suitable for successive integral tests of activation cross section data.

  8. Space-Charge Effect

    International Nuclear Information System (INIS)

    Chauvin, N

    2013-01-01

    First, this chapter introduces the expressions for the electric and magnetic space-charge internal fields and forces induced by high-intensity beams. Then, the root-mean-square equation with space charge is derived and discussed. In the third section, the one-dimensional Child-Langmuir law, which gives the maximum current density that can be extracted from an ion source, is exposed. Space-charge compensation can occur in the low-energy beam transport lines (located after the ion source). This phenomenon, which counteracts the spacecharge defocusing effect, is explained and its main parameters are presented. The fifth section presents an overview of the principal methods to perform beam dynamics numerical simulations. An example of a particles-in-cells code, SolMaxP, which takes into account space-charge compensation, is given. Finally, beam dynamics simulation results obtained with this code in the case of the IFMIF injector are presented. (author)

  9. Space-Charge Effect

    CERN Document Server

    Chauvin, N.

    2013-12-16

    First, this chapter introduces the expressions for the electric and magnetic space-charge internal fields and forces induced by high-intensity beams. Then, the root-mean-square equation with space charge is derived and discussed. In the third section, the one-dimensional Child-Langmuir law, which gives the maximum current density that can be extracted from an ion source, is exposed. Space-charge compensation can occur in the low-energy beam transport lines (located after the ion source). This phenomenon, which counteracts the spacecharge defocusing effect, is explained and its main parameters are presented. The fifth section presents an overview of the principal methods to perform beam dynamics numerical simulations. An example of a particles-in-cells code, SolMaxP, which takes into account space-charge compensation, is given. Finally, beam dynamics simulation results obtained with this code in the case of the IFMIF injector are presented.

  10. Fusion - 2050 perspective (in Polish)

    CERN Document Server

    Romaniuk, R S

    2013-01-01

    The results of strongly exothermic reaction of thermonuclear fusion between nuclei of deuterium and tritium are: helium nuclei and neutrons, plus considerable kinetic energy of neutrons of over 14 MeV. DT nuclides synthesis reaction is probably not the most favorable one for energy production, but is the most advanced technologically. More efficient would be possibly aneutronic fusion. The EU by its EURATOM agenda prepared a Road Map for research and implementation of Fusion as a commercial method of thermonuclear energy generation in the time horizon of 2050.The milestones on this road are tokomak experiments JET, ITER and DEMO, and neutron experiment IFMIF. There is a hope, that by engagement of the national government, and all research and technical fusion communities, part of this Road Map may be realized in Poland. The infrastructure build for fusion experiments may be also used for material engineering research, chemistry, biomedical, associated with environment protection, power engineering, security, ...

  11. Dynamic mechanical properties of reduced activation ferritic steels

    International Nuclear Information System (INIS)

    Hirose, T.; Kohyama, A.; Tanigawa, H.; Ando, M.; Jitsukawa, S.

    2003-01-01

    A fatigue test method by a miniaturized hourglass-shaped fatigue specimen has been developed for International Fusion Materials Irradiation Facility (IFMIF) and sufficient potential as the alternative to a conventional large specimen was presented. Furthermore, focused ion beam micro- sampling method was successfully applied to microstructural analysis on fracture process. Where, the effects of displacement damage and transmutation helium on the fatigue properties of Reduced Activation Ferritic/Martensitic Steels, RAFs, were investigated. Neutron irradiation and helium-ion-implantation at ambient temperature caused radiation hardening to degrade fatigue lifetime of F82H steel. Microstructural analysis revealed that local brittle fractures occurred at early stage of fatigue tests was the origin of the degradation.. No significant difference in fatigue life degradation was detected with and without implanted helium. This result suggests that 100 appm helium implanted has no impact on fracture life time under neutron irradiation. (author)

  12. High current ion source development at Frankfurt

    Energy Technology Data Exchange (ETDEWEB)

    Volk, K.; Klein, H.; Lakatos, A.; Maaser, A.; Weber, M. [Frankfurt Univ. (Germany). Inst. fuer Angewandte Physik

    1995-11-01

    The development of high current positive and negative ion sources is an essential issue for the next generation of high current linear accelerators. Especially, the design of the European Spallation Source facility (ESS) and the International Fusion Material Irradiation Test Facility (IFMIF) have increased the significance of high brightness hydrogen and deuterium sources. As an example, for the ESS facility, two H{sup -}-sources each delivering a 70 mA H{sup -}-beam in 1.45 ms pulses at a repetition rate of 50 Hz are necessary. A low emittance is another important prerequisite. The source must operate, while meeting the performance requirements, with a constancy and reliability over an acceptable period of time. The present paper summarizes the progress achieved in ion sources development of intense, single charge, positive and negative ion beams. (author) 16 figs., 7 refs.

  13. High current ion source development at Frankfurt

    International Nuclear Information System (INIS)

    Volk, K.; Klein, H.; Lakatos, A.; Maaser, A.; Weber, M.

    1995-01-01

    The development of high current positive and negative ion sources is an essential issue for the next generation of high current linear accelerators. Especially, the design of the European Spallation Source facility (ESS) and the International Fusion Material Irradiation Test Facility (IFMIF) have increased the significance of high brightness hydrogen and deuterium sources. As an example, for the ESS facility, two H - -sources each delivering a 70 mA H - -beam in 1.45 ms pulses at a repetition rate of 50 Hz are necessary. A low emittance is another important prerequisite. The source must operate, while meeting the performance requirements, with a constancy and reliability over an acceptable period of time. The present paper summarizes the progress achieved in ion sources development of intense, single charge, positive and negative ion beams. (author) 16 figs., 7 refs

  14. Accelerator conceptual design of the international fusion materials irradiation facility

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, M.; Kinsho, M. [Japan Atomic Energy Res. Inst., Tokai, Ibaraki (Japan). Intense Neutron Source Lab.; Jameson, R.A.; Blind, B. [Los Alamos National Lab., NM (United States); Teplyakov, V. [Institute for High Energy Physics, Moscow (Russian Federation); Berwald, D.; Bruhwiler, D.; Peakock, M.; Rathke, J. [Northrop Grumman Corp., Bethpage, NY (United States); Deitinghoff, H.; Klein, H.; Pozimski, Y.; Volk, K. [Johann Wolfgang Goethe Univ., Frankfurt (Germany). Inst. fur Angewandte Phys.; Ferdinand, R.; Lagniel, J.-M. [CEA Saclay LNS, Gif-sur-Yvette (France); Miyahara, A. [Teikyo Univ., Tokyo (Japan); Olivier, M. [CEA DSM, Saclay, Gif-sur-Yvette (France); Piechowiak, E. [Northrop Grumman Corp., Baltimore, MD (United States); Tanabe, Y. [Toshiba Corp., Tsurumi-ku, Yokohama (Japan)

    1998-10-01

    The accelerator system of the international fusion materials irradiation facility (IFMIF) provides the 250-mA, 40-MeV continuous-wave deuteron beam at one of the two lithium target stations. It consists of two identical linear accelerator modules, each of which independently delivers a 125-mA beam to the common footprint of 20 cm x 5 cm at the target surface. The accelerator module consists of an ion injector, a 175 MHz RFQ and eight DTL tanks, and rf power supply system. The requirements for the accelerator system and the design concept are described. The interface issues and operational considerations to attain the proposed availability are also discussed. (orig.) 8 refs.

  15. Accelerator conceptual design of the international fusion materials irradiation facility

    International Nuclear Information System (INIS)

    Sugimoto, M.; Kinsho, M.; Teplyakov, V.; Berwald, D.; Bruhwiler, D.; Peakock, M.; Rathke, J.; Deitinghoff, H.; Klein, H.; Pozimski, Y.; Volk, K.; Miyahara, A.; Olivier, M.; Piechowiak, E.; Tanabe, Y.

    1998-01-01

    The accelerator system of the international fusion materials irradiation facility (IFMIF) provides the 250-mA, 40-MeV continuous-wave deuteron beam at one of the two lithium target stations. It consists of two identical linear accelerator modules, each of which independently delivers a 125-mA beam to the common footprint of 20 cm x 5 cm at the target surface. The accelerator module consists of an ion injector, a 175 MHz RFQ and eight DTL tanks, and rf power supply system. The requirements for the accelerator system and the design concept are described. The interface issues and operational considerations to attain the proposed availability are also discussed. (orig.)

  16. Calculation of damage function of Al{sub 2}O{sub 3} in irradiation facilities for fusion reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Mota, F., E-mail: fernando.mota@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Ortiz, C.J., E-mail: christophe.ortiz@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Vila, R., E-mail: rafael.vila@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Casal, N., E-mail: natalia.casal@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); García, A., E-mail: angela.garcia@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Ibarra, A., E-mail: Angel.ibarra@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain)

    2013-11-15

    A rigorous material testing program is essential for the development of the nuclear fusion world program. In particular, it is very important to predict the generation of the displacement damage in materials, because the irradiation intensity expected in fusion conditions is such that the performance of materials and components under these extreme conditions is unknown. To study the damage produced by neutrons in materials of interest for fusion, a specific computational methodology was developed. Neutron fluxes expected in different irradiation facilities (International Fusion Materials Irradiation Facility [IFMIF] and DEMO-HCLL) and in different irradiation spots were obtained with particles transport codes (McDeLicious, MCNP). The energy differential cross sections of primary knock-on atoms were calculated using the NJOY code. Resulting data were input into the Monte Carlo code MARLOWE to calculate the corresponding displacements (i.e., interstitials (I) and vacancies (V)). However, the number of Frenkel pairs created during irradiation strongly depends on the recombination radius between interstitials and vacancies. This parameter corresponds to the minimum distance below which instantaneous recombination occurs. Mainly, the influence of such parameter on the damage function in Al{sub 2}O{sub 3} was assessed in this report. The displacements per atom values calculated as a function of the recombination radius considered are compared to experimental data to determine the most appropriate capture radius. In addition, the damage function and damage dose generated at different experimental irradiation facilities are compared with those expected in DEMO. The conclusion is that both IFMIF and TechnoFusión (future triple beam ion accelerator to emulate fusion neutron irradiation effects in materials) facilities are suited to perform relevant irradiation experiments for the design of DEMO.

  17. An assessment of the evaporation and condensation phenomena of lithium during the operation of a Li(d,xn fusion relevant neutron source

    Directory of Open Access Journals (Sweden)

    J. Knaster

    2016-12-01

    Full Text Available The flowing lithium target of a Li(d,xn fusion relevant neutron source must evacuate the deuteron beam power and generate in a stable manner a flux of neutrons with a broad peak at 14 MeV capable to cause similar phenomena as would undergo the structural materials of plasma facing components of a DEMO like reactors. Whereas the physics of the beam-target interaction are understood and the stability of the lithium screen flowing at the nominal conditions of IFMIF (25 mm thick screen with +/–1 mm surface amplitudes flowing at 15 m/s and 523 K has been demonstrated, a conclusive assessment of the evaporation and condensation of lithium during operation was missing. First attempts to determine evaporation rates started by Hertz in 1882 and have since been subject of continuous efforts driven by its practical importance; however intense surface evaporation is essentially a non-equilibrium process with its inherent theoretical difficulties. Hertz-Knudsen-Langmuir (HKL equation with Schrage’s ‘accommodation factor’ η = 1.66 provide excellent agreement with experiments for weak evaporation under certain conditions, which are present during a Li(d,xn facility operation. An assessment of the impact under the known operational conditions for IFMIF (574 K and 10−3Pa on the free surface, with the sticking probability of 1 inherent to a hot lithium gas contained in room temperature steel walls, is carried out. An explanation of the main physical concepts to adequately place needed assumptions is included.

  18. Hazard evaluation of The International Fusion Materials Irradiation Facility

    Energy Technology Data Exchange (ETDEWEB)

    Burgazzi, Luciano [ENEA-Centro Ricerche ' Ezio Clementel' , Advanced Physics Technology Division, Via Martiri di Monte Sole, 4, 40129 Bologna (Italy)]. E-mail: burgazzi@bologna.enea.it

    2005-01-15

    The International Fusion Materials Irradiation Facility (IFMIF) is aimed to provide an intense neutron source by a high current deuteron linear accelerator and a high-speed lithium flow target, for testing candidate materials for fusion. Liquid lithium is being circulated through a loop and is kept at a temperature above its freezing point. In the frame of the design phase called Key Element technology Phase (KEP), jointly performed by an international team to verify the most important risk factors, safety assessment of the whole plant has been required in order to identify the hazards associated with the plant operation. This paper discusses the safety assessments that were performed and their outcome: Failure Mode and Effect Analysis (FMEA) approach has been adopted in order to accomplish the task. Main conclusions of the study is that, on account of the safety and preventive measures adopted, potential plant related hazards are confined within the IFMIF security boundaries and great care must be exercised to protect workers and site personnel from operating the plant. The analysis has provided as a result a set of Postulated Initiating Events (PIEs), that is off-normal events, that could result in hazardous consequences for the plant, together with the total frequency and the list of component failures which could induce the PIE: this assures the exhaustive list of major initiating events of accident sequences, helpful to the further accident sequence analysis phase. Finally, for each one of the individuated PIEs, the evaluation of the accident evolution, in terms of effects on the plant and relative countermeasures, has allowed to verify that adequate measures are being taken both to prevent the accident occurrence and to cope with the accident consequences, thus assuring the fulfilment of the safety requirements.

  19. The EU/JA Broader Approach Activities

    International Nuclear Information System (INIS)

    Shinzaburo Matsuda

    2006-01-01

    At the time of ITER site decision in Moscow on 28 June 2005, representatives of EU and Japan jointly declared their intention to implement Broader Approach Activities in support of ITER on a time frame compatible with its construction phase. On the basis of this declaration, working groups from the Parties were established to identify possible key areas of joint activities. As a result of intense discussions, agreement has been reached by both Parties as follows. The Broader Approach Activities comprise the following three projects: 1) Engineering Validation and Engineering Design Activities for the International Fusion Materials Irradiation Facility (IFMIF/EVEDA) 2) the International Fusion Energy Research Center (IFERC), comprising: a) A DEMO Design, R (and) D coordination Center aiming at establishing a common basis for a DEMO design, b) A Computational Simulation Center composed of super-computer facilities for large scale simulation activities, and c) An ITER Remote Experimentation Center to facilitate broad participation of scientists into ITER experiments. 3) the Satellite Tokamak Programme including participation in the upgrade of JT-60 Tokamak to an advanced superconducting tokamak and participation in its exploitation, to support ITER and research towards DEMO. The Parties shall establish a Steering Committee responsible for the overall direction and supervision of the activities. Each project is lead by the respective Project Leader supported by the Project Team. A Project Committee is established for each project to make recommendations to the Steering Committee and monitor the progress of the project. Each Party shall nominate an agency to discharge its obligations for the implementation of these projects. Resources for the Broader Approach Activities shall be equally shared by EU and Japan, contributed mostly in-kind, and allocation of procurements, tasks and responsibilities have been identified. The IFMIF/EVEDA and IFERC projects will be implemented

  20. Primary Displacement Damage Calculation Induced by Neutron and Ion Using Binary Collision Approximation Techniques (Marlowe Code)

    International Nuclear Information System (INIS)

    Mota, F.; Ortiz, C. J.; Vila, R.

    2012-01-01

    The level of damage expected in future fusion reactors conditions is such that the performance of materials and components under these extreme irradiation conditions is still unknown. Considering this scenario, the study of the effects of energetic neutrons generated in fusion reactors on materials is one of the most important research topics to be carried out during next years. The effects of neutron irradiation on materials involve, from a fundamental point of view, two physical phenomena: i) the displacement of atoms from their equilibrium positions in the lattice, which creates point defects, and ii) the generation of nuclear transmutation reactions that contribute to the formation of impurities inside the material, with He and H as the most important ones. The ratio between the levels of He and H, and the amount of point defects is one of the main parameters to understand the effect of the radiation on materials. In order to emulate the neutron irradiation that would prevail under fusion conditions, two approaches are contemplated: a) on one hand different kinds of current neutron sources to emulate the fusion irradiation environment are available, as for example - Fission power reactor - Spallation sources - Striping Sources: The objective of the International Fusion Materials Irradiation Facility (IFMIF) will be to provide an intense neutron source with adequate energy spectrum to test the suitability of candidate materials for future nuclear fusion power reactor (DEMO). IFMIF will constitute an essential tool in the international strategy towards the achievement of future fusion reactors. b) on the other hand, as these neutron sources have a number of problems and very strict operating conditions, (e.g. the radiological risks), to emulate the effects of fusion neutron on materials, some other facilities can be used. One example is the Spanish initiative TechnoFusion facility which purpose is to serve as technological support for IFMIF and DEMO. The Material

  1. Optimization of thermal design for nitrogen shield of JET cryopump

    International Nuclear Information System (INIS)

    Baxi, C.B.; Obert, W.

    1991-11-01

    The reference design of JET cryopump nitrogen shield consists of an outer section made of copper chevrons fastened to two cooling tubes and an inner stainless steel section and backing plate with two cooling tubes. These tubes are fed in a parallel flow arrangement. The inlet flow is divided into two parallel paths so that both tubes on either section are always at the same temperature. This arrangement was selected due to concern about conduction between warm and cold parts of the shield during cooldown transients. If the heat loads are unequal, such a parallel flow arrangement can result in flow starvation in the path with higher heat load. This will cause large temperature differences and, ultimately, structural failure. Hence, an analysis was undertaken to investigate the conduction effects in the shield for other flow arrangements. 4 refs., 8 figs

  2. The Kansas State University revolving sputter source

    International Nuclear Information System (INIS)

    Tipping, T.N.

    1989-01-01

    It has been that the perfect ion source is one which runs in a very stable mode, runs continuously, and has the ability to change ion species without sacrificing the previous two requirements. This paper presents an approximation to the perfect ion source, the KSU Revolving Sputter Source. The source consists of an Aarhus-geometry sputter source with the addition of a rotating wheel which holds eight sputter cathodes. The wheel consists of a front plate with eight fixed Macor insulators and a back plate with eight Macor insulators held in place by the tension of eight springs. The cathode assembly consists of a copper cartridge with a threaded rod on one end and a sputter cathode with a threaded hole on the back. The cathode is screwed onto the cartridge and the whole assembly may be loaded into the wheel. A small spring on the side of the cartridge holds the assembly in the wheel

  3. Design Feature and Result of PFCs Baking System for the KSTAR

    International Nuclear Information System (INIS)

    Bang, Eun Nam; Kim, Kyung Min; Kim, Hong Tack; Kim, Hak Kun; Lee, Kun Su; Kim, Sang Tae; Yang, Hyung Lyeol; Kwon, Myeun

    2010-01-01

    The Korea Superconducting Tokamak Advanced Research (KSTAR) is being majorly updated for 2010's operation which mainly aims to achieve the plasma shaping and diverted plasmas. The Plasma Facing Components (PFCs) such as inboard and outboard limiters, divertors, and passive stabilizers have been finally installed in the vacuum vessel (VV) by middle of June 2010. The baking and cooling (B and C) pipe system for all the PFCs were installed inside of the vacuum vessel to fulfill baking and active cooling of each PFC components. The PFCs are to be baked by circulating hot nitrogen gas through internal tubes of back-plates of the PFCs. While VV is baked-out, the PFCs temperature was raised from room temperature to 120 .deg. C, and the baking temperature was raised again to 200 .deg. C in spite of the VV being maintained at room temperature

  4. Design Feature and Result of PFCs Baking System for the KSTAR

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Eun Nam; Kim, Kyung Min; Kim, Hong Tack; Kim, Hak Kun; Lee, Kun Su; Kim, Sang Tae; Yang, Hyung Lyeol; Kwon, Myeun [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The Korea Superconducting Tokamak Advanced Research (KSTAR) is being majorly updated for 2010's operation which mainly aims to achieve the plasma shaping and diverted plasmas. The Plasma Facing Components (PFCs) such as inboard and outboard limiters, divertors, and passive stabilizers have been finally installed in the vacuum vessel (VV) by middle of June 2010. The baking and cooling (B and C) pipe system for all the PFCs were installed inside of the vacuum vessel to fulfill baking and active cooling of each PFC components. The PFCs are to be baked by circulating hot nitrogen gas through internal tubes of back-plates of the PFCs. While VV is baked-out, the PFCs temperature was raised from room temperature to 120 .deg. C, and the baking temperature was raised again to 200 .deg. C in spite of the VV being maintained at room temperature

  5. FW/Blanket and vacuum vessel for RTO/RC ITER

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Iida, H.; Johnson, G.; Kalinin, G.; Miki, N.; Onozuka, M.; Sannazzaro, G.; Utin, Y.; Yamada, M.

    2000-01-01

    The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, ∼50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste

  6. FW/Blanket and vacuum vessel for RTO/RC ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

    The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, {approx}50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste.

  7. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  8. Analysis of classical guitars' vibrational behavior based on scanning laser vibrometer measurements

    Science.gov (United States)

    Czajkowska, Marzena

    2012-06-01

    One of the main goals in musical acoustics research is to link measurable, physical properties of a musical instrument with subjective assessments of its tone quality. The aim of the research discussed in this paper was to observe the structural vibrations of different class classical guitars in relation to their quality. This work focuses on mid-low-and low-class classical (nylon-stringed) guitars. The main source of guitar body vibrations come from top and back plate vibrations therefore these were the objects of structural mode measurements and analysis. Sixteen classical guitars have been investigated, nine with cedar and seven with spruce top plate. Structural modes of top and back plates have been measured with the aid of a scanning laser vibrometer and the instruments were excited with a chirp signal transferred by bone vibrator. The issues related to excitor selection have been discussed. Correlation and descriptive statistics of top and back plates measurement results have been investigated in relation to guitar quality. The frequency range of 300 Hz to 5 kHz as well as selected narrowed frequency bands have been analyzed for cedar and spruce guitars. Furthermore, the influence of top plate wood type on vibration characteristics have been observed on three pairs of guitars. The instruments were of the same model but different top plate material. Determination and visualization of both guitar plates' modal patterns in relation to frequency are a significant attainment of the research. Scanning laser vibrometer measurements allow particular mode observation and therefore mode identification, as opposed to sound pressure response measurements. When correlating vibration characteristics of top and back plates it appears that Pearson productmoment correlation coefficient is not a parameter that associates with guitar quality. However, for best instruments with cedar top, top-back correlation coefficient has relatively greater value in 1-2 kHz band and lower in

  9. Behavior of Combined Dielectric-Metallic Systems in a Charged Particle Environment

    Science.gov (United States)

    Gordon, W. L.; Hoffman, R. W.

    1984-01-01

    The charging and discharging characteristics of an electrically isolated solar array segment were studied in order to simulate discharges seen during geomagnetic substorms. A solar array segment was floated while bombarded with monoenergetic electrons at various angles of incidence. The potentials of the array surface and of the interconnects were monitored using Trek voltage probes to maintain electrical isolation. A back plate was capacitively coupled to the array to provide information on the characteristics of the transients accompanying the discharges. Several modes of discharging of the array were observed at relatively low differential and absolute potentials (a few kilovolts). A relatively slow discharge response in the array was observed, discharging over one second with currents of nanoamps. Two types of faster discharges were also seen which lasted a few hundredths of a millisecond and with currents on the order of microamps. Some results indicate an electron emission process associated with the arcs.

  10. Modelling of fracture processes in the ballistic impact on ceramic armours

    International Nuclear Information System (INIS)

    Zaera, R.; Sanchez-Galvez, V.

    1997-01-01

    This work examines the essential physical processes in the perforation of metal backed ceramic armours which include projectile erosion, fracture of the ceramic tile and ductile deformation of the metal backing plate. The impact of projectiles onto alumina and aluminium nitride ceramic materials is studied experimentally and numerically. Observations were performed using an X-ray shadowgraph technique to obtain accurate data of the penetration process at different times. From the examination of computer simulations and corresponding impact experiments a simple analytical model is developed by assuming some hypotheses simplifying the actual mechanisms of the penetration process. Material description is simplified by using simple equations and a few material parameters easily obtained experimentally, such as the elastic modulus, the compressive and tensile strength and the rupture strain. (orig.)

  11. High current capacity electrical connector

    International Nuclear Information System (INIS)

    Bettis, E.S.; Watts, H.L.

    1976-01-01

    An electrical connector is provided for coupling high current capacity electrical conductors such as copper busses or the like. The connector is arranged in a ''sandwiched'' configuration in which a conductor plate contacts the busses along major surfaces clamped between two stainless steel backing plates. The conductor plate is provided with contact buttons in a spaced array such that the caps of the buttons extend above the conductor plate surface to contact the busses. When clamping bolts provided through openings in the sandwiched arrangement are tightened, Belleville springs provided under the rim of each button cap are compressed and resiliently force the caps into contact with the busses' contacting surfaces to maintain a predetermined electrical contact area provided by the button cap tops. The contact area does not change with changing thermal or mechanical stresses applied to the coupled conductors

  12. Conceptual design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Sato, Satoshi; Takatsu, Hideyuki; Kurasawa, Toshimasa

    1995-03-01

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

  13. Water jacket for solid particle solar receiver

    Science.gov (United States)

    Wasyluk, David T.

    2018-03-20

    A solar receiver includes: water jacket panels each having a light-receiving side and a back side with a watertight sealed plenum defined in-between; light apertures passing through the watertight sealed plenums to receive light from the light-receiving sides of the water jacket panels; a heat transfer medium gap defined between the back sides of the water jacket panels and a cylindrical back plate; and light channeling tubes optically coupled with the light apertures and extending into the heat transfer medium gap. In some embodiments ends of the light apertures at the light receiving side of the water jacket panel are welded together to define at least a portion of the light-receiving side. A cylindrical solar receiver may be constructed using a plurality of such water jacket panels arranged with their light-receiving sides facing outward.

  14. Pre-conceptual design requirements and system description for FED frame seal welder and cutter

    International Nuclear Information System (INIS)

    Masson, L.S.; Longhurst, G.R.; Watts, K.D.; Williams, S.A.

    1981-03-01

    The Fusion Engineering Device (FED) is being designed in a torus shape using ten removable segments to form the torus geometry. The torus consists of a frame and ten shield assemblies which fit into the frame and are held in place structurally using electrically insulated backing plates. It is then necessary to seal the shield segment to the frame for the assembly to sustain an internal vacuum of 10 -7 torr. This task is intended to be accomplished by welding a frame seal between the frame and the shield segment. An example of this concept is shown. This document covers the equipment requirements and pre-conceptual design description for installing and removing the frame seal

  15. Edge plasma control using an LID configuration on CHS

    Energy Technology Data Exchange (ETDEWEB)

    Masuzaki, S.; Komori, A.; Morisaki, T. [National Inst. for Fusion Science, Oroshi, Toki (Japan)] [and others

    1997-07-01

    A Local Island Divertor (LID) has been proposed to enhance energy confinement through neutral particle control. For the case of the Large Helical Device (LHD), the separatrix of an m/n = 1/1 magnetic island, formed at the edge region, will be utilized as a divertor configuration. The divertor head is inserted in the island, and the island separatrix provides connection between the edge plasma region surrounding the core plasma and the back plate of the divertor head through the field lines. The particle flux and associated heat flux from the core plasma strike the back plate of the divertor head, and thus particle recycling is localized in this region. A pumping duct covers the divertor head to form a closed divertor system for efficient particle exhaust. The advantages of the LID are ease of hydrogen pumping because of the localized particle recycling and avoidance of the high heat load that would be localized on the leading edge of the divertor head. With efficient pumping, the neutral pressure in the edge plasma region will be reduced, and hence the edge plasma temperature will be higher, hopefully leading to a better core confinement region. A LID configuration experiment was done on the Compact Helical System (CHS) to confirm the effect of the LID. The typical effects of the LID configuration on the core plasma are reduction of the line averaged density to a half, and small or no reduction of the stored energy. In this contribution, the experimental results which were obtained in edge plasma control experiments with the LID configuration in the CHS are presented.

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

    International Nuclear Information System (INIS)

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

    1996-07-01

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

  17. Investigation of plasma parameters at BATMAN for variation of the Cs evaporation asymmetry and comparing two driver geometries

    Science.gov (United States)

    Wimmer, C.; Fantz, U.; Aza, E.; Jovović, J.; Kraus, W.; Mimo, A.; Schiesko, L.

    2017-08-01

    The Neutral Beam Injection (NBI) system for fusion devices like ITER and, beyond ITER, DEMO requires large scale sources for negative hydrogen ions. BATMAN (Bavarian Test Machine for Negative ions) is a test facility attached with the prototype source for the ITER NBI (1/8 source size of the ITER source), dedicated to physical investigations due to its flexible access for diagnostics and exchange of source components. The required amount of negative ions is produced by surface conversion of hydrogen atoms or ions on caesiated surfaces. Several diagnostic tools (Optical Emission Spectroscopy, Cavity Ring-Down Spectroscopy for H-, Langmuir probes, Tunable Diode Laser Absorption Spectroscopy for Cs) allow the determination of plasma parameters in the ion source. Plasma parameters for two modifications of the standard prototype source have been investigated: Firstly, a second Cs oven has been installed in the bottom part of the back plate in addition to the regularly used oven in the top part of the back plate. Evaporation from the top oven only can lead to a vertically asymmetric Cs distribution in front of the plasma grid. Using both ovens, a symmetric Cs distribution can be reached - however, in most cases no significant change of the extracted ion current has been determined for varying Cs symmetry if the source is well-conditioned. Secondly, BATMAN has been equipped with a much larger, racetrack-shaped RF driver (area of 32×58 cm2) instead of the cylindrical RF driver (diameter of 24.5 cm). The main idea is that one racetrack driver could substitute two cylindrical drivers in larger sources with increased reliability and power efficiency. For the same applied RF power, the electron density is lower in the racetrack driver due to its five times higher volume. The fraction of hydrogen atoms to molecules, however, is at a similar level or even slightly higher, which is a promising result for application in larger sources.

  18. "CLASS APPROACH" AND "PROLETARIAN CHARACTER" OF RUSSIAN REVOLUTION OF 1917

    Directory of Open Access Journals (Sweden)

    Эдуард Эдуардович Шульц

    2014-12-01

    Full Text Available Study of the problem of “class character” of 1917’ revolution and competency of the term “proletarian revolution”. The author considers questions of participation of various social groups in the Russian revolution, draws analogies of social composition of previous revolutions, considers the principle of “proletarian revolution”, as an ideology element for positioning of Bolsheviks and power capture. It is necessary to consider that an age, gender and national factor played much bigger role un Russian revolution than class factor. Revolution in Russia in many respects leaned on young generations which made more than a third of the population of the Russian Empire by 1917. In fight against tsarism separate calculation was based on the non-russian population and national suburbs of the empire. The special role in the Russian revolution was played by the peasantry. Revolution happened in the capital (in two capitals in Russia, the peasantry remained indifferent to revolution while Bolsheviks didn't begin to take away from them the food violently. This period:(summer - fall of 1919 became the time of peak of the Civil war. However return of landowners and their claim for property of the land forced peasants to turn bayonets for revolution and the earth and, eventually, to provide to Bolsheviks a victory in the Civil war.

  19. FRIB Cryogenic Distribution System and Status

    Energy Technology Data Exchange (ETDEWEB)

    Ganni, Venkatarao [Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Dixon, Kelly D. [Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Laverdure, Nathaniel A. [Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Yang, Shuo [Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Nellis, Timothy [Michigan State Univ., East Lansing, MI (United States); Jones, S. [Michigan State Univ., East Lansing, MI (United States); Casagrande, Fabio [Michigan State Univ., East Lansing, MI (United States)

    2015-12-01

    The MSU-FRIB cryogenic distribution system supports the 2 K primary, 4 K primary, and 35 - 55 K shield operation of more than 70 loads in the accelerator and the experimental areas. It is based on JLab and SNS experience with bayonet-type disconnects between the loads and the distribution system for phased commissioning and maintenance. The linac transfer line, which features three separate transfer line segments for additional independence during phased commissioning at 4 K and 2 K, connects the folded arrangement of 49 cryomodules and 4 superconducting dipole magnets and a fourth transfer line supports the separator area cryo loads. The pressure reliefs for the transfer line process lines, located in the refrigeration room outside the tunnel/accelerator area, are piped to be vented outdoors. The transfer line designs integrate supply and return flow paths into a combined vacuum space. The main linac distribution segments are produced in a small number of standard configurations; a prototype of one such configuration has been fabricated at Jefferson Lab and has been installed at MSU to support testing of a prototype FRIB cryomodule.

  20. Neurosurgical hand-held optical coherence tomography (OCT) forward-viewing probe

    Science.gov (United States)

    Sun, Cuiru; Lee, Kenneth K. C.; Vuong, Barry; Cusimano, Michael; Brukson, Alexander; Mariampillai, Adrian; Standish, Beau A.; Yang, Victor X. D.

    2012-02-01

    A prototype neurosurgical hand-held optical coherence tomography (OCT) imaging probe has been developed to provide micron resolution cross-sectional images of subsurface tissue during open surgery. This new ergonomic hand-held probe has been designed based on our group's previous work on electrostatically driven optical fibers. It has been packaged into a catheter probe in the familiar form factor of the clinically accepted Bayonet shaped neurosurgical non-imaging Doppler ultrasound probes. The optical design was optimized using ZEMAX simulation. Optical properties of the probe were tested to yield an ~20 um spot size, 5 mm working distance and a 3.5 mm field of view. The scan frequency can be increased or decreased by changing the applied voltage. Typically a scan frequency of less than 60Hz is chosen to keep the applied voltage to less than 2000V. The axial resolution of the probe was ~15 um (in air) as determined by the OCT system. A custom-triggering methodology has been developed to provide continuous stable imaging, which is crucial for clinical utility. Feasibility of this probe, in combination with a 1310 nm swept source OCT system was tested and images are presented to highlight the usefulness of such a forward viewing handheld OCT imaging probe. Knowledge gained from this research will lay the foundation for developing new OCT technologies for endovascular management of cerebral aneurysms and transsphenoidal neuroendoscopic treatment of pituitary tumors.

  1. Externally fired gas turbine cycles with high temperature heat exchangers utilising Fe-based ODS alloy tubing

    International Nuclear Information System (INIS)

    Olsson, F.; Svensson, S.-A.; Duncan, R.

    2001-01-01

    This work is part of the BRITE / EuRAM Project 'Development of Torsional Grain Structures to Improve Biaxial Creep Performance of Fe-based ODS Alloy Tubing for Biomass Power Plant'. The main goal of this project is to heat exchanger tubes working at 1100 o C and above. The paper deals with design implications of a biomass power plant, using an indirectly fired gas turbine with a high temperature heat exchanger containing Fe-based ODS alloy tubing. In the current heat exchanger design, ODS alloy tubing is used in a radiant section, using a bayonet type tube arrangement. This enables the use of straight sections of ODS tubing and reduces the amount of material required. In order to assess the potential of the power plant system, thermodynamic calculations have been conducted. Both co-generation and condensing applications are studied and results so far indicate that the electrical efficiency is high, compared to values reached by conventional steam cycle power plants of the same size (approx. 5 MW e ). (author)

  2. Heat transfer and flow characteristics of a cooling thimble in a molten salt reactor residual heat removal system

    Directory of Open Access Journals (Sweden)

    Zonghao Yang

    2017-12-01

    Full Text Available In the passive residual heat removal system of a molten salt reactor, one of the residual heat removal methods is to use the thimble-type heat transfer elements of the drain salt tank to remove the residual heat of fuel salts. An experimental loop is designed and built with a single heat transfer element to analyze the heat transfer and flow characteristics. In this research, the influence of the size of a three-layer thimble-type heat transfer element on the heat transfer rate is analyzed. Two methods are used to obtain the heat transfer rate, and a difference of results between methods is approximately 5%. The gas gap width between the thimble and the bayonet has a large effect on the heat transfer rate. As the gas gap width increases from 1.0 mm to 11.0 mm, the heat transfer rate decreases from 5.2 kW to 1.6 kW. In addition, a natural circulation startup process is described in this paper. Finally, flashing natural circulation instability has been observed in this thimble-type heat transfer element.

  3. Design of a 0-50 mbar pressure measurement channel compatible with the LHC tunnel radiation environment

    Science.gov (United States)

    Casas, Juan; Jelen, Dorota; Trikoupis, Nikolaos

    2017-02-01

    The monitoring of cryogenic facilities often require the measurement of pressure in the sub 5’000 Pa range that are used for flow metering applications, for saturated superfluid helium, etc. The pressure measurement is based on the minute displacement of a sensing diaphragm often through contactless techniques by using capacitive or inductive methods. The LHC radiation environment forbid the use of standard commercial sensors because of the embedded electronics that are affected both by radiation induced drift and transient Single Event Effects (SEE). Passive pressure sensors from two manufacturers were investigated and a CERN designed radiation-tolerant electronics has been developed for measuring variable-reluctance sensors. During the last maintenance stop of the LHC accelerator, four absolute pressure sensors were installed in some of the low pressure bayonet heat exchangers and four differential pressure sensors on the venturi flowmeters that monitor the cooling flow of the 20.5 kA current leads of the ATLAS end-cap superconducting toroids. The pressure sensors operating range is about 1000 to 5000 Pa and the targeted uncertainty is +/- 50 Pa which would permit to measure the equivalent saturation temperature at 1.8 K within better than 0.01 K. This paper describes the radiation hard measuring head that is based on an inductive bridge, its associated radiation-tolerant electronics that is installed under the LHC superconducting magnets or the ATLAS detector cavern; and the first operational experience.

  4. Design Document for Control Dewar and Vacuum Pump Platforms

    International Nuclear Information System (INIS)

    Rucinksi, R.

    1997-01-01

    This engineering note documents the design of the control dewar and vacuum pump platform that is to be installed on the D-Zero detector. It's purpose is twofold. Firstly it is a summary and repository of the final design calculations of the structure. Secondly, it documents that design follows the American Institute of Steel Construction (AISC) manual and applicable OSHA requirements with respect to walking working surfaces. The information contained in the main body of this note is supported by raw calculations included as the appendix. The platform is a truss type frame strucrure constructed primarily of rectangular steel tubing. The upper platform is for support of the control dewar (cryogenic/electrical interface for the solenoid), visible light photon counter (VLPC) cryogenic bayonet can, and infrequently, personnel during the connection and disconnection of the detector to building services. Figure 1 shows a layout of the structure as mounted on the detector and with the installed equipment. The connection of the platform to the detector is not conventional. Two main booms cantilever the structure to a location outside of the detector. The mounting location and support booms allow for the uninhibited motion of the detector components.

  5. Coal gasification by indirect heating in a single moving bed reactor: Process development & simulation

    Directory of Open Access Journals (Sweden)

    Junaid Akhlas

    2015-10-01

    Full Text Available In this work, the development and simulation of a new coal gasification process with indirect heat supply is performed. In this way, the need of pure oxygen production as in a conventional gasification process is avoided. The feasibility and energetic self-sufficiency of the proposed processes are addressed. To avoid the need of Air Separation Unit, the heat required by gasification reactions is supplied by the combustion flue gases, and transferred to the reacting mixture through a bayonet heat exchanger installed inside the gasifier. Two alternatives for the flue gas generation have been investigated and compared. The proposed processes are modeled using chemical kinetics validated on experimental gasification data by means of a standard process simulator (Aspen PlusTM, integrated with a spreadsheet for the modeling of a special type of heat exchanger. Simulation results are presented and discussed for proposed integrated process schemes. It is shown that they do not need external energy supply and ensure overall efficiencies comparable to conventional processes while producing syngas with lower content of carbon dioxide.

  6. Apparatus and methods for regeneration of precipitating solvent

    Science.gov (United States)

    Liu, Guohai; Vimalchand, Pannalal; Peng, Wan Wang; Bonsu, Alexander

    2015-08-25

    A regenerator that can handle rich loaded chemical solvent containing precipitated absorption reaction products is disclosed. The invention is particularly suitable for separating CO.sub.2 from large gas streams that are typical of power plant processes. The internally circulating liquid stream in the regenerator (ICLS regenerator) rapidly heats-up the in-coming rich solvent stream in a downcomer standpipe as well as decreases the overall concentration of CO.sub.2 in the mixed stream. Both these actions lead to dissolution of precipitates. Any remaining precipitate further dissolves as heat is transferred to the mixed solution with an inverted bayonet tube heat exchanger in the riser portion of the regenerator. The evolving CO.sub.2 bubbles in the riser portion of the regenerator lead to substantial gas hold-up and the large density difference between the solutions in the downcomer standpipe and riser portions promotes internal circulation of the liquid stream in the regenerator. As minor amounts of solvent components present in the exit gas stream are condensed and returned back to the regenerator, pure CO.sub.2 gas stream exits the disclosed regenerator and condenser system.

  7. FCJ-165 Obama Trolling: Memes, Salutes and an Agonistic Politics in the 2012 Presidential Election

    Directory of Open Access Journals (Sweden)

    Benjamin Burroughs

    2013-12-01

    Full Text Available During the 2012 presidential campaign an explosion of photo-shopped images circulated that depicted President Obama as unpatriotic. The ‘crotch salute’, ‘left-hand salute’, and ‘Veterans Day non-salute’ serve as case studies for understanding the role of trolling in the public sphere and Internet politics in an era of social networks and circulation. This paper tracks the cultural practices and logics of ‘sharing’ political memes and conceptualises memes as part of an agonistic public sphere and media ecology. Obama trolling is facilitated through the techno-cultural affordances of memes, which can only become public because of their mimetic form and sterilised partial anonymity. The paper seeks to conceptualise trolling as a broader cultural practice, which can be considered political. Normative assumptions about these memes would portray this trafficking as destructive to deliberative democracy but when understood as a generative cultural practice, trolling becomes central to articulating political emotions in social networks. Photo-shopped Obama salutes, in addition to Big Bird, binder, and bayonet memes, express not only political identities but also larger cultural values within networked popular culture.

  8. Using human-centered design to improve the assault rifle.

    Science.gov (United States)

    Kuo, Cheng-Lang; Yuan, Cheng-Kang; Liu, Bor-Shong

    2012-11-01

    The objective of the present study was to interview infantry soldiers to determine their preferences with respect to rifle design and to examine the effect of buttstocks on shooting performance. Factor analysis showed that seven main factors should be considered in rifle redesign including tactics necessary, interface design, saving weight, bullpup configuration, sight design, other devices, and bayonet lug. For the shooting experiment, a total of four shooting trials were performed with the T-91 rifle, with buttstock lengths of 26 mm, 34 mm, self-adjusting stock, and bullpup stock. The analysis revealed that buttstock length had a significant effect on shooting performance. The redesigned rifle weight and total length should be reduced to 3.2 kg and 750 mm, respectively. The rifle buttstock should be a non-adjustable bullpup style. The buttstock shape should be curved and the hand-guard type should be more deeply and density seams, while the trigger handle shape should be slanted. Copyright © 2012 Elsevier Ltd and The Ergonomics Society. All rights reserved.

  9. The Spear: An Effective Weapon Since Antiquity

    Directory of Open Access Journals (Sweden)

    Robert E. Dohrenwend

    2012-07-01

    Full Text Available The spear is perhaps man’s oldest weapon, and may even be his oldest tool. Over the hundreds of thousands of years of the weapon’s existence, it evolved from a pointed stick into an efficient hunting missile, and then became the most effective hand-held bladed weapon humans ever devised. The spear and its use is the only martial art originally devised for use against species other than our own, and more than any other weapon, the spear emphasizes the relationship between hunting and warfare. Virtually without exception, the spear is present wherever there are humans. The spear may claim to be the senior martial art, and the weapon is still in use today. Early techniques are preserved by the small number of Japanese sojutsu schools, and modern Chinese martial artists have developed elegant and impressive gymnastic routines for the spear. The javelin throw is a modern Olympic track and field event, and there are people who have resurrected the Aztec atlatl for sporting competition. Today, the spear is still used in Europe for hunting wild boar, and the continued issue of the obsolete bayonet to modern soldiers testifies to a deep, almost instinctive respect still possessed by the military for the spear.

  10. Design of the fill/transfer station cryostat for the OMEGA cryogenic target system

    International Nuclear Information System (INIS)

    Gibson, C.R.; Charmin, C.M.; Del Bene, J.V.; Hoffmann, E.H.; Besenbruch, G.E.; Anteby, I.

    1997-09-01

    General Atomics is designing, testing and fabricating a system for supplying cryogenic targets for the University of Rochester's OMEGA laser system. A prototype system has demonstrated the filling of 1 mm diameter, 3 microm wall plastic spheres to 111 MPa (1,100 atm) with deuterium and then cooling to 18 K to condense the fuel. The production design must be capable of routinely filling and cooling targets with a 50/50 mix of deuterium and tritium and transferring them to a device which places the targets into the focus of 60 laser beams. This paper discusses the design and analysis of the production Fill/Transfer Station cryostat. The cryostat has two major components, a fixed base and a removable dome. The joint between the base and the dome is similar to a bayonet fitting and is sealed by a room temperature elastomeric o-ring. Since the cryostat must be housed in a glovebox, its design is driven strongly by maintenance requirements. To reach the equipment inside the cryostat, the dome is simply unbolted and lifted. The inside of the cryostat is maintained at 16 K by a closed loop helium flow system. Gaseous helium at about 1.4 MPa (200 psi) flows through tubes which are brazed to the inner walls. Cooling is provided by several cryocoolers which are located external to the cryostat. Liquid nitrogen is used as a heat intercept and to precool the helium gas

  11. Detection of antipersonnel (AP) mines using mechatronics approach

    Science.gov (United States)

    Shahri, Ali M.; Naghdy, Fazel

    1998-09-01

    At present there are approximately 110 million land-mines scattered around the world in 64 countries. The clearance of these mines takes place manually. Unfortunately, on average for every 5000 mines cleared one mine clearer is killed. A Mine Detector Arm (MDA) using mechatronics approach is under development in this work. The robot arm imitates manual hand- prodding technique for mine detection. It inserts a bayonet into the soil and models the dynamics of the manipulator and environment parameters, such as stiffness variation in the soil to control the impact caused by contacting a stiff object. An explicit impact control scheme is applied as the main control scheme, while two different intelligent control methods are designed to deal with uncertainties and varying environmental parameters. Firstly, a neuro-fuzzy adaptive gain controller (NFAGC) is designed to adapt the force gain control according to the estimated environment stiffness. Then, an adaptive neuro-fuzzy plus PID controller is employed to switch from a conventional PID controller to neuro-fuzzy impact control (NFIC), when an impact is detected. The developed control schemes are validated through computer simulation and experimental work.

  12. Neurosurgical simulation and navigation with three-dimensional computer graphics.

    Science.gov (United States)

    Hayashi, N; Endo, S; Shibata, T; Ikeda, H; Takaku, A

    1999-01-01

    We developed a pre-operative simulation and intra-operative navigation system with three-dimensional computer graphics (3D-CG). Because the 3D-CG created by the present system enables visualization of lesions via semitransparent imaging of the scalp surface and brain, the expected operative field could be visualized on the computer display pre-operatively. We used two different configurative navigators. One is assembled by an arciform arm and a laser pointer. The arciform arm consists of 3 joints mounted with rotary encoders forming an iso-center system. The distal end of the arm has a laser pointer, which has a CCD for measurement of the distance between the outlet of the laser beam, and the position illuminated by the laser pointer. Using this navigator, surgeons could accurately estimate the trajectory to the target lesion, and the boundaries of the lesion. Because the other navigator has six degrees of freedom and an interchangeable probe shaped like a bayonet on its tip, it can be used in deep structures through narrow openings. Our system proved efficient and yielded an unobstructed view of deep structures during microscopic neurosurgical procedures.

  13. D0 Silicon Upgrade: West End Assembly Hall Platform Design Calculations

    International Nuclear Information System (INIS)

    Rucinski, Russ

    1996-01-01

    This engineering note documents design calculations done for the bayonet feed can platform installed at the far west end of the assembly hall. The platform is mounted off of a cast concrete wall directly south of where the shielding block wall is stacked. A summary of the loading, reaction forces and stresses is shown on the page 3. As can be seen, the calculated stresses are very small, maximum value = 2540 psi. The material used is structural steel tubing, ASTM A500 Gr. B, with a minimum yield strength of 46 ksi and minimum ultimate tensile strength of 58 ksi. The reaction forces for the upper two members will be carried together by a 1/2-inch mounting plate. The mounting plate is attached to the wall by four 1/2-inch Hilti wedge anchors. The allowables for each wedge anchor are 2400 lbs. tensile, 1960 lbs. shear. The major reaction load for the top members is a combined 3627 lbs. tensile load which can easily be handled by the four bolt pattern. Some small moment reactions not listed on the summary page add negligible (400 lbs.) force couples to the axial loading. The bottom members are also attached to a mounting plate that is bolted to the wall. See page 15 for Hilti wedge anchor data.

  14. Neutronics analysis of the conceptual design of a component test facility based on the spherical tokamak

    International Nuclear Information System (INIS)

    Zheng, S.; Voss, G.M.; Pampin, R.

    2010-01-01

    One of the crucial aspects of fusion research is the optimisation and qualification of suitable materials and components. To enable the design and construction of DEMO in the future, ITER is taken to demonstrate the scientific and technological feasibility and IFMIF will provide rigorous testing of small material samples. Meanwhile, a dedicated, small-scale components testing facility (CTF) is proposed to complement and extend the functions of ITER and IFMIF and operate in association with DEMO so as to reduce the risk of delays during this phase of fusion power development. The design of a spherical tokamak (ST)-based CTF is being developed which offers many advantages over conventional machines, including lower tritium consumption, easier maintenance, and a compact assembly. The neutronics analysis of this system is presented here. Based on a three-dimensional neutronics model generated by the interface programme MCAM from CAD models, a series of nuclear and radiation protection analyses were carried out using the MCNP code and FENDL2.1 nuclear data library to assess the current design and guide its development if needed. The nuclear analyses addresses key neutronics issues such as the neutron wall loading (NWL) profile, nuclear heat loads, and radiation damage to the coil insulation and to structural components, particularly the stainless steel vessel wall close to the NBI ports where shielding is limited. The shielding of the divertor coil and the internal Poloidal Field (PF) coil, which is introduced in the expanded divertor design, are optimised to reduce their radiation damage. The preliminary results show that the peak radiation damage to the structure of martensitic/ferritic steel is about 29 dpa at the mid-plane assuming a life of 12 years at a duty factor 33%, which is much lower than its ∼150 dpa limit. In addition, TBMs installed in 8 mid-plane ports and 6 lower ports, and 60% 6 Li enrichment in the Li 4 SiO 4 breeder, the total tritium generation is

  15. Development, simulation and testing of structural materials for DEMO

    International Nuclear Information System (INIS)

    Laesser, R.; Baluc, N.; Boutard, J.-L.; Diegele, E.; Gasparotto, M.; Riccardi, B.; Dudarev, S.; Moeslang, A.; Pippan, R.; Schaaf, B. van der

    2006-01-01

    In DEMO the structural and functional materials of the in-vessel components will be exposed to a very intense flux of fusion neutrons with energies up to 14 MeV creating displacement cascades and gaseous transmutation products. Point defects and transmutations will induce new microstructures leading to changes in mechanical and physical properties such as hardening, swelling, loss of fracture toughness and creep strength. The kinetics of microstructural evolution depends on time, temperature and defect production rates. The structural materials to be used in DEMO should have very special properties: high radiation resistance up to the dose of 100 dpa, low residual activation, high creep strength and good compatibility with the cooling media in as wide a temperature operational window as possible for the achievement of high thermal efficiency. The most promising materials are: Reduced Activation Ferritic Martensitic (RAFM) steels (Eurofer and F82H), Oxide Dispersion Strengthened (ODS) RAFM and RAF steels, SiC fibres reinforced SiC matrix composites (SiCf/SiC), tungsten (W) and W-alloys. Each of these materials has its advantages and drawbacks and will be best used under certain conditions. Presently the best studied group of materials are the RAFM steels. They require the smallest extrapolation for use in DEMO but also offer the lowest upper temperature limit of operation (550 o C) and thus the lowest thermal efficiency. The other materials foreseen for more advanced breeder blanket and divertor concepts require intense fundamental R(and)D and testing before their acceptance, whereas the so-called Test Blanket Modules (TBMs) will be constructed using RAFM steel and tested in ITER. Validation of the DEMO structural materials will be done in IFMIF, the International Fusion Materials Irradiation Facility, which will produce neutron damage and transmutation products very similar to those characterising a fusion device and will allow accelerated testing with damage rates

  16. Hydrogel based sensor arrays (2 × 2) with perforated piezoresistive diaphragms for metabolic monitoring (in vitro).

    Science.gov (United States)

    Orthner, M P; Lin, G; Avula, M; Buetefisch, S; Magda, J; Rieth, L W; Solzbacher, F

    2010-03-19

    This report details the first experimental results from novel hydrogel sensor array (2 × 2) which incorporates analyte diffusion pores into a piezoresistive diaphragm for the detection of hydrogel swelling pressures and hence chemical concentrations. The sensor assembly was comprised of three components, the active four sensors, HPMA/DMA/TEGDMA (hydroxypropyl methacrylate (HPMA), N,N-dimethylaminoethyl methacrylate (DMA) and crosslinker tetra-ethyleneglycol dimethacrylate (TEGDMA)) hydrogel, and backing plate. Each of the individual sensors of the array can be used with various hydrogels used to measure the presence of a number of stimuli including pH, ionic strength, and glucose concentrations. Ideally, in the future, these sensors will be used for continuous metabolic monitoring applications and implanted subcutaneously. In this paper and to properly characterize the sensor assembly, hydrogels sensitive to changes ionic strength were synthesized using hydroxypropyl methacrylate (HPMA), N,N-dimethylaminoethyl methacrylate (DMA) and crosslinker tetra-ethyleneglycol dimethacrylate (TEGDMA) and inserted into the sensor assembly. This hydrogel quickly and reversibly swells when placed environments of physiological buffer solutions (PBS) with ionic strengths ranging from 0.025 to 0.15 M, making it ideal for proof-of-concept testing and initial characterization. The assembly was wire bonded to a printed circuit board and coated with 3 ± 0.5 μm of Parylene-C using chemical vapor deposition (CVD) to protect the sensor and electrical connections during ionic strength wet testing. Two versions of sensors were fabricated for comparison, the first incorporated diffusion pores into the diaphragm, and the second used a solid diaphragm with perforated backing plate. This new design (perforated diaphragm) was shown to have slightly higher sensitivity than solid diaphragm sensors with separate diffuse backing plates when coupled with the hydrogel. The sensitivities for the 1 mm

  17. The influence of humanities on the teaching of technical vocabulary = La influencia de las humanidades en la enseñanza del vocabulario técnico

    Directory of Open Access Journals (Sweden)

    Verónica Vivanco Cervero

    2017-12-01

    Full Text Available Abstract A part of research in education is related to innovation. This means trying to create new approaches or teaching techniques to the object or process being taught. It has always being said that there is a gap between science/technology and arts/humanities. For this reason, this paper tries to show some examples of how to link and apply these branches of knowledge among them, as part of an innovative and creative teaching/learning process. As lexis and languages apply to everything, because all the objects and concepts need significant to express their significations, the terms of study are coupled in Spanish/English pairs. This way, students can learn vocabulary through contrast, comparison and semantic connections which trace the evolution of meanings. The complement of visuals and images helps capture the attention of students who can see the language class as a complete mosaic that links humanistic and technical knowledge. Both similarities and differences may act as a mental trigger to store vocabulary in the memory. The study has focused three war terms such as bayoneta/bayonet, bazuca/bazooka and cañón/cannon to see what type of historical, musical or cultural implications could make students preserve the meaning of these words in their mental reservoir. In addition, the complement of visuals and images helps capture the attention of students who can see the language class as a more complete approximation to knowledge. This approach to teaching tries to capture attention in a way that makes students not forget the contents of the class in their long-term memory. Resumen Una parte de la investigación en educación se relaciona con la investigación. Lo anterior implica que se han de intentar crear nuevos enfoques o métodos de enseñanza para el objeto o proceso que se estudia. Siempre se ha dicho que hay un hueco entre la ciencia y la tecnología, por un lado, y, las humanidades, por otro. Por tal motivo, este artículo intenta

  18. Computer-enhanced stereoscopic vision in a head-mounted operating binocular

    International Nuclear Information System (INIS)

    Birkfellner, Wolfgang; Figl, Michael; Matula, Christian; Hummel, Johann; Hanel, Rudolf; Imhof, Herwig; Wanschitz, Felix; Wagner, Arne; Watzinger, Franz; Bergmann, Helmar

    2003-01-01

    Based on the Varioscope, a commercially available head-mounted operating binocular, we have developed the Varioscope AR, a see through head-mounted display (HMD) for augmented reality visualization that seamlessly fits into the infrastructure of a surgical navigation system. We have assessed the extent to which stereoscopic visualization improves target localization in computer-aided surgery in a phantom study. In order to quantify the depth perception of a user aiming at a given target, we have designed a phantom simulating typical clinical situations in skull base surgery. Sixteen steel spheres were fixed at the base of a bony skull, and several typical craniotomies were applied. After having taken CT scans, the skull was filled with opaque jelly in order to simulate brain tissue. The positions of the spheres were registered using VISIT, a system for computer-aided surgical navigation. Then attempts were made to locate the steel spheres with a bayonet probe through the craniotomies using VISIT and the Varioscope AR as a stereoscopic display device. Localization of targets 4 mm in diameter using stereoscopic vision and additional visual cues indicating target proximity had a success rate (defined as a first-trial hit rate) of 87.5%. Using monoscopic vision and target proximity indication, the success rate was found to be 66.6%. Omission of visual hints on reaching a target yielded a success rate of 79.2% in the stereo case and 56.25% with monoscopic vision. Time requirements for localizing all 16 targets ranged from 7.5 min (stereo, with proximity cues) to 10 min (mono, without proximity cues). Navigation error is primarily governed by the accuracy of registration in the navigation system, whereas the HMD does not appear to influence localization significantly. We conclude that stereo vision is a valuable tool in augmented reality guided interventions. (note)

  19. Out-of-pile demonstration test of HTTR hydrogen production system structure and fabrication technology of steam reformer. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Inagaki, Yoshiyuki; Ouchi, Yoshihiro; Fujisaki, Katsuo; Kato, Michio; Uno, Hisao; Hayashi, Koji; Aita, Hideki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1999-10-01

    A hydrogen production system by steam reforming of natural gas, chemical reaction; CH{sub 4}+H{sub 2}O = 3H{sub 2}+CO, is to be the first heat utilization system of the HTTR. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile test facility is presently under construction in order to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the HTTR hydrogen production system. The out-of-pile test facility, using an electric heater as a reactor substitute, simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30 with a hydrogen production rate of 110 Nm{sup 3}/h. A steam reformer (SR) is a key component to produce hydrogen by steam reforming of natural gas. A bayonet-type catalyst tube was applied to the SR of the out-of-pile test facility in order to enhance the heat utilization rate. Also to promote heat transfer, the thickness of the catalyst tube should be decreased to 10 mm while augmenting heat transfer by fins formed on the outer surface of the catalyst tube. Therefore, the catalyst tube was designed on the basis of pressure difference between helium and process gases instead of total pressure of them. This design method was authorized for the first time in Japan. Furthermore, a function of explosion proof was applied to the SR because it contains inflammable gas and electric heater. This report describes the structure of the SR as well as the authorization both of the design method of the catalyst tube and the explosion proof function of the SR. (author)

  20. Improvements of reforming performance of a nuclear heated steam reforming process

    International Nuclear Information System (INIS)

    Hada, Kazuhiko

    1996-10-01

    Performance of an energy production process by utilizing high temperature nuclear process heat was not competitive to that by utilizing non-nuclear process heat, especially fossil-fired process heat due to its less favorable chemical reaction conditions. Less favorable conditions are because a temperature of the nuclear generated heat is around 950degC and the heat transferring fluid is the helium gas pressurized at around 4 MPa. Improvements of reforming performance of nuclear heated steam reforming process were proposed in the present report. The steam reforming process, one of hydrogen production processes, has the possibility to be industrialized as a nuclear heated process as early as expected, and technical solutions to resolve issues for coupling an HTGR with the steam reforming system are applicable to other nuclear-heated hydrogen production systems. The improvements are as follows: As for the steam reformer, (1) increase in heat input to process gas by applying a bayonet type of reformer tubes and so on, (2) increase in reforming temperature by enhancing heat transfer rate by the use of combined promoters of orifice baffles, cylindrical thermal radiation pipes and other proposal, and (3) increase in conversion rate of methane to hydrogen by optimizing chemical compositions of feed process gas. Regarding system arrangement, a steam generator and superheater are set in the helium loop as downstream coolers of the steam reformer, so as to effectively utilize the residual nuclear heat for generating feed steam. The improvements are estimated to achieve the hydrogen production rate of approximately 3800 STP-m 3 /h for the heat source of 10 MW and therefore will provide the potential competitiveness to a fossil-fired steam reforming process. Those improvements also provide the compactness of reformer tubes, giving the applicability of seamless tubes. (J.P.N.)

  1. Out-of-pile demonstration test of HTTR hydrogen production system structure and fabrication technology of steam reformer. Contract research

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Ouchi, Yoshihiro; Fujisaki, Katsuo; Kato, Michio; Uno, Hisao; Hayashi, Koji; Aita, Hideki

    1999-10-01

    A hydrogen production system by steam reforming of natural gas, chemical reaction; CH 4 +H 2 O = 3H 2 +CO, is to be the first heat utilization system of the HTTR. Prior to coupling of the steam reforming system with the HTTR, an out-of-pile test facility is presently under construction in order to confirm safety, controllability and performance of the steam reforming system under simulated operational conditions of the HTTR hydrogen production system. The out-of-pile test facility, using an electric heater as a reactor substitute, simulates key components downstream an intermediate heat exchanger of the HTTR hydrogen production system on a scale of 1 to 30 with a hydrogen production rate of 110 Nm 3 /h. A steam reformer (SR) is a key component to produce hydrogen by steam reforming of natural gas. A bayonet-type catalyst tube was applied to the SR of the out-of-pile test facility in order to enhance the heat utilization rate. Also to promote heat transfer, the thickness of the catalyst tube should be decreased to 10 mm while augmenting heat transfer by fins formed on the outer surface of the catalyst tube. Therefore, the catalyst tube was designed on the basis of pressure difference between helium and process gases instead of total pressure of them. This design method was authorized for the first time in Japan. Furthermore, a function of explosion proof was applied to the SR because it contains inflammable gas and electric heater. This report describes the structure of the SR as well as the authorization both of the design method of the catalyst tube and the explosion proof function of the SR. (author)

  2. Predictive factors for beneficial application of high-frequency electromagnetics for tumour vaporization and coagulation in neurosurgery

    Directory of Open Access Journals (Sweden)

    Koerbel Andrei

    2008-04-01

    Full Text Available Abstract Objective To identify preoperative and intraoperative factors and conditions that predicts the beneficial application of a high-frequency electromagnetic field (EMF system for tumor vaporization and coagulation. Methods One hundred three subsequent patients with brain tumors were microsurgically treated using the EMF system in addition to the standard neurosurgical instrumentarium. A multivariate analysis was performed regarding the usefulness (ineffective/useful/very helpful/essential of the new technology for tumor vaporization and coagulation, with respect to tumor histology and location, tissue consistency and texture, patients' age and sex. Results The EMF system could be used effectively during tumor surgery in 83 cases with an essential contribution to the overall success in 14 cases. In the advanced category of effectiveness (very helpful/essential, there was a significant difference between hard and soft tissue consistency (50 of 66 cases vs. 3 of 37 cases. The coagulation function worked well (very helpful/essential for surface (73 of 103 cases and spot (46 of 103 cases coagulation when vessels with a diameter of less than one millimeter were involved. The light-weight bayonet hand piece and long malleable electrodes made the system especially suited for the resection of deep-seated lesions (34 of 52 cases compared to superficial tumors (19 of 50 cases. The EMF system was less effective than traditional electrosurgical devices in reducing soft glial tumors. Standard methods where also required for coagulation of larger vessels. Conclusion It is possible to identify factors and conditions that predict a beneficial application of high-frequency electromagnetics for tumor vaporization and coagulation. This allows focusing the use of this technology on selective indications.

  3. Endoscopic cubital tunnel release using the Hoffmann technique.

    Science.gov (United States)

    Hoffmann, Reimer; Lubahn, John

    2013-06-01

    Endoscopic cubital tunnel release was originally described in 1989 by Tsai, and his technique has been modified by other surgeons including Mirza and Cobb. In 2006, Hoffmann and Siemionow described an endoscopic technique quite different from Tsai's original description. Instead of working from the "inside out," Hoffmann's technique is performed through an incision similar to that which would be used for an in situ release of the ulnar nerve. The main difference being that the nerve can be explored and decompressed 10 cm proximal and distal to the arcuate ligament as the surgeon looks down on the nerve and the surrounding tissues while viewing the anatomy through a camera attached to a soft tissue endoscope that is inserted in the wound. The arcuate (Osborne's) ligament is released under direct vision much like a standard in situ decompression. Using a blunt dissection instrument, a workspace is created proximally and distally to the cubital tunnel. Next an illuminated speculum is introduced, the nerve is directly visualized between 4 and 5 cm proximal and distal to the cubital tunnel, and potential compressive forearm fasciae or fibrous bands are released. Finally, a 15-cm, 30° soft tissue endoscope is introduced into the incision, and viewing the internal anatomy on a video monitor, the decompression continues using longer scissors. Any potential bleeding is controlled with a long bayonet bipolar cautery. The authors discuss indications, contraindications, and the surgical technique. Postoperative management and associated complications are also discussed. Copyright © 2013 American Society for Surgery of the Hand. Published by Elsevier Inc. All rights reserved.

  4. High-Voltage, Low-Power BNC Feedthrough Terminator

    Science.gov (United States)

    Bearden, Douglas

    2012-01-01

    This innovation is a high-voltage, lowpower BNC (Bayonet Neill-Concelman) feedthrough that enables the user to terminate an instrumentation cable properly while connected to a high voltage, without the use of a voltage divider. This feedthrough is low power, which will not load the source, and will properly terminate the instrumentation cable to the instrumentation, even if the cable impedance is not constant. The Space Shuttle Program had a requirement to measure voltage transients on the orbiter bus through the Ground Lightning Measurement System (GLMS). This measurement has a bandwidth requirement of 1 MHz. The GLMS voltage measurement is connected to the orbiter through a DC panel. The DC panel is connected to the bus through a nonuniform cable that is approximately 75 ft (approximately equal to 23 m) long. A 15-ft (approximately equal to 5-m), 50-ohm triaxial cable is connected between the DC panel and the digitizer. Based on calculations and simulations, cable resonances and reflections due to mismatched impedances of the cable connecting the orbiter bus and the digitizer causes the output not to reflect accurately what is on the bus. A voltage divider at the DC panel, and terminating the 50-ohm cable properly, would eliminate this issue. Due to implementation issues, an alternative design was needed to terminate the cable properly without the use of a voltage divider. Analysis shows how the cable resonances and reflections due to the mismatched impedances of the cable connecting the orbiter bus and the digitizer causes the output not to reflect accurately what is on the bus. After simulating a dampening circuit located at the digitizer, simulations were performed to show how the cable resonances were dampened and the accuracy was improved significantly. Test cables built to verify simulations were accurate. Since the dampening circuit is low power, it can be packaged in a BNC feedthrough.

  5. Performance characterization of the FLEX low pressure helium facility for fusion technology experiments

    Energy Technology Data Exchange (ETDEWEB)

    Schlindwein, Georg, E-mail: schlindwein@kit.edu; Arbeiter, Frederik

    2014-10-15

    Highlights: • A gas loop for fusion R and D has been built and tested. • Facility requirements and their implementation are given. • The loop's functions and instrumentation are explained. • The loops performance has been characterized. - Abstract: FLEX (Fluid Dynamics Experimental Facility) is a multi-purpose small scale gas loop for research on fluid and thermodynamic investigations, especially heat transfer, flow field measurements and gas purification. Initially it was built for investigation on mini-channel gas-flow to design the HFTM module of IFMIF. Because of its versatility it offers a wide range of further applications, e.g. the research of pressure drops in mockups of breeder units of the helium cooled pebble bed (HCPB) test blanket module for ITER. The main parameters of the loop, which can be operated with inert gases and air are: (i) operation gas pressure 0.02–0.38 MPa abs., (ii) test section pressure head up to 0.12 MPa, (iii) tolerable gas temperature RT – 200 °C and (iv) mass flow rate 0.2–12 × 10{sup −3} kg/s for Helium. This paper gives a detailed view of the loop assembly with the components that generate and regulate the mass flow and loop pressure. The measurement instrumentation will be presented as well as a representative mass flow-pressure drop characteristic. Furthermore, the achievable gas purity will be discussed.

  6. Preliminary results of the International Fusion Materials Irradiation Facility deuteron injector

    Energy Technology Data Exchange (ETDEWEB)

    Gobin, R.; Adroit, G.; Bogard, D.; Bourdelle, G.; Chauvin, N.; Delferriere, O.; Gauthier, Y.; Girardot, P.; Guiho, P.; Harrault, F.; Jannin, J. L.; Loiseau, D.; Mattei, P.; Roger, A.; Sauce, Y.; Senee, F.; Vacher, T. [Commissariat a l' Energie Atomique et aux Energie Alternatives, CEA/Saclay, DSM/IRFU, 91191-Gif/Yvette (France)

    2012-02-15

    In the framework of the IFMIF-EVEDA project (International Fusion Materials Irradiation Facility-Engineering Validation and Engineering Design Activities), CEA/IRFU is in charge of the design, construction, and characterization of the 140 mA continuous deuteron injector, including the source and the low energy beam line. The electron cyclotron resonance ion source which operates at 2.45 GHz is associated with a 4-electrode extraction system in order to minimize beam divergence at the source exit. Krypton gas injection is foreseen in the 2-solenoid low energy beam line. Such Kr injection will allow reaching a high level of space charge compensation in order to improve the beam matching at the radio frequency quadrupole (RFQ) entrance. The injector construction is now completed on the Saclay site and the first plasma and beam production has been produced in May 2011. This installation will be tested with proton and deuteron beams either in pulsed or continuous mode at Saclay before shipping to Japan. In this paper, after a brief description of the installation, the preliminary results obtained with hydrogen gas injection into the plasma chamber will be reported.

  7. Proceedings of the IEA-technical workshop for an international fusion materials irradiation facility. IEA-implementing agreement for a programme of research and development on fusion materials

    International Nuclear Information System (INIS)

    Ehrlich, K.; Lindau, R.

    1995-07-01

    The workshop was initiated to deal with the following objectives: (1) Critical review of the requirements for IFMIF from the user's point of view; (2) Definition of a baseline concept for the CDA-study; (3) Formation of working groups for main fields of activities; (4) Identification of tasks and critical issues for main components; (5) Development of a working break-down structure, distribution of work and milestones for CDA-activities; (6) Documentation of main results. According to the enclosed agenda the mission for a Conceptual Design Study, the requirements for an intense neutron source from the user's point of view and the baseline concept for an accelerator-driven D-Li neutron source were discussed in several plenary sessions. In three subgroups (SG 1 Accelerators, SG2 Lithium Target and SG3 Users and Test Cell) technical concepts for the different components and facilities were discussed in detail, critical issues and tasks for the concept study were identified. Finally, the sharing of tasks to the different national parties, questions of organisation of the work, flow of information and definition of milestones was agreed upon. The detailled summary reports of the subgroups and the contributions of the plenary sessions are presented in the proceedings. (orig./HP)

  8. Three-dimensional coupled Monte Carlo-discrete ordinates computational scheme for shielding calculations of large and complex nuclear facilities

    International Nuclear Information System (INIS)

    Chen, Y.; Fischer, U.

    2005-01-01

    Shielding calculations of advanced nuclear facilities such as accelerator based neutron sources or fusion devices of the tokamak type are complicated due to their complex geometries and their large dimensions, including bulk shields of several meters thickness. While the complexity of the geometry in the shielding calculation can be hardly handled by the discrete ordinates method, the deep penetration of radiation through bulk shields is a severe challenge for the Monte Carlo particle transport technique. This work proposes a dedicated computational scheme for coupled Monte Carlo-Discrete Ordinates transport calculations to handle this kind of shielding problems. The Monte Carlo technique is used to simulate the particle generation and transport in the target region with both complex geometry and reaction physics, and the discrete ordinates method is used to treat the deep penetration problem in the bulk shield. The coupling scheme has been implemented in a program system by loosely integrating the Monte Carlo transport code MCNP, the three-dimensional discrete ordinates code TORT and a newly developed coupling interface program for mapping process. Test calculations were performed with comparison to MCNP solutions. Satisfactory agreements were obtained between these two approaches. The program system has been chosen to treat the complicated shielding problem of the accelerator-based IFMIF neutron source. The successful application demonstrates that coupling scheme with the program system is a useful computational tool for the shielding analysis of complex and large nuclear facilities. (authors)

  9. Development of small specimen test techniques. Development of a remote controlled small punch testing apparatus

    International Nuclear Information System (INIS)

    Ohmi, Masao; Saito, Junichi; Ooka, Norikazu; Jitsukawa, Shiro; Hishinuma, Akimichi; Umino, Akira.

    1997-01-01

    An accelerator-driven deuterium-lithium (d-Li) stripping reaction-type neutron source, such as the International Fusion Materials Irradiation Facility (IFMIF) planned by the International Energy Agency is recognized as one of the most promising facility to obtain test environments of high-energy neutrons for fusion reactor materials development. The limitation on the available irradiation volume of the irradiation facility requires the development of the small specimen test techniques (SSTT). Application of SSTT to evaluate the degradation of various components in the light water reactor for the life extension is expected to be also quite beneficial. A remote-controlled testing machine for the Small Punch (SP) and miniaturized tensile tests was developed at the hot laboratory of the Japan Materials Testing Reactor (JMTR). The machine is designed for testing at temperatures ranging between 93 and 1,123 K to evaluate the temperature dependence of the strength of materials including the embrittlement at low temperatures and the softening at elevated temperatures. The tests are performed in a vacuum or in an inert gas environment. The machine has been installed in a hot cell and is being used for the round robin test program of the SP test method. The round robin test program is planned to identify the capability of the test method and to establish a standard test procedure. The configuration and the specifications of the test machine are introduced and the results of the SP tests are also shown. (author)

  10. DEMO concepts and their roles within the fusion programme

    International Nuclear Information System (INIS)

    Tran, Minh Quang

    2007-01-01

    In the past years, the international fusion community has developed models of fusion power plants, which were extremely useful in showing the key advantages of fusion energy and pointing out he areas of development. The present view is that between ITER and such power plants (even of ''first of kind'' type), there is a need for one or two intermediate steps. The need to have a ''fast rack'' towards such a fusion reactor, suggested that the steps after ITER, which are usually considered to be a Demonstration power plant followed by a Prototypical one, could be combines into one known as a DEMO. DEMO would then be a device capable of producing electricity, paving the way towards fusion power plants which would be economically viable. This talk outlines the DEMO concepts as the necessary physics and technological extrapolation from the envisaged future steps (ITER, IFMIF) are discussed. It attempts to provide a coverage of the different concepts developed by various countries, The key issues, as foreseen today, and their implications for the programme are highlighted. (orig.)

  11. Overview of the European Union fusion nuclear technologies development and essential elements on the way to DEMO

    International Nuclear Information System (INIS)

    Andreani, R.; Diegele, E.; Gulden, W.; Laesser, R.; Maisonnier, D.; Murdoch, D.; Pick, M.; Poitevin, Y.

    2006-01-01

    EU is strongly preparing ITER construction involving the system of EU Associations, universities and industry. The European programme has been steered to be in line with the present conception of a future power reactor. Thirty percent of the fusion research budget has been spent on long-term related activities managed by EFDA. These include Power Plant Conceptual Studies (PPCS), the recently undertaken DEMO Conceptual Studies, design and R and D for breeder blankets, low activation materials and IFMIF. Developments on fuel cycle, neutronics, safety and socio-economics complement those specifically performed for ITER. Two EU helium-cooled DEMO blankets will be tested in ITER, using liquid lithium-lead and solid ceramics as breeders. The blanket structures will use EUROFER. Irradiations to 70-80 dpa will qualify EUROFER for DEMO. Advanced materials, in particular SiC f /SiC, under development, could provide more thermodynamically efficient blankets. Even with a fully successful ITER, a number of issues will remain open in technology. The application of high temperature superconductors, essential progress in materials, blanket design and remote handling, are required to produce environmentally safe and economically competitive fusion. A fully integrated world wide international programme is the best way to efficiently progress in these fields

  12. Research and development plan of fusion technologies in JAERI toward DEMO reactors

    International Nuclear Information System (INIS)

    Nishitani, Takeo; Hayashi, Takumi; Abe, Tetsuya; Akiba, Masato; Isono, Takaaki; Inoue, Takashi; Enoeda, Mikio; Okuno, Kiyoshi; Koizumi, Norikiyo; Sakamoto, Keishi; Sato, Satoshi; Jitsukawa, Shiro; Sugimoto, Masayoshi; Suzuki, Satoshi; Seki, Shogo; Takatsu, Hideyuki; Tanzawa, Sadamitsu; Tsuchiya, Kunihiko; Nishi, Masataka; Hayashi, Kimio; Matsui, Hideki; Yamanishi, Toshihiko; Watanabe, Kazuhiro

    2005-03-01

    In accordance with the 'Third Phase Basic Program on Fusion Research and Development' established by the Fusion Council of the Japan Atomic Energy Commission, research and development (R and D) of fusion technologies aim at realization of two elements: development of ITER key components and their improvement for higher performances; and construction of sound technical basis of fusion nuclear technologies essential for fusion energy utilization. JAERI has been assigned in the Third Phase Basic Program as a responsible institute for developing the above two elements, and accordingly has been implementing technology R and Ds categorized in the following three areas: R and D for ITER construction and operation; R and D for ITER utilization (blanket testing in ITER) and toward DEMO; and R and D on basic fusion technologies. The present report reviews the status and the plan of fusion technology R and Ds in the latter two areas, and presents the technical objectives, technical issues, status of R and D and near-term R and D plans for: breeding blankets; structural materials; the IFMIF program; improvements of the key ITER components for higher performances toward DEMO; and basic fusion technologies. (author)

  13. TBM/MTM for HTS-FNSF: An Innovative Testing Strategy to Qualify/Validate Fusion Technologies for U.S. DEMO

    Directory of Open Access Journals (Sweden)

    Laila El-Guebaly

    2016-08-01

    Full Text Available The qualification and validation of nuclear technologies are daunting tasks for fusion demonstration (DEMO and power plants. This is particularly true for advanced designs that involve harsh radiation environment with 14 MeV neutrons and high-temperature operating regimes. This paper outlines the unique qualification and validation processes developed in the U.S., offering the only access to the complete fusion environment, focusing on the most prominent U.S. blanket concept (the dual cooled PbLi (DCLL along with testing new generations of structural and functional materials in dedicated test modules. The venue for such activities is the proposed Fusion Nuclear Science Facility (FNSF, which is viewed as an essential element of the U.S. fusion roadmap. A staged blanket testing strategy has been developed to test and enhance the DCLL blanket performance during each phase of FNSF D-T operation. A materials testing module (MTM is critically important to include in the FNSF as well to test a broad range of specimens of future, more advanced generations of materials in a relevant fusion environment. The most important attributes for MTM are the relevant He/dpa ratio (10–15 and the much larger specimen volumes compared to the 10–500 mL range available in the International Fusion Materials Irradiation Facility (IFMIF and European DEMO-Oriented Neutron Source (DONES.

  14. Report of the 13th IEA workshop on radiation effects in ceramic insulators

    International Nuclear Information System (INIS)

    2004-03-01

    The 13th IEA Workshop on Radiation Effects in Ceramic Insulators, based on Annex II: Experimentation on Radiation Damage in Fusion Materials, to the IEA Implementing Agreement for a Programme of Research and Development on Radiation Damage in Fusion Materials, was held on the 9th, December, 2003, at Kyoto International Conference Center, in Kyoto, Japan, in conjunction with the 11th International Conference on Fusion Reactor Materials (ICFRM-11). 44 participants from 10 countries (26 from Japan, 5 from Spain, 3 from Belgium, 3 from USA, 2 from RF, each 1 from Austria, Greece, Italy, Romania, and UK) gathered together and discussed following issues extensively, with the newest experimental results, after the welcome remarks by one of the organizer and chairpersons, Dr. E.R. Hodgson of CIEMAT. Effects of electric field on radiation induced microstructural evolution, parasitic electrical current and voltage induced in cables and wires by radiation effects, optical materials, IFMIF related issues and fundamental aspects were discussed. Significant results such as an observation of γ-alumina and aluminum colloid formation for the Radiation Induced Electrical Degradation mechanism are obtained. This report is workshop summary, abstracts and documents of the 13th IEA Workshop on Radiation Effects in Ceramic Insulators. (author)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  16. 23. Symposium On Fusion Technology (SOFT), Venice - A personal view

    International Nuclear Information System (INIS)

    Spears, W.R.

    2004-01-01

    This conference, examining the advances in our leading-edge technology, took place on 22-24 September 2004 against the wonderful and historic backdrop of Venice, at a monastery of the Cini Foundation, on the Island of St. Giorgio, directly opposite St. Marks. The strong connection between the ancient and modern was brought home to us in the very first talk, from the Mayor of Venice and MEP Prof. P. Costa, who reminded us of Venice's particular problem with global warming, and urged us to do our part to develop an energy source that should help to avoid it drowning. Prof. Sir C. Llewellyn-Smith, head of the UK Fusion Programme and Chairman of Euratom CCE-FU, took up this theme and elaborated how we should reach our goal, showing in particular the urgency of pursuing a fast track, proceeding with ITER and the International Fusion Materials Irradiation Facility (IFMIF) without further delay, and envisaging that the subsequent machine would be prototypical of future commercial fusion power plants. The conference proceeded through plenary and oral sessions, and through poster sessions, covering plasma heating, fuelling, control and diagnostics, magnets and power supplies, plasma-facing components, blanket and vessel, remote handling, materials technology, the experiences gained on existing experiments, and projections for future experiments and fusion power plants. There were 570 participants, from 25 countries, of whom a third came from outside Europe

  17. The development of EUROFER reduced activation steel

    Energy Technology Data Exchange (ETDEWEB)

    Schaaf, B. van der E-mail: vanderschaaf@nrg-nl.com; Tavassoli, F.; Fazio, C.; Rigal, E.; Diegele, E.; Lindau, R.; LeMarois, G

    2003-09-01

    Ferritic martensitic steels show limited swelling and susceptibility to helium effects and can be made with low activation chemical compositions. These properties make them the reference steel for the development of breeding blankets in fusion power plants. EUROFER97 is the European implementation of such a steel, where experience gained from an IEA co-operation with Japan and the US is also implemented. Results obtained so far show that EUROFER steel has attractive mechanical properties even after long ageing times. Compatibility tests in water and PbLi17 are in progress. Oxidised aluminium is the most effective protective layer in PbLi17. The displacement damage and helium formation strongly influence the hydrogen transport in the steel. Present experiments should be backed by tests in a more fusion relevant environment, e.g. IFMIF. The 2.5 dpa neutron irradiations at low temperatures result in a higher DBTT. High dose irradiations, up to 80 dpa, are underway. The early results of ODS grades with EUROFER steel composition show potential of these grades for increasing the operating temperature with 100-150 K.

  18. Assessment of the gas dynamic trap mirror facility as intense neutron source for fusion material test irradiations

    International Nuclear Information System (INIS)

    Fischer, U.; Moeslang, A.; Ivanov, A.A.

    2000-01-01

    The gas dynamic trap (GDT) mirror machine has been proposed by the Budker Institute of nuclear physics, Novosibirsk, as a volumetric neutron source for fusion material test irradiations. On the basis of the GDT plasma confinement concept, 14 MeV neutrons are generated at high production rates in the two end sections of the axially symmetrical central mirror cell, serving as suitable irradiation test regions. In this paper, we present an assessment of the GDT as intense neutron source for fusion material test irradiations. This includes comparisons to irradiation conditions in fusion reactor systems (ITER, Demo) and the International Fusion Material Irradiation Facility (IFMIF), as well as a conceptual design for a helium-cooled tubular test assembly elaborated for the largest of the two test zones taking proper account of neutronics, thermal-hydraulic and mechanical aspects. This tubular test assembly incorporates ten rigs of about 200 cm length used for inserting instrumented test capsules with miniaturized specimens taking advantage of the 'small specimen test technology'. The proposed design allows individual temperatures in each of the rigs, and active heating systems inside the capsules ensures specimen temperature stability even during beam-off periods. The major concern is about the maximum achievable dpa accumulation of less than 15 dpa per full power year on the basis of the present design parameters of the GDT neutron source. A design upgrading is proposed to allow for higher neutron wall loadings in the material test regions

  19. Fracture toughness evaluation of Eurofer'97 by testing small specimens

    International Nuclear Information System (INIS)

    Serrano, M.; Fernandez, P.; Lapena, J.

    2006-01-01

    The Eurofer'97 is the structural reference material that will be tested in the ITER modules. Its metallurgical properties have been well characterized during the last years. However, more investigations related with the fracture toughness of this material are necessary because this property is one of the most important to design structural components and to study their integrity assessment. In the case of structural materials for fusion reactor the small specimen technology (SSTT) are being actively developed to investigate the fracture toughness among other mechanical properties. The use of small specimens is due to the small available irradiation volume of IFMIF and also due to the high fluence expected in the fusion reactor. The aim of this paper is to determine the fracture toughness of the Eurofer'97 steel by testing small specimens of different geometry in the ductile to brittle transition region, with the application of the Master Curve methodology, and to evaluate this method to assess the decrease in fracture toughness due to neutron irradiation. The tests and data analysis have been performed following the Master Curve approach included in the ASTM Standard E1921-05. Specimen size effect and comparison of the fracture toughness results with data available in the literature are also considered. (author)

  20. High energy neutron source for materials research and development

    International Nuclear Information System (INIS)

    Odera, M.

    1989-01-01

    Requirements for neutron source for nuclear materials research are reviewed and ESNIT, Energy Selective Neutron Irradiation Test facility proposed by JAERI is discussed. Its principal aims of a wide neutron energy tunability and spectra peaking at each energy to enable characterization of material damage process are demanding but attractive goals which deserve detailed study. It is also to be noted that the requirements make a difference in facility design from those of FMIT, IFMIF and other high energy intense neutron sources built or planned to date. Areas of technologies to be addressed to realize the ESNIT facility are defined and discussed. In order to get neutron source having desired spectral characteristics keeping moderate intensity, projectile and target combinations must be examined including experimentation if necessary. It is also desired to minimize change of flux density and energy spectrum according to location inside irradiation chamber. Extended target or multiple targets configuration might be a solution as well as specimen rotation and choice of combination of projectile and target which has minimum velocity of the center of mass. Though relevant accelerator technology exists, it is to be stressed that considerable efforts must be paid, especially in the area of target and irradiation devices to get ESNIT goal. Design considerations to allow hands-on maintenance and future upgrading possibility are important either, in order to exploit the facility fully for nuclear materials research and development. (author)

  1. Overview of fusion nuclear technology in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Andreani, R. E-mail: roberto.andreani@tech.efda.org; Gasparotto, M. E-mail: maurizio.gasparotto@tech.efda.org

    2002-11-01

    The fusion nuclear technology programme in the EU is focussed on materials and breeding blankets development, tritium and high heat flux component technologies. A strong effort is also devoted to the validation of the design of an intense 14 MeV neutron source (IFMIF). The material programme includes the development of reduced activation ferritic martensitic steel (EUROFER) to be used as structural material in a DEMO reactor, and potentially more attractive higher performance materials: ODS and SiC/SiC composites. The breeding blanket activities are focussed in the preparation of the two European Test Blanket Moduli to be installed in ITER. The Fuel Cycle activities for ITER include development of the torus exhaust cryopump, fuel storage system, performance characterisation of the torus exhaust processing and design of water detritiation system. High heat flux components have been developed in the framework of ITER R and D programme and based on copper alloy heat sink protected by an armour of beryllium, CFC or tungsten. Studies give an important contribution in defining the nuclear technology programme strategy.

  2. Biological Derived Nanomotors in a ``Domino Fashion''

    Science.gov (United States)

    Maksoed, W. H.

    2015-11-01

    For disproportionation of H2 O2 , we also considers an electrokinetic mechanism they appear.So far, the more efficient micro/nanoscale motors are derived from biological systems [2003]. Besides, a control experimenting using 3 stripped Au/Pt/Au rods with catalyzed the composition of H2 O2 , at a similar rate-Walter F Paxton: ``Catalytic Nanomotors,'' JACS, 2004. We also intended to accomaplishes the HCCI quotes from Marcin Frackowiak, dissertation, 2009, just in several characters seems as twin of IGNITION through IceCube document project held since Oct 11, 2001 ever concludes as ``saw none'' so they can be follows the ITER/IFMIF. Refers to S29286 file in UI retrieved: ``magnetic quantum-dot cellular automata which is nonvolatile & lower power consist of nanomagnets. Since they are magnetically coupled, logic can be performed by switching, on the other hand in a DOMINO fashion..'' [A. Klenm: ``Fabrication of Magnetic Tunnel Junction-based Spintronic Devices..,'' convocation, Aug 11-14, 2010]. Acknowledgments devotes to BB Mandelbrot: ``Fractal Geometry: What is it & What Does it do?''.

  3. Overview of fusion nuclear technology in Europe

    International Nuclear Information System (INIS)

    Andreani, R.; Gasparotto, M.

    2002-01-01

    The fusion nuclear technology programme in the EU is focussed on materials and breeding blankets development, tritium and high heat flux component technologies. A strong effort is also devoted to the validation of the design of an intense 14 MeV neutron source (IFMIF). The material programme includes the development of reduced activation ferritic martensitic steel (EUROFER) to be used as structural material in a DEMO reactor, and potentially more attractive higher performance materials: ODS and SiC/SiC composites. The breeding blanket activities are focussed in the preparation of the two European Test Blanket Moduli to be installed in ITER. The Fuel Cycle activities for ITER include development of the torus exhaust cryopump, fuel storage system, performance characterisation of the torus exhaust processing and design of water detritiation system. High heat flux components have been developed in the framework of ITER R and D programme and based on copper alloy heat sink protected by an armour of beryllium, CFC or tungsten. Studies give an important contribution in defining the nuclear technology programme strategy

  4. The European Activation File, EAF-2005

    International Nuclear Information System (INIS)

    Forrest, R.A.

    2005-01-01

    The current version of the European Activation File is EAF-2003. This contains various libraries of nuclear data required for activation calculations. An important component is the neutron-induced cross-section library. Plans to expose fusion components to high neutron fluxes include the IFMIF materials testing facility. This accelerator-based device will produce neutrons with a high-energy tail up to about 55 MeV. In order to carry out activation calculations on materials exposed to such neutrons it is necessary to extend the energy range of the cross-section library. Work on extending the energy range to 60 MeV is nearing completion. A test version (EAF-2004) was produced at the end of 2003 showing the feasibility of the chosen approach. This library required calculated data to extend the existing data from 20-60 MeV and to enlarge it with new classes of reactions with high thresholds. A summary of the new library EAF-2005, which is under development and is planned for distribution at the beginning of 2005, is given. The other files in EAF-2005 are briefly described; these cover cross-section uncertainty information and decay data. Both these have been extended beyond the current version to allow activation calculations at energies up to 60 MeV

  5. Applications of proton and deuteron accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Todd, A.M.M. (Grumman Corporate Research Center, Princeton, NJ (United States))

    1993-06-01

    Applications of positive and negative hydrogen and deuterium ion accelerators beyond basic research are increasing. Large scale proposed national laboratory/industrial projects include the Accelerator Production of Tritium (APT) which will utilize protons, and the International Fusion Material Irradiation Facility (IFMIF) which will accelerate a deuteron beam into a lithium target. At the small scale end, radio-frequency quadrupole (RFQ) accelerator based systems have been built for neutron activation analysis and for applications such as explosive detection. At an intermediate scale, the Loma Linda proton therapy accelerator is now successfully treating a full schedule of patients, and more than half a dozen such hospital based units are under active study world-wide. At this same scale, there are also several ongoing negative ion, military accelerator projects which include the Continuous Wave Deuterium Demonstrator (CWDD) and the Neutral Particle Beam Space Experiment (NPBSE). These respective deuterium and hydrogen accelerators, which have not been previously described, are the focus of this paper. (orig.)

  6. High-order adaptive secondary mirrors: where are we?

    Science.gov (United States)

    Salinari, Piero; Sandler, David G.

    1998-09-01

    We discuss the current developments and the perspective performances of adaptive secondary mirrors for high order adaptive a correction on large ground based telescopes. The development of the basic techniques involved a large collaborative effort of public research Institutes and of private companies is now essentially complete. The next crucial step will be the construction of an adaptive secondary mirror for the 6.5 m MMT. Problems such as the fabrication of very thin mirrors, the low cost implementation of fast position sensors, of efficient and compact electromagnetic actuators, of the control and communication electronics, of the actuator control system, of the thermal control and of the mechanical layout can be considered as solved, in some cases with more than one viable solution. To verify performances at system level two complete prototypes have been built and tested, one at ThermoTrex and the other at Arcetri. The two prototypes adopt the same basic approach concerning actuators, sensor and support of the thin mirror, but differ in a number of aspects such as the material of the rigid back plate used as reference for the thin mirror, the number and surface density of the actuators, the solution adopted for the removal of the heat, and the design of the electronics. We discuss how the results obtained by of the two prototypes and by numerical simulations will guide the design of full size adaptive secondary units.

  7. Water-based metamaterial absorbers for optical transparency and broadband microwave absorption

    Science.gov (United States)

    Pang, Yongqiang; Shen, Yang; Li, Yongfeng; Wang, Jiafu; Xu, Zhuo; Qu, Shaobo

    2018-04-01

    Naturally occurring water is a promising candidate for achieving broadband absorption. In this work, by virtue of the optically transparent character of the water, the water-based metamaterial absorbers (MAs) are proposed to achieve the broadband absorption at microwave frequencies and optical transparence simultaneously. For this purpose, the transparent indium tin oxide (ITO) and polymethyl methacrylate (PMMA) are chosen as the constitutive materials. The water is encapsulated between the ITO backed plate and PMMA, serving as the microwave loss as well as optically transparent material. Numerical simulations show that the broadband absorption with the efficiency over 90% in the frequency band of 6.4-30 GHz and highly optical transparency of about 85% in the visible region can be achieved and have been well demonstrated experimentally. Additionally, the proposed water-based MA displays a wide-angle absorption performance for both TE and TM waves and is also robust to the variations of the structure parameters, which is much desired in a practical application.

  8. Densitometric HPTLC analysis of 8-gingerol in Zingiber officinale extract and ginger-containing dietary supplements, teas and commercial creams.

    Science.gov (United States)

    Alam, Prawez

    2013-08-01

    To develop and validate a simple, accurate HPTLC method for the analysis of 8-gingerol and to determine the quantity of 8-gingerol in Zingiber officinale extract and ginger-containing dietary supplements, teas and commercial creams. The analysis was performed on 10×20 cm aluminium-backed plates coated with 0.2 mm layers of silica gel 60 F254 (E-Merck, Germany) with n-hexane: ethyl acetate 60: 40 (v/v) as mobile phase. Camag TLC Scanner III was used for the UV densitometric scanning at 569. This system was found to give a compact spot of 8-gingerol at retention factor (Rf) value of (0.39±0.04) and linearity was found in the ranges 50-500 ng/spot (r (2)=0.9987). Limit of detection (12.76 ng/spot), limit of quantification (26.32 ng/spot), accuracy (less than 2 %) and recovery (ranging from 98.22-99.20) were found satisfactory. The HPTLC method developed for quantification of 8-gingerol was found to be simple, accurate, reproducible, sensitive and is applicable to the analysis of 8-gingerol in Zingiber officinale extract and ginger-containing dietary supplements, teas and commercial creams.

  9. The design of the poloidal divertor experiment tokamak wall armor and inner limiter system

    International Nuclear Information System (INIS)

    Kugel, H.W.; Ulrickson, M.

    1982-01-01

    The inner wall protective plates for the Poloidal Divertor Experiment Tokamak are designed to absorb 8 MW of neutral deuterium beam power at maximum power densities of 3 kW/cm 2 for pulse lengths of 0.5 s. Preliminary studies indicate that the design could survive several pulses of l-s duration. The design consists of a tile and mounting plate structure. The mounting plates are water cooled to allow short duty cycles and beam calorimetry. The temperature and flow of the coolant are measured to obtain the injected power. A thermocouple array on the tiles provides beam position and power density profiles. Several material combinations for the tiles were subjected to thermal tests using both electron and neutral beams, and titanium-carbidecoated graphite was selected as the tile material. The heat transfer coefficient of the tile backing plate structure was measured to determine the maximum pulse rate allowable. The design of the armor system allows the structure to be used as a neutral beam power diagnostic and as an inner plasma limiter. The electrical and cooling systems external to the vacuum vessel are discussed

  10. Effect of process parameters on optimum welding condition of DP590 steel by friction stir welding

    International Nuclear Information System (INIS)

    Kim, Young Gon; Kim, Ji Sun; Kim, In Ju

    2014-01-01

    In the automotive industry, vehicle weight reduction techniques have been actively studied to improve the rate of fuel consumption and to cope with the regulation restricting exhaust gas. For this reason, advanced high-strength steel (AHSS) is preferred in the automobile industry as its tensile strength is 590 MPa and over. In this study, to obtain the optimum welding condition, the friction stir welding (FSW) process applied to AHSS was considered. The FSW experiment was performed on a stir plate using a Si 3 N 4 tool and a 1.4-mm thick DP590 steel sheet manufactured by cold rolling. In addition, to investigate the temperature distribution of the advancing and retreating sides in the welding state, the tool rotation speed of 800 rpm, and the welding speed of 180 mm/min, a K-type thermocouple was inserted in the backing plate, and the peak temperature was evaluated at each point. Especially, the correlation between the heat input per unit length and the formation of the FSW zone was minutely analyzed.

  11. Microscopic Examination of Cold Spray Cermet Sn+In2O3 Coatings for Sputtering Target Materials.

    Science.gov (United States)

    Winnicki, M; Baszczuk, A; Rutkowska-Gorczyca, M; Jasiorski, M; Małachowska, A; Posadowski, W; Znamirowski, Z; Ambroziak, A

    2017-01-01

    Low-pressure cold spraying is a newly developed technology with high application potential. The aim of this study was to investigate potential application of this technique for producing a new type of transparent conductive oxide films target. Cold spraying technique allows the manufacture of target directly on the backing plate; therefore the proposed sputtering target has a form of Sn+In 2 O 3 coating sprayed onto copper substrate. The microstructure and properties of the feedstock powder prepared using three various methods as well as the deposited ones by low-pressure cold spraying coatings were evaluated, compared, and analysed. Produced cermet Sn+In 2 O 3 targets were employed in first magnetron sputtering process to deposit preliminary, thin, transparent conducting oxide films onto the glass substrates. The resistivity of obtained preliminary films was measured and allows believing that fabrication of TCO (transparent conducting oxide) films using targets produced by cold spraying is possible in the future, after optimization of the deposition conditions.

  12. Violin

    Science.gov (United States)

    Curtin, Joseph; Rossing, Thomas D.

    The first known violins were built in Italy in the early 1500s. While not much is yet known about the instrument's prior development, European forebears include the rebec and the Renaissance fiddle, which themselves evolved from instruments found in the ancient Eastern world. The violin brought together in a particularly happy way features seen in a variety of earlier stringed instruments. Arched plates increased the stiffness-to-mass ratio of the body, creating a more brilliant sound and helping resist long-term deformation. A pronounced waist gave the bow access to the outermost strings, while the precisely calibrated curves of fingerboard and bridge enabled the strings to be played individually as well as in two-, three-, and even four-part chords. In contrast to the viola da gamba and guitar, the violin's top and back plates overhung the ribs, allowing easy removal for repairs, thus contributing to the instrument's fabled longevity. A graceful outline, harmonious proportions, and the minimal use of ornamentation together lent the violin a timeless beauty - explaining in part why it has resisted significant stylistic modification to this day. For a discussion of historical string instruments, see Chap. 17.

  13. Spectroscopic Analysis to Characterize Finishing Treatments of Ancient Bowed String Instruments.

    Science.gov (United States)

    Fiocco, Giacomo; Rovetta, Tommaso; Gulmini, Monica; Piccirillo, Anna; Licchelli, Maurizio; Malagodi, Marco

    2017-11-01

    Historical bowed string instruments exhibit acoustic features and aesthetic appeal that are still considered inimitable. These characteristics seem to be in large part determined by the materials used in the ground and varnishing treatments after the assembly of the instrument. These finishing processes were kept secret by the violinmakers and the traditional methods were handed down orally from master craftsmen to apprentices. Today, the methods of the past can represent a secret to be revealed through scientific investigations. The "Cremonese" methods used in the 17th and 18th centuries were lost as the last Great Masters from the Amati, Guarneri, and Stradivari families passed away. In this study, we had the chance of combining noninvasive and microinvasive techniques on six fragments of historical musical instruments. The fragments were detached from different instruments during extraordinary maintenance and restoration treatments, which involved the substitution of severely damaged structural parts like top plates, back plates, or ribs. Therefore, the fragments can offer to the scientists a valuable overview on the materials and techniques used by the violinmakers. The results obtained by portable X-ray fluorescence, optical microscopy, scanning electron microscopy coupled with energy dispersive X-ray spectrometry, and Fourier transform infrared microscopy allowed us to: (1) determine the stratigraphy of six instruments; (2) obtain new information about the materials involved in the finishing processes employed in Cremona; and (3) elucidate the technological relationship among the procedures adopted in the violin making workshops during the considered period.

  14. ELECTROMAGNETIC CALORIMETER (ECAL)

    CERN Multimedia

    Philippe Bloch

    ECAL Barrel (EB) Great progress has been achieved during the last few months on Barrel commissioning. All 36 supermodules have been run concurrently during the CRUZET in early May. The EB readout has reached the expected performance and is included regularly with central DAQ.  ECAL has been used as a source of triggers during cosmic runs. ECAL Endcaps (EE) Important milestones have been recently achieved: The Endcaps crystal production was completed in mid March. The gluing of the VPTs (Vacuum Photo Triodes) on the crystals, the assembly of Supercrystals (a set of 25 crystals) and their mounting on the Dee backplates (including the connection of the laser monitoring fibers) were finished during May. The mechanical assembly of the four endcap Dees is therefore completed. The assembly of the services and electronics on the backside of the Dees’ back-plates is also proceeding at a fast speed. The laying of the high voltage cables, the inner moderator, the optical fibers for the LED stabilit...

  15. Material Flow and Oxide Particle Distributions in Friction-Stir Welded F/M-ODS Sheets

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Suk Hoon; Noh, Sanghoon; Jin, Hyun Ju; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    It is well known that uniform nano-oxide dispersoids act as pinning points to obstruct dislocation and grain boundary motion in ODS(Oxide dispersion strengthened) steel. However, these advantages will disappear while the material is subjected to the high temperature of conventional fusion welding. There is only limited literature available on the joining of ODS steels. Friction stir welding (FSW) is considered to be the best welding technique for welding ODS steels as the technique helps in retaining the homogeneous nano-oxide particles distributions in matrix. FSW is a solid.state, hot.shear joining process in which a rotating tool with a shoulder and terminating in a threaded pin, moves along the butting surfaces of two rigidly clamped plates placed on a backing plate. Heat generated by friction at the shoulder and to a lesser extent at the pin surface, softens the material being welded. Severe plastic deformation and flow of this plasticised metal occurs as the tool is translated along the welding direction. Material is transported from the front of the tool to the trailing edge where it is forged into a joint. Friction stir welding appears to be a very promising technique for the welding of FMS and ODS steels. This study found that, during FSW, the forward movement of the tool pin results in loose contact between the tool pin and the receding material on the advancing side.

  16. Characterization of retrokeratoprosthetic membranes in the Boston type 1 keratoprosthesis.

    Science.gov (United States)

    Stacy, Rebecca C; Jakobiec, Frederick A; Michaud, Norman A; Dohlman, Claes H; Colby, Kathryn A

    2011-03-01

    To evaluate retroprosthetic membranes that can occur in 25% to 65% of patients with the Boston type 1 keratoprosthesis (KPro). Two patients with Peter anomaly and 2 with neurotrophic scarred corneas underwent revisions of their type 1 KPros because of visually compromising retroprosthetic membranes. The excised membranes were studied by light microscopy with hematoxylin-eosin, periodic acid-Schiff, and toluidine blue stains. Immunohistochemical and transmission electron microscopic examination were also used. Light microscopic examination revealed that the retro-KPro fibrous membranes originated from the host's corneal stroma. These mildly to moderately vascularized membranes grew through gaps in the Descemet membrane to reach behind the KPro back plate and adhere to the anterior iris surface, which had undergone partial lysis. In 2 cases, the fibrous membranes merged at the pupil with matrical portions of metaplastic lens epithelium, forming a bilayered structure that crossed the optical axis. Retro-KPro membranes stained positively for α-smooth muscle actin but negatively for pancytokeratin. Electron microscopy confirmed the presence of actin filaments within myofibroblasts and small surviving clusters of metaplastic lens epithelial cells. Stromal downgrowth, rather than epithelial downgrowth, was the major element of the retro-KPro membranes in this series. Metaplastic lens epithelium also contributed to opacification of the visual axis. Florid membranous inflammation was not a prominent finding and thus probably not a requisite stimulus for membrane development. Further advances in prosthetic design and newer antifibroproliferative agents may reduce membrane formation.

  17. Effect of process parameters on optimum welding condition of DP590 steel by friction stir welding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Gon; Kim, Ji Sun; Kim, In Ju [Korea Institute of Industrial Technology, Gwangju (Korea, Republic of)

    2014-12-15

    In the automotive industry, vehicle weight reduction techniques have been actively studied to improve the rate of fuel consumption and to cope with the regulation restricting exhaust gas. For this reason, advanced high-strength steel (AHSS) is preferred in the automobile industry as its tensile strength is 590 MPa and over. In this study, to obtain the optimum welding condition, the friction stir welding (FSW) process applied to AHSS was considered. The FSW experiment was performed on a stir plate using a Si{sub 3}N{sub 4} tool and a 1.4-mm thick DP590 steel sheet manufactured by cold rolling. In addition, to investigate the temperature distribution of the advancing and retreating sides in the welding state, the tool rotation speed of 800 rpm, and the welding speed of 180 mm/min, a K-type thermocouple was inserted in the backing plate, and the peak temperature was evaluated at each point. Especially, the correlation between the heat input per unit length and the formation of the FSW zone was minutely analyzed.

  18. Performance of the BATMAN RF source with a large racetrack shaped driver

    Science.gov (United States)

    Kraus, W.; Schiesko, L.; Wimmer, C.; Fantz, U.; Heinemann, B.

    2017-08-01

    In the negative ion sources in neutral beam injection systems (NBI) of future fusion reactors the plasma is generated in up to eight cylindrical RF sources ("drivers") from which it expands into the main volume. For these large sources, in particular those used in the future DEMO NBI, a high RF efficiency and operational reliability is required. To achieve this it could be favorable to substitute each pair of drivers by one larger one. To investigate this option the cylindrical driver of the BATMAN source at IPP Garching has been replaced by a large source with a racetrack shaped base area and tested using the same extraction system. The main differences are a five times larger source volume and another position of the Cs oven which is mounted onto the driver`s back plate and not onto the expansion volume. The conditioning characteristics and the plasma symmetry in front of the plasma grid were very similar. The extracted H- current densities jex are comparable to that achieved with the small driver at the same power. Because no saturation of jex occurred at 0.6 Pa at high power and the source allows high power operation, a maximum value 45.1 mA/cm2 at 103 kW has been reached. Sputtered Cu from the walls of the expansion volume affected the performance at low pressure, particularly in deuterium. The experiments will be therefore continued with Mo coating of all inner walls.

  19. Image characterization of computed radiography

    International Nuclear Information System (INIS)

    Candeias, Janaina P.; Saddock, Aline; Oliveira, Davi F.; Lopes, Ricardo T.

    2007-01-01

    The digital radiographic image became a reality as of the 80's decade. Since then, several works have been developed with the aim of reducing the exposure time to ionizing radiation obtaining in this way an excellent image quality with a minimum exposure. In the Computerized Radiography, the conventional film is substituted for Image Plate (IP) which consists of a radiosensitive layer of phosphor crystals on a polyester backing plate. The unique design makes it reusable and easy to handle. When exposed, the IP accumulates and stores the irradiated radioactive energy. In order to qualify a computerized radiography system it is necessary to evaluate the Image Plate. In this work it was performed a series of experimental procedures with the aim of evaluating the responses characteristics for different plates. For this purpose it was used a computerized radiographic system CR Tower Scanner - GE, with three different types of IPs, all of them manufactured by GE, whose nomenclatures are IPC, IPX and IPS. It was used the Rhythm Acquire and Review programs for image acquisition and treatment, respectively. (author)

  20. Investigation on prediction capability of nuclear design parameters for gap configuration in ITER through analysis of the FNS gap streaming experiment

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Konno, Chikara; Kasugai, Yoshimi; Oyama, Yukio; Uno, Yoshitomo; Maekawa, Hiroshi; Ikeda, Yujiro

    2000-01-01

    As an R and D Task of shielding neutronics experiment under the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER), streaming experiments with simulating a gap configuration formed by two neighboring blanket modules of ITER were carried out at the FNS (Fusion Neutron Source) facility. In this work, prediction capability of various nuclear design parameters was investigated through analysis of the experiments. The Monte Carlo transport calculation code MCNP-4A and the FENDL/E-1.0 and JENDL Fusion File cross section data libraries were used for the analysis with detailed modeling of the experimental conditions. As a result, all the measured quantities were reproduced within about ±30% by the calculations. It was concluded that these calculation tools were capable of predicting nuclear design parameters, such as helium production rates at connection legs of blanket modules to the back plate and nuclear responses in toroidal field coils, with uncertainty of ±30% for the geometry where gap-streaming effect was significant. (author)

  1. Metal/graphite - composites in fusion engineering

    International Nuclear Information System (INIS)

    Staffler, R.; Kneringer, G.; Kny, E.; Reheis, N.

    1989-01-01

    Metal/graphite composites have been well known in medical industry for many years. X-ray tubes used in modern radiography, particularly in computerized tomography are equipped with rotating targets able to absorb a maximum of heat in a given time. Modern rotating targets consist of a refractory metal/graphite composite. Today the use of graphite as a plasma facing material is one predominant concept in fusion engineering. Depending on the thermal load, the graphite components have to be directly cooled (i.e. divertor plates) or inertially cooled (i.e. firstwall tiles). In case of direct cooling a metallurgical joining such as high temperature brazing between graphite and a metallic cooling structure shows the most promising results /1/. Inertially cooled graphite tiles have to be joined to a metallic backing plate in order to get a stable attachment to the supporting structure. The main requirements on the metallic partner of a metal/graphite composite used in the first wall area are: high melting point, high thermal strength, high thermal conductivity, low vapor pressure and a thermal expansion matching that of graphite. These properties are typical for the refractory metals such as molybdenum, tungsten and their alloys. 4 refs., 13 figs., 1 tab

  2. Metal/graphite - composites in fusion engineering

    International Nuclear Information System (INIS)

    Staffler, R.; Kneringer, G.; Kny, E.; Reheis, N.

    1995-01-01

    Metal/graphite composites have been well known in medical industry for many years. X-ray tubes used in modern radiography, particulary in computerized tomography are equipped with rotating targets able to absorb a maximum of heat in a given time. Modern rotating targets consist of a refractory metal/graphite composite. Today the use of graphite as a plasma facing material is one predominant concept in fusion engineering. Depending on the thermal load, the graphite components have to be directly cooled (i.e. divertor plates) or inertially cooled (i.e. firstwall tiles). In case of direct cooling a metallurgical joining such as high temperature brazing between graphite and a metalic cooling structure shows the most promising results /1/. Inertially cooled graphite tiles have to be joined to a metallic backing plate in order to get a stable attachment to the supporting structure. The main requirements on the metallic partner of a metal/graphite composite and in the first wall area are: high melting point, high thermal strength, high thermal conductivity, low vapour pressure and a thermal expansion matching that of graphite. These properties are typical for the refractory metals such as molybdenum, tungsten and their alloys. (author)

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  4. 3D eddy-current distribution in a tokamak first wall during a plasma disruption using 'TRIFOU'

    International Nuclear Information System (INIS)

    Chaussecourte, P.; Bossavit, A.; Verite, J.C.; Crutzen, Y.R.

    1989-01-01

    In fusion reactor studies there is a lack of knowledge concerning the electromagnetic-type of phenomena generated by a plasma disruption event (rapid quenching of the plasma current). The induced eddy current distribution in space and time in the passive conducting structural components surrounding the plasma ring needs to be accurately investigated. TRIFOU is a full 3D eddy-current computer program based on a mixed FEM and BIEM technique, using the magnetic field, h, as a state variable, It has already been used in various areas of interest including static or rotating machines, non-destructive testing, induction heating, and research devices such as tokamaks. It can take into account various geometries and a wide range of physical situations (time dependency, physical properties, etc.). The present application is related to the eddy-current situation arising from a strong electromagnetic transient generated in the NET (Next European Torus) first wall segment. With respect to previous numerical simulations, the general 3D approach for the current density shows different eddy current circulations in the front/side shells and in the stiff back plate. The results obtained by TRIFOU are illustrated by means of advanced computer graphic displays and an animation movie. (orig.)

  5. Engineering design and analysis of Indian LLCB TBM set

    Energy Technology Data Exchange (ETDEWEB)

    Ranjithkumar, S., E-mail: ranjith@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat (India); Sharma, Deepak; Chaudhuri, Paritosh; Danani, Chandan; Kumar, E. Rajendra [Institute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat (India); Khan, Istiyak; Bhattacharya, Sujay; Vyas, K.N. [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2016-11-01

    India is developing Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) for testing in ITER for the validation of fusion blanket design tools, tritium breeding performance and high grade heat extraction capability relevant to Indian DEMO. LLCB TBM is designed to withstand various ITER loads and its combinations, like thermal, mechanical including the high pressure coolant loads, electromagnetic loads during plasma disruption and seismic loading conditions. LLCB TBM system is designed in compliance with ITER Safety requirements and guidelines. A few challenging part in the design includes the design of helium cooled First Wall (FW) and back plates, the attachment system between TBM and the shield block to withstand loads for all the ITER operational modes, routing of high temperature process pipes between TBM and the shield block, interfaces between process pipes and the connecting flange, design of manifolds of different process fluids etc. Analysis has been performed on the LLCB TBM set using a Finite Element code, ANSYS. Relevant codes and standards, namely the French code RCC-MR, has been followed for the design analysis. The details of the analysis and further plans and proposals for improvement in design will be discussed in this paper.

  6. Structural materials for ITER in-vessel component design

    Energy Technology Data Exchange (ETDEWEB)

    Kalinin, G. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Gauster, W. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Matera, R. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Tavassoli, A.-A.F. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France); Rowcliffe, A. [Oak Ridge National Lab., TN (United States); Fabritsiev, S. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kawamura, H. [JAERI, IMTR Project, Ibaraki (Japan). Blanket Irradiation Lab.

    1996-10-01

    The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m{sup 2} in the basic performance phase (BPP)) within a temperature range from 20 to 300 C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. (orig.).

  7. Investigation of the thermal performance of {sup nat}Te target for {sup 124}I production in the RARS cyclotron

    Energy Technology Data Exchange (ETDEWEB)

    Azizakram, Hamid; Zolfagharpour, Farhad [Univ. of Mohaghegh Ardabili, Ardabil (Iran, Islamic Republic of). Dept. of Physics; Sadeghi, Mahdi [Iran Univ. of Medical Science, Tehran (Iran, Islamic Republic of). Medical Physics Dept.; Ashtari, Parviz [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Radiation Application Research School (RARS); Nikjou, Amir [Payam Noor Univ., Tehran (Iran, Islamic Republic of). Dept. of Physics

    2017-12-15

    Regarding the low thermal conductivity of {sup nat}Te element, the provision of an effective cooling system is one of the critical issues in cyclotron targetry to prevent melting of target matter during the irradiation to {sup 124}I production via {sup nat}Te(p,xn){sup 124}I reaction. Heat transfer on Te target and efficiency of cooling fluid in the solid target system have been simulated based on a Finite Element Analysis (FEA) code for the thermal behavior of the target during the irradiation and under different beam currents, coolant flow rates, substrate matters and target geometry. The results on the routinely used solid target in Radiation Application Research School (RARS) cyclotron showed that in a 3 m/s coolant flow rate, by using a fined-cooling system and a nickel substrate coated on copper backing plate, the irradiation beam current can be raised up to 180 μA without any risk of melting. The cooling flow rates greater than 3 m/s do not noticeably improve the heat dispersion of target layer. As expected, a linear increase was observed for the temperature and temperature gradient of plates in the beam currents of 100-300 μA.

  8. Design of ITER neutron monitor using micro fission chambers

    International Nuclear Information System (INIS)

    Nishitani, Takeo; Ebisawa, Katsuyuki; Ando, Toshiro; Kasai, Satoshi; Johnson, L.C.; Walker, C.

    1998-08-01

    We are designing micro fission chambers, which are pencil size gas counters with fissile material inside, to be installed in the vacuum vessel as neutron flux monitors for ITER. We found that the 238 U micro fission chambers are not suitable because the detection efficiency will increase up to 50% in the ITER life time by breading 239 Pu. We propose to install 235 U micro fission chambers on the front side of the back plate in the gap between adjacent blanket modules and behind the blankets at 10 poloidal locations. One chamber will be installed in the divertor cassette just under the dome. Employing both pulse counting mode and Campbelling mode in the electronics, we can accomplish the ITER requirement of 10 7 dynamic range with 1 ms temporal resolution, and eliminate the effect of gamma-rays. We demonstrate by neutron Monte Carlo calculation with three-dimensional modeling that we avoid those detection efficiency changes by installing micro fission chambers at several poloidal locations inside the vacuum vessel. (author)

  9. Detailed design of the RF source for the 1 MV neutral beam test facility

    International Nuclear Information System (INIS)

    Marcuzzi, D.; Palma, M. Dalla; Pavei, M.; Heinemann, B.; Kraus, W.; Riedl, R.

    2009-01-01

    In the framework of the EU activities for the development of the Neutral Beam Injector for ITER, the detailed design of the Radio Frequency (RF) driven negative ion source to be installed in the 1 MV ITER Neutral Beam Test Facility (NBTF) has been carried out. Results coming from ongoing R and D on IPP test beds [A. Staebler et al., Development of a RF-Driven Ion Source for the ITER NBI System, this conference] and the design of the new ELISE facility [B. Heinemann et al., Design of the Half-Size ITER Neutral Beam Source Test Facility ELISE, this conference] brought several modifications to the solution based on the previous design. An assessment was carried out regarding the Back-Streaming positive Ions (BSI+) that impinge on the back plates of the ion source and cause high and localized heat loads. This led to the redesign of most heated components to increase cooling, and to different choices for the plasma facing materials to reduce the effects of sputtering. The design of the electric circuit, gas supply and the other auxiliary systems has been optimized. Integration with other components of the beam source has been revised, with regards to the interfaces with the supporting structure, the plasma grid and the flexible connections. In the paper the design will be presented in detail, as well as the results of the analyses performed for the thermo-mechanical verification of the components.

  10. Microscopic Examination of Cold Spray Cermet Sn+In2O3 Coatings for Sputtering Target Materials

    Directory of Open Access Journals (Sweden)

    M. Winnicki

    2017-01-01

    Full Text Available Low-pressure cold spraying is a newly developed technology with high application potential. The aim of this study was to investigate potential application of this technique for producing a new type of transparent conductive oxide films target. Cold spraying technique allows the manufacture of target directly on the backing plate; therefore the proposed sputtering target has a form of Sn+In2O3 coating sprayed onto copper substrate. The microstructure and properties of the feedstock powder prepared using three various methods as well as the deposited ones by low-pressure cold spraying coatings were evaluated, compared, and analysed. Produced cermet Sn+In2O3 targets were employed in first magnetron sputtering process to deposit preliminary, thin, transparent conducting oxide films onto the glass substrates. The resistivity of obtained preliminary films was measured and allows believing that fabrication of TCO (transparent conducting oxide films using targets produced by cold spraying is possible in the future, after optimization of the deposition conditions.

  11. Aeroacoustic Simulation for NASA CC3 Centrifugal Compressor Operating at off Design Condition

    Directory of Open Access Journals (Sweden)

    Alqaradawi Mohamed

    2016-01-01

    Full Text Available This paper covers the characterization of the acoustic noise and the unsteady flow field of a high speed centrifugal compressor NASA CC3. In order to accurately predict the noise, all analyses are carried out through the use of Large Eddy Simulation and Ffowcs Williams–Hawkings model for noise prediction. The relative effect of hub cavity on flow characteristics and sound levels is investigated, for a compressor stage with a total pressure ratio equal to 4, working from surge to near choke condition. In comparison with the experimental results from literature, the predicted compressor performance and flow field are predicted well. The hub cavity flow effect on the compressor aeroacoustic generated noise is shown in the paper. The unsteady static pressure and sound pressure levels are compared not only at different location but also for design and off design operating points. The internal flow results inside the hub cavity are presented at surge, design and near choke points. The conclusion is that the cavity effect of the centrifugal compressor cannot be ignored in the numerical prediction of aerodynamic generated noise. The impeller back plate of the rotor experiences a strong pressure fluctuation, which is maxima at the impeller outer radius for all operating point, but higher pressure values at the surge point.

  12. Apparatus for precision micromachining with lasers

    Science.gov (United States)

    Chang, J.J.; Dragon, E.P.; Warner, B.E.

    1998-04-28

    A new material processing apparatus using a short-pulsed, high-repetition-rate visible laser for precision micromachining utilizes a near diffraction limited laser, a high-speed precision two-axis tilt-mirror for steering the laser beam, an optical system for either focusing or imaging the laser beam on the part, and a part holder that may consist of a cover plate and a back plate. The system is generally useful for precision drilling, cutting, milling and polishing of metals and ceramics, and has broad application in manufacturing precision components. Precision machining has been demonstrated through percussion drilling and trepanning using this system. With a 30 W copper vapor laser running at multi-kHz pulse repetition frequency, straight parallel holes with size varying from 500 microns to less than 25 microns and with aspect ratios up to 1:40 have been consistently drilled with good surface finish on a variety of metals. Micromilling and microdrilling on ceramics using a 250 W copper vapor laser have also been demonstrated with good results. Materialographic sections of machined parts show little (submicron scale) recast layer and heat affected zone. 1 fig.

  13. PIV analysis of merging flow in a simplified model of a rotary kiln

    Energy Technology Data Exchange (ETDEWEB)

    Larsson, I.A.S.; Granstroem, B.R.; Lundstroem, T.S. [Luleaa University of Technology, Division of Fluid and Experimental Mechanics, Luleaa (Sweden); Marjavaara, B.D. [LKAB, Kiruna (Sweden)

    2012-08-15

    Rotary kilns are used in a variety of industrial applications. The focus in this work is on characterizing the non-reacting, isothermal flow field in a rotary kiln used for iron ore pelletization. A downscaled, simplified model of the kiln is experimentally investigated using particle image velocimetry. Five different momentum flux ratios of the two inlet ducts to the kiln are investigated in order to evaluate its effect on the flow field in general and the recirculation zone in particular. Time-averaged and phase-averaged analyses are reported, and it is found that the flow field resembles that of two parallel merging jets, with the same characteristic flow zones. The back plate separating the inlet ducts acts as a bluff body to the flow and creates a region of reversed flow behind it. Due to the semicircular cross-section of the jets, the wake is elongated along the walls. Conclusions are that the flow field shows a dependence on momentum flux ratio of the jets; as the momentum flux ratio approaches unity, there is an increasing presence of von Karman-type coherent structures with a Strouhal number of between 0.16 and 0.18. These large-scale structures enhance the mixing of the jets and also affect the size of the recirculation zone. It is also shown that the inclination of the upper inlet duct leads to a decrease in length of the recirculation zone in certain cases. (orig.)

  14. A preliminary study on the local impact behavior of Steel-plate Concrete walls

    International Nuclear Information System (INIS)

    Kim, Kap-sun; Moon, Il-hwan; Choi, Hyung-jin; Nam, Deok-woo

    2017-01-01

    International regulations for nuclear power plants strictly prescribe the design requirements for local impact loads, such as aircraft engine impact, and internal and external missile impact. However, the local impact characteristics of Steel-plate Concrete (SC) walls are not easy to evaluate precisely because the dynamic impact behavior of SC walls which include external steel plate, internal concrete, tie-bars, and studs, is so complex. In this study, dynamic impact characteristics of SC walls subjected to local missile impact load are investigated via actual high-speed impact test and numerical simulation. Three velocity checkout tests and four SC wall tests were performed at the Energetic Materials Research and Testing Center (EMRTC) site in the USA. Initial and residual velocity of the missile, strain and acceleration of the back plate, local failure mode (penetration, bulging, splitting and perforation) and deformation size, etc. were measured to study the local behavior of the specimen using high speed cameras and various other instrumentation devices. In addition, a more advanced and applicable numerical simulation method using the finite element (FE) method is proposed and verified by the experimental results. Finally, the experimental results are compared with the local failure evaluation formula for SC walls recently proposed, and future research directions for the development of a refined design method for SC walls are reviewed.

  15. Design of an Experimental Facility for Passive Heat Removal in Advanced Nuclear Reactors

    Science.gov (United States)

    Bersano, Andrea

    With reference to innovative heat exchangers to be used in passive safety system of Gen- eration IV nuclear reactors and Small Modular Reactors it is necessary to study the natural circulation and the efficiency of heat removal systems. Especially in safety systems, as the decay heat removal system of many reactors, it is increasing the use of passive components in order to improve their availability and reliability during possible accidental scenarios, reducing the need of human intervention. Many of these systems are based on natural circulation, so they require an intense analysis due to the possible instability of the related phenomena. The aim of this thesis work is to build a scaled facility which can reproduce, in a simplified way, the decay heat removal system (DHR2) of the lead-cooled fast reactor ALFRED and, in particular, the bayonet heat exchanger, which transfers heat from lead to water. Given the thermal power to be removed, the natural circulation flow rate and the pressure drops will be studied both experimentally and numerically using the code RELAP5 3D. The first phase of preliminary analysis and project includes: the calculations to design the heat source and heat sink, the choice of materials and components and CAD drawings of the facility. After that, the numerical study is performed using the thermal-hydraulic code RELAP5 3D in order to simulate the behavior of the system. The purpose is to run pretest simulations of the facility to optimize the dimensioning setting the operative parameters (temperature, pressure, etc.) and to chose the most adequate measurement devices. The model of the system is continually developed to better simulate the system studied. High attention is dedicated to the control logic of the system to obtain acceptable results. The initial experimental tests phase consists in cold zero power tests of the facility in order to characterize and to calibrate the pressure drops. In future works the experimental results will be

  16. Accelerators for Fusion Materials Testing

    Science.gov (United States)

    Knaster, Juan; Okumura, Yoshikazu

    with the International Fusion Materials Irradiation Facility (IFMIF) under discussion at the time. Worldwide technological efforts are maturing soundly and the time for a fusion-relevant neutron source has arrived according to world fusion roadmaps; if decisions are taken we could count the next decade with a powerful source of 14 MeV neutrons thanks to the expected significant results of the Engineering Validation and Engineering Design Activity (EVEDA) phase of the IFMIF project. The accelerator know-how has matured in all possible aspects since the times of FMIT conception in the 1970s; today, operating 125 mA deuteron beam at 40 MeV in CW with high availabilities seems feasible thanks to the understanding of the beam halo physics and the three main technological breakthroughs in accelerator technology: (1) the ECR ion source for light ions developed at Chalk River Laboratories in the early 1990s, (2) the RFQ operation of H+ in CW with 100 mA demonstrated by LEDA in LANL in the late 1990s, and (3) the growing maturity of superconducting resonators for light hadrons and low β beams achieved in recent years.

  17. Development of Fusion Nuclear Technologies and the role of MTR's

    International Nuclear Information System (INIS)

    Laan, J.G. van der; Schaaf, B. van der

    2006-01-01

    design for the EU ITER Test Blanket Module. The duration of the irradiation is relevant for the total TBM operation during the ITER lifetime. The major result is that the basic design soundness has been demonstrated under ITER relevant conditions. Besides the ceramic breeder concept experiments with lithium lead breeder sub components are continued to measure the effects of transmutation product helium on the liquid metal properties. Similarly, activities are ongoing to perform in-pile testing of primary wall components, allowing to address fatigue type loading conditions. In the next decade 14 MeV sources such as ITER, IFMIF and maybe a volumetric source will support the crucial demonstration of components under near fusion plasma nuclear conditions. These sources have limitations in accumulated total damage (ITER) irradiation volume (IFMIF) and control. MTR's will thus continue to supply essential facts on component behaviour and materials in parallel to 14 MeV sources. The present generation of MTR's will be closed in this and next decade because they reach their end of life. The new generation will be utilised for 4 major areas of nuclear interest: energy, science, health and environmental issues. Fusion and the next generation fission (Generation 4) power plant development will share the areas energy and science in the next decades. The design and concept of the new MTR's will centre on faster development cycles, thus higher fluxes up to 5 x 10 18 nm -2 . Several MTR replacements in the EU are in different design stages such as the Reacteur Jules Horowitz in France and PALLAS in the Netherlands. The conceptual design of the replacement for the HFR, Petten, named PALLAS envisages a fruitful co-operation of the experimenters for advanced fission power reactor and fusion plant components. Materials science will also be able to use modern MTR facilities for the modelling of radiation damage in both fission and fusion environments. The development of primary fusion

  18. Design Concept of a Seal-off Type 14 MeV Neutron Generator of 10''1''1n/s Range

    Energy Technology Data Exchange (ETDEWEB)

    In, S. R.; Oh, B. H. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The total neutron fluence during the life time is expected to be around 10MW·yr/m''2 which may cause a damage of -100 dpa in materials. To estimate the adaptability of candidate materials in a few years, a 14MeV neutron source with a flux level of 3 - 5 x 10''1''8 n/s·m''2, which is the goal of the IFMIF facility costing more than ¤1000M, is necessitated. The problem in making an intense neutron generator of beam target type is really not on the neutron production rate, but on the huge heat generated in the target, because the fusion power is only one of thousands of beam power exerted on the target. We have a plan to develop neutron generators step by step from a 10''8 n/s level. The final goal is establishing a 14MeV neutron irradiation facility at 10''1''4 intensity level.. Up to the 10''1''0 n/s level, there occurs basically no critical thermal problem, because beam power density is in the range of tens W/cm''2. The neutron generator designed in a sealed-off type because of tritium safety is mainly composed of an ion source, target, reaction chamber, and getter pump.. The major design concepts for the neutron generator with the neutron production rate of 10''1''1 n/s range were presented. The specifications of the ion source, target and getter have been determined for attaining the goal of the neutron generation rate.

  19. Neutrons for science (NFS) at spiral-2

    International Nuclear Information System (INIS)

    Ridikas, D.

    2005-01-01

    Both cross section measurements and various applications could be realised successfully using the high energy neutrons that will be produced at SPIRAL-2. Two particular cases were examined in more detail, namely: (a) neutron time-of-flight (nToF) measurements with pulsed neutron beams, and (b) material activation-irradiation with high-energy high-intensity neutron fluxes. Thanks to the high energy and high intensity neutron flux available, SPIRAL-2 offers a unique opportunity for material irradiations both for fission and fusion related research, tests of various detection systems and of resistance of electronics components to irradiations, etc. SPIRAL-2 also could be considered as an intermediate step towards new generation dedicated irradiation facilities as IFMIF previewed only beyond 2015. Equally, the interval from 0.1 MeV to 40 MeV for neutron cross section measurements is an energy range that is of particular importance for energy applications, notably accelerator driven systems (ADS) and Gen-IV fast reactors, as well as for fusion related devices. It is also the region where pre-equilibrium approaches are often used to link the low (evaporation) and high energy (intra-nuclear cascade) reaction models. With very intense neutron beams of SPIRAL-2 measurements of very low mass (often radioactive) targets and small cross sections become feasible in short experimental campaigns. Production of radioactive targets for dedicated physics experiments is also an attractive feature of SPIRAL-2. In brief, it was shown that SPIRAL-2 has got a remarkable potential for neutron based research both for fundamental physics and various applications. In addition, in the neutron energy range from a few MeV to, say, 35 MeV this research would have a leading position for the next 10-15 years if compared to other neutron facilities in operation or under construction worldwide. (author)

  20. Report of Activities of the Association Euratom/Ciemat

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    The focal point of the work at the Spanish Association has been the flexible Heliac TJII, which since 2002 is the only stellarator in operation in Europe. The main milestone of TJ-II operation has been the generation of plasmas sustained by NBI heating (which lead to a record in TJ-II stored energy) but significant physics results have been also obtained in the continuation of existing lines (improved confinement scenarios and the role of rational surfaces, iota scaling with boronized walls, turbulence studies, impurity transport and rotation experiments, suprathermal electrons studies, plasma wall effects, etc). TJ-II improvements include the progress in the second NBI, the preparations for the Bernstein wave heating system, the installation of a Diagnostic NB and the fast camera (Ha) diagnostic (on temporal loan from PPPL- Princeton). Other activities of the Association include the Materials research programme, both in the areas of insulator materials properties and structural materials (with a new line open: studies of Tritium barriers during irradiation), the studies on the socio-economic impact of fusion and a reinforced participation in the EFDA technology work programme. The Association wants to increase technology activities and, along this line, a number of expression of interest have been submitted, leading to several task contracts : design of the European Dipole, design of the magnet for ITER field simulation on NBI test bed, IFMIF security analysis, Demo Blanket support system (finished), Main plasma reflectometry system (finished), Tritium retention/ removal studies. Finally, the Association has keep its involvement in the PhD programme Fusion and Plasma Physics that has been carried since 2001 in collaboration with several Universities and other Spanish research centres. (Author)

  1. Report of Activities of the Association Euratom/Ciemat. Annual Report 2004

    International Nuclear Information System (INIS)

    2005-01-01

    The focal point of the work at the Spanish Association has been the flexible Heliac TJ-II, which at present is the only stellarator in operation in Europe. The main milestone of TJ-II operation has been the generation of plasmas sustained by NBI heating (which lead to a record in TJ-II stored energy) but significant physics results have been also obtained in the continuation of existing lines (improved confinement scenarios and the role of rational surfaces, iota scaling with boronized walls, turbulence studies, impurity transport and rotation experiments, suprathermal electrons studies, plasma wall effects). TJ-II improvements include the progress in the second NBI, the preparations for the Bernstein wave heating system, the installation of a Diagnostic NB and the fast camera (Ha) diagnostic (on temporal loan from PPPL- Princeton). Other activities of the Association include the Materials research programme, both in the areas of insulator materials properties and structural materials (with a new line open: studies of Tritium barriers during irradiation), the studies on the socio-economic impact of fusion and a reinforced participation in the EFDA technology work programme. The Association wants to increase technology activities and, along this line, a number of expression of interest have been submitted, leading to several task contracts : design of the European Dipole, design of the magnet for ITER field simulation on NBI test bed, IFMIF security analysis, Demo Blanket support system (finished), Main plasma reflectometry system (finished), Tritium retention/ removal studies. Finally, the Association has keep its involvement in the PhD programme Fusion and Plasma Physics that has been carried since 2001 in collaboration with several Universities and other Spanish research centres. (Author)

  2. Report of Activities of the Association Euratom/Ciemat. Annual Report 2004

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    The focal point of the work at the Spanish Association has been the flexible Heliac TJ-II, which at present is the only stellarator in operation in Europe. The main milestone of TJ-II operation has been the generation of plasmas sustained by NBI heating (which lead to a record in TJ-II stored energy) but significant physics results have been also obtained in the continuation of existing lines (improved confinement scenarios and the role of rational surfaces, iota scaling with boronized walls, turbulence studies, impurity transport and rotation experiments, suprathermal electrons studies, plasma wall effects). TJ-II improvements include the progress in the second NBI, the preparations for the Bernstein wave heating system, the installation of a Diagnostic NB and the fast camera (Ha) diagnostic (on temporal loan from PPPL- Princeton). Other activities of the Association include the Materials research programme, both in the areas of insulator materials properties and structural materials (with a new line open: studies of Tritium barriers during irradiation), the studies on the socio-economic impact of fusion and a reinforced participation in the EFDA technology work programme. The Association wants to increase technology activities and, along this line, a number of expression of interest have been submitted, leading to several task contracts : design of the European Dipole, design of the magnet for ITER field simulation on NBI test bed, IFMIF security analysis, Demo Blanket support system (finished), Main plasma reflectometry system (finished), Tritium retention/ removal studies. Finally, the Association has keep its involvement in the PhD programme Fusion and Plasma Physics that has been carried since 2001 in collaboration with several Universities and other Spanish research centres. (Author)

  3. Report of Activities of the Association Euratom/Ciemat

    International Nuclear Information System (INIS)

    2006-01-01

    The focal point of the work at the Spanish Association has been the flexible Heliac TJII, which since 2002 is the only stellarator in operation in Europe. The main milestone of TJ-II operation has been the generation of plasmas sustained by NBI heating (which lead to a record in TJ-II stored energy) but significant physics results have been also obtained in the continuation of existing lines (improved confinement scenarios and the role of rational surfaces, iota scaling with boronized walls, turbulence studies, impurity transport and rotation experiments, suprathermal electrons studies, plasma wall effects, etc). TJ-II improvements include the progress in the second NBI, the preparations for the Bernstein wave heating system, the installation of a Diagnostic NB and the fast camera (Ha) diagnostic (on temporal loan from PPPL- Princeton). Other activities of the Association include the Materials research programme, both in the areas of insulator materials properties and structural materials (with a new line open: studies of Tritium barriers during irradiation), the studies on the socio-economic impact of fusion and a reinforced participation in the EFDA technology work programme. The Association wants to increase technology activities and, along this line, a number of expression of interest have been submitted, leading to several task contracts : design of the European Dipole, design of the magnet for ITER field simulation on NBI test bed, IFMIF security analysis, Demo Blanket support system (finished), Main plasma reflectometry system (finished), Tritium retention/ removal studies. Finally, the Association has keep its involvement in the PhD programme Fusion and Plasma Physics that has been carried since 2001 in collaboration with several Universities and other Spanish research centres. (Author)

  4. Development of a zonal applicability tool for remote handling equipment in DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Madzharov, Vladimir, E-mail: vladimir.madzharov@kit.edu [Karlsruhe Institute of Technology, Institute for Material Handling and Logistics, Karlsruhe (Germany); Mittwollen, Martin [Karlsruhe Institute of Technology, Institute for Material Handling and Logistics, Karlsruhe (Germany); Leichtle, Dieter [Fusion for Energy F4E, Barcelona (Spain); Hermon, Gary [Culham Center for Fusion Energy, Culham Science Centre, OX14 3DB Abingdon (United Kingdom)

    2015-10-15

    Highlights: • Radiation-hardness assessment of remote handling (RH) components used in DEMO. • A radiation assessment tool for supporting remote handling engineers. • Connecting data from the radiation field analysis to the radiation hardness data. • Output is the expected lifetime of the selected RH component used for maintenance. - Abstract: A radiation-induced damage caused by the ionizing radiation can induce a malfunctioning of the remote handling equipment (RHE) used during maintenance in fusion power plants, other nuclear power stations and high-energy accelerators facilities like e.g. IFMIF. Therefore to achieve a sufficient length of operational time inside future fusion power plants, a suitable radiation tolerant RHE for maintenance operations in radiation environments is inevitably required. To assess the influence of the radiation on remote handling equipment (RHE), an investigation about radiation hardness assessment of typically used RHE components, has been performed. Additionally, information about the absorbed total dose that every component can withstand before failure was collected. Furthermore, the development of a zonal applicability tool for supporting RHE designers has been started using Excel VBA. The tool connects the data from the radiation field analysis (3-D radiation map) to the radiation hardness data of the planned RHE for DEMO remote maintenance. The intelligent combination of the available information for the radiation behaviour and radiation level at certain time and certain location may help with the taking of decisions about the application of RHE in radiation environment. The user inputs the following parameters: the specific device used in the RHE, the planned location and the maintenance period. The output is the expected lifetime of the selected RHE component at the given location and maintenance period. Planned action times have to be also considered. After having all the parameters it can be decided, if specific RHE

  5. Numerical analysis of high-speed Lithium jet flow under vacuum conditions

    International Nuclear Information System (INIS)

    Gordeev, Sergej; Groeschel, Friedrich; Stieglitz, Robert

    2016-01-01

    The EVEDA Li test loop (ELTL) [1] is aimed at validating the hydraulic stability of the Lithium (Li) target at a velocity up to 20 m/s at vacuum (≈10 −3 Pa). The ELTL has been designed to demonstrate the feasibility of the major components providing a neutron production liquid Li target for IFMIF. The rectangular shaped Li jet (cross-section 25 mm × 100 mm) necessitates for heat removal flow velocities of 15–20 m/s along a concave shaped back wall (curvature radius 250 mm) towards the outlet pipe, where the Li jet is subjected to vacuum before it finally enters the collecting quench tank. During the validation experiments within the ELTL acoustic waves within the target outlet pipe have been recorded, indicating potential cavitation processes in the jet impinging region, which may cause premature failures. In order to identify potential cavitation phenomena in correlation with the free jet flow in the outlet pipe a numerical study has been performed. The comparison measured and simulated acoustic emissions exhibits that experimentally deduced cavitation area coincides with the location of the jet wall impingement. The simulations further reveal that a part of the fluid after striking the wall even flows opposite to the gravity vector. This reversed flow is inherently unstable and characterized by waves at first growing and then bursting into droplets. The intense generation of small droplets increases significantly the Li free surface area and lead to a production of Li vapour, which is captured by the jet flow and reintroduced in the main flow. As the static pressure is recovered downstream due to jet impact, the vapour bubbles collapse and hence cavitation likely occurs.

  6. The cross-section data from neutron activation experiments on niobium in the NPI p-7Li quasi-monoenergetic neutron field

    Directory of Open Access Journals (Sweden)

    Simakov S.P.

    2010-10-01

    Full Text Available The reaction of protons on 7Li target produces the high-energy quasi- monoenergetic neutron spectrum with the tail to lower energies. Proton energies of 19.8, 25.1, 27.6, 30.1, 32.6, 35.0 and 37.4 MeV were used to obtain quasi-monoenergetic neutrons with energies of 18, 21.6, 24.8, 27.6, 30.3, 32.9 and 35.6 MeV, respectively. Nb cross-section data for neutron energies higher than 22.5 MeV do not exist in the literature. Nb is the important material for fusion applications (IFMIF as well. The variable-energy proton beam of NPI cyclotron is utilized for the production of neutron field using thin lithium target. The carbon backing serves as the beam stopper. The system permits to produce neutron flux density about 109  n/cm2/s in peak at 30 MeV neutron energy. The niobium foils of 15 mm in diameter and approx. 0.75 g weight were activated. The nuclear spectroscopy methods with HPGe detector technique were used to obtain the activities of produced isotopes. The large set of neutron energies used in the experiment allows us to make the complex study of the cross-section values. The reactions (n,2n, (n,3n, (n,4n, (n,He3, (n,α and (n,2nα are studied. The cross-sections data of the (n,4n and (n,2nα are obtained for the first time. The cross-sections of (n,2n and (n,α reactions for higher neutron energies are strongly influenced by low energy tail of neutron spectra. This effect is discussed. The results are compared with the EAF-2007 library.

  7. Design of intense neutron source for fusion material study and the role of universities

    International Nuclear Information System (INIS)

    Ishino, Shiori

    1993-01-01

    Need and requirement for the intense neutron source for fusion materials study have been discussed for many years. Recently, international climate has been becoming gradually maturing to consider this problem more seriously because of the recognition of crucial importance of solving materials problems for fusion energy development. The present symposium was designed to discuss the problems associated with the intense neutron source for material irradiation studies which will have a potential for the National Institute for Fusion Science to become one of the important future research areas. The symposium comprises five sessions; first, the role of materials research in fusion development strategies was discussed followed by a brief summary of current IFMIF (International Fusion Materials Irradiation Facility) activity. Despite the pressing need for intense fusion neutron source, currently available neutron sources are reactor or accelerator based sources of which FFTF and LASREF were discussed. Then, various concepts of intense neutron source candidates were presented including ESNIT, which are currently under design by JAERI. In the fourth session, discussions were made on the study of materials with the intense neutron source from the viewpoint of materials scientists and engineers as the user of the facility. This is followed by discussions on the role of universities from the two stand points, namely, fusion irradiation studies and fusion materials development. Finally summary discussions were made by the participants, indicating important role fundamental studies in universities for the full utilization of irradiation data and the need of pure 14 MeV neutron source for fundamental studies together with the intense surrogate neutron sources. (author)

  8. Integrated Approach to Dense Magnetized Plasmas Applications in Nuclear Fusion Technology. Report of a Coordinated Research Project 2007-2011

    International Nuclear Information System (INIS)

    2013-04-01

    Through its coordinated research activities, the IAEA promotes the development and application of nuclear technologies in Member States. The scientific and technical knowledge required for the construction and operation of large nuclear fusion research facilities, including ITER and the Laser Megajoule in France, and the Z machine and the National Ignition Facility in the United States of America, necessitates several accompanying research and development programmes in physics and technology. This is particularly true in the areas of materials science and fusion technology. Hence, the long standing IAEA effort to conduct coordinated research projects (CRPs) in these areas is aimed at: (i) the development of appropriate technical tools to investigate the issue of materials damage and degradation in a fusion plasma environment; and (ii) the emergence of a knowledge based understanding of the various processes underlying materials damage and degradation, thereby leading to the identification of suitable candidate materials fulfilling the stringent requirements of a fusion environment in any next step facility. Dense magnetized plasma (DMP) devices serve as a first test bench for testing of fusion relevant plasma facing materials, diagnostic development and calibration, technologies and scaling to conceptual principles of larger devices while sophisticated testing facilities such as the International Fusion Materials Irradiation Facility (IFMIF) are being designed. The CRP on Integrated Approach to Dense Magnetized Plasmas Applications in Nuclear Fusion Technology described herein was initiated in 2007 with the participation of 12 research institutions in 8 Member States and was concluded in 2011. It was designed with specific research objectives falling into two main categories: support to mainstream fusion research and development of DMP technology. This publication is a compilation of the individual reports submitted by the 12 CRP participants. These reports discuss

  9. Investigation and prevention of droplet splashing during operation of a sodium free jet flow

    International Nuclear Information System (INIS)

    Stoppel, L.; Gordeev, S.; Wetzel, T.; Fellmoser, F.; Daubner, M.

    2010-01-01

    Many accelerator application concepts consider liquid metal as a windowless target, at which the particle beam does directly hit the liquid. One of such concepts is studied in the European project ''DIRAC-Secondary beams - Design Study''. This project is focused on the preliminary research work for construction of a new international particle accelerator - Facility for Antiproton and Ion Research (FAIR) in Darmstadt. The planned accelerator is aimed to work with high-energy heavy-ions, such as U 238 . One of the key elements of the FAIR facility is a liquid-metal-target, made in the form of a rectangular shaped Lithium jet aligned with the gravity vector. In the course of preliminary investigations the theoretical and practical conditions for a stable liquid-metal-jet conforming to the FAIR-requirements have been studied in the Karlsruhe Liquid Metal Laboratory (KALLA) sodium facility. The acquired scientific and technological results can be transferred to liquid-metal targets in nuclear applications, for example, the IFMIF-Target for the study of fusion reactor materials and the Myrrah/XT-ADS target. The main goals of the KALLA-part of the project were to design and build a facility for experimental research on hydrodynamic phenomena of the free surface liquid metal flow as well as to look at technological problems influencing the hydrodynamic stability of such flows. One of such problems emerged already during the startup of the facility: Splashing of liquid metal drops in the vacuum volume of the target box. As a result of such splashing process, liquid metal droplets are accumulated on various internal constructional elements of the target box, for example, on the inspection windows. This effect prohibits long term operation with the facility. The present paper describes the methods used to reduce the splashing to a minimum. (orig.)

  10. Numerical analysis of high-speed Lithium jet flow under vacuum conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, Sergej, E-mail: sergej.gordeev@kit.edu; Groeschel, Friedrich; Stieglitz, Robert

    2016-11-01

    The EVEDA Li test loop (ELTL) [1] is aimed at validating the hydraulic stability of the Lithium (Li) target at a velocity up to 20 m/s at vacuum (≈10{sup −3} Pa). The ELTL has been designed to demonstrate the feasibility of the major components providing a neutron production liquid Li target for IFMIF. The rectangular shaped Li jet (cross-section 25 mm × 100 mm) necessitates for heat removal flow velocities of 15–20 m/s along a concave shaped back wall (curvature radius 250 mm) towards the outlet pipe, where the Li jet is subjected to vacuum before it finally enters the collecting quench tank. During the validation experiments within the ELTL acoustic waves within the target outlet pipe have been recorded, indicating potential cavitation processes in the jet impinging region, which may cause premature failures. In order to identify potential cavitation phenomena in correlation with the free jet flow in the outlet pipe a numerical study has been performed. The comparison measured and simulated acoustic emissions exhibits that experimentally deduced cavitation area coincides with the location of the jet wall impingement. The simulations further reveal that a part of the fluid after striking the wall even flows opposite to the gravity vector. This reversed flow is inherently unstable and characterized by waves at first growing and then bursting into droplets. The intense generation of small droplets increases significantly the Li free surface area and lead to a production of Li vapour, which is captured by the jet flow and reintroduced in the main flow. As the static pressure is recovered downstream due to jet impact, the vapour bubbles collapse and hence cavitation likely occurs.

  11. LIPAc personnel protection system for realizing radiation licensing conditions on injector commissioning with deuteron beam

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Hiroki, E-mail: takahashi.hiroki@jaea.go.jp [IFMIF/EVEDA Accelerator Group, Japan Atomic Energy Agency (JAEA), Rokkasho, Aomori (Japan); Narita, Takahiro; Kasugai, Atsushi [IFMIF/EVEDA Accelerator Group, Japan Atomic Energy Agency (JAEA), Rokkasho, Aomori (Japan); Kojima, Toshiyuki [Gitec Co. Ltd., Hachinohe, Aomori (Japan); Marqueta, Alvaro; Nishiyama, Koichi [IFMIF/EVEDA Project Team, Rokkasho, Aomori (Japan); Sakaki, Hironao [Quantum Beam Science Center, JAEA, Kizu, Kyoto (Japan); Gobin, Raphael [Commissariat à l’Energie Atomique et aux Energies Alternatives, CEA/Saclay, DSM/IRFU, Gif/Yvette (France)

    2016-11-01

    Highlights: • Personnel Protection System (PPS) is developed to adapt the radiation licensing. • PPS achieves the target performance to secure the personnel safety. • Pulse Duty Management System (PDMS) is developed to manage the beam-operation-time. • Satisfying performance of PDMS is confirmed by injector operation with H+ beam. • By the result of PPS and PDMS tests, the radiation license was successfully obtained. - Abstract: The performance validation of the Linear IFMIF Prototype Accelerator (LIPAc), up to the energy of 9 MeV deuteron beam with 125 mA continuous wave (CW), is planned in Rokkasho, Japan. There are three main phases of LIPAc performance validation: Injector commissioning, RFQ commissioning and LIPAc commissioning. Injector commissioning was started by H{sup +} and D{sup +} beam. To apply the radiation licensing for the Injector commissioning, the entering/leaving to/from accelerator vault should be under control, and access to the accelerator vault has to be prohibited for any person during the beam operation. The Personnel Protection System (PPS) was developed to adapt the radiation licensing conditions. The licensing requests that PPS must manage the accumulated D{sup +} current. So, to manage the overall D{sup +} beam time during injector operation, Pulse Duty Management System (PDMS) was developed as a configurable subsystem as part of the PPS. The PDMS was tested during H{sup +} beam (as simulated D{sup +}) operation, to confirm that it can handle the beam inhibit from Injector before the beam accumulation is above the threshold value specified in the radiation licensing condition. In this paper, the design and configuration of these systems and the result of the tests are presented.

  12. Vacuum-Flex Figuring of Primary Telescope Mirrors

    Science.gov (United States)

    Albin, E. F. M.

    2004-12-01

    In the current investigation, details on the construction and performance of a vacuum-flexed (i.e., figured) 51 cm (20-inch) mirror, with a fast f/4 focal ratio, are presented. A vacuum has the chief advantage of being able to pull with a uniform or isotropic stress across a large surface area, which will naturally form a parabolic surface. The essence of the idea is to grind and polish a spherical mirror and then warp or flex it into a near perfect paraboloid, thus avoiding tedious figuring altogether. To date, telescope makers around the globe have experimented with small flexed mirrors with considerable success. In these instances, mirrors have been flexed by exerting tension on a bolt or sponge-pad adhered to the back of the mirror. The prototype mirror consists of two 51 cm disks of plate glass -- each slumped to an f/4 focal ratio. The front-plate (19 mm in thickness) is separated from the back-plate (13 mm in thickness) back a flexible 9.5 mm air filled gasket. Although the rubber gasket makes a fairly good vacuum seal, silicon cement was placed about the outer edge in order to produce a perfectly tight seal. A vacuum of 8 kPa on the back of the mirror resulted in approximately 164 kilograms of negative pressure, which is required to flex the mirror into the required paraboloid. Ronchi test show a nice smooth paraboloid free from astigmatism while foucault zonal measurements display a figure better than 1/20 wave. Preliminary star testing show promising results as well. Vacuum-flexed mirrors may have benefits for both amateur and professional telescope makers alike. A US patent is pending on the aforementioned design.

  13. A theoretical analysis of the impact of atmospheric parameters on the spectral, electrical and thermal performance of a concentrating III–V triple-junction solar cell

    International Nuclear Information System (INIS)

    Theristis, Marios; Fernández, Eduardo F.; Stark, Cameron; O’Donovan, Tadhg S.

    2016-01-01

    Highlights: • An integrated spectral dependent electrical–thermal model has been developed. • The effect of atmospheric parameters on system’s performance is evaluated. • The HCPV cooling requirements under “hot & dry” conditions are quantified. • Case studies show the impact of heat transfer coefficient on annual energy yield. • The integrated modelling allows the system’s optimisation. - Abstract: The spectral sensitivity of a concentrating triple-junction (3J) solar cell has been investigated. The atmospheric parameters such as the air mass (AM), aerosol optical depth (AOD) and precipitable water (PW) change the distribution of the solar spectrum in a way that the spectral, electrical and thermal performance of a 3J solar cell is affected. In this paper, the influence of the spectral changes on the performance of each subcell and whole cell has been analysed. It has been shown that increasing the AM and AOD have a negative impact on the spectral and electrical performance of 3J solar cells while increasing the PW has a positive effect, although, to a lesser degree. A three-dimensional finite element analysis model is used to quantify the effect of each atmospheric parameter on the thermal performance for a range of heat transfer coefficients from the back-plate to the ambient air and also ambient temperature. It is shown that a heat transfer coefficient greater than 1300 W/(m"2 K) is required to keep the solar cell under 100 °C at all times. In order to get a more realistic assessment and also to investigate the effect of heat transfer coefficient on the annual energy yield, the methodology is applied for four US locations using data from a typical meteorological year (TMY3).

  14. Diagnostic procedure on brake pad assembly based on Young's modulus estimation

    International Nuclear Information System (INIS)

    Chiariotti, P; Santolini, C; Tomasini, E P; Martarelli, M

    2013-01-01

    Quality control of brake pads is an important issue, since the pad is a key component of the braking system. Typical damage of a brake pad assembly is the pad–backing plate detachment that affects and modifies the mechanical properties of the whole system. The most sensitive parameter to the damage is the effective Young's modulus, since the damage induces a decrease of the pad assembly stiffness and therefore of its effective Young's modulus: indeed its variation could be used for diagnostic purposes. The effective Young's modulus can be estimated from the first bending resonance frequency identified from the frequency response function measured on the pad assembly. Two kinds of excitation methods, i.e. conventional impulse excitation and magnetic actuation, will be presented and two different measurement sensors, e.g. laser Doppler vibrometer and microphone, analyzed. The robustness of the effective Young's modulus as a diagnostic feature will be demonstrated in comparison to the first bending resonance frequency, which is more sensitive to geometrical dimensions. Variability in the sample dimension, in fact, will induce a variation of the resonance frequency which could be mistaken for damage. The diagnostic approach has been applied to a set of undamaged and damaged pad assemblies showing good performance in terms of damage identification. The environmental temperature can be an important interfering input for the diagnostic procedure, since it influences the effective Young's modulus of the assembly. For that reason, a test at different temperatures in the range between 15 °C and 30 °C has been performed, evidencing that damage identification technique is efficient at any temperature. The robustness of the Young's modulus as a diagnostic feature with respect to damping is also presented. (paper)

  15. Alignment of the TFTR bumper limiter

    International Nuclear Information System (INIS)

    Barnes, G.W.; Owens, D.K.; Loesser, G.D.; Ulrickson, M.

    1989-01-01

    The TFTR Bumper Limiter (BL) is an axisymmetric toroidal limiter mounted on the inner wall of the vacuum vessel. It subtends 120 degree poloidally and has a surface area of 22 m 2 . The plasma facing surface consists of 1,000 kg of graphite tiles mounted on watercooled Inconel backing plates. During the initial installation in the Spring of 1985, the limiter surface was aligned to the toroidal magnetic field by mechanical and magnetic measurements to an estimated accuracy of ±2 mm. During subsequent operation, especially in the 1988 run period in which 30 MW of Neutral Beam Injection routinely occurred, several tiles at points on the limiter which protruded slightly into the plasma were severely damaged. The damage, cracked and spalled tiles, is believed to be initiated by high energy disruptions and aggravated by normal high power operation. The damage pattern and temperature rise during normal operation are consistent with this interpretation. A vacuum vessel opening to replace the damaged tiles and realign the limiter was required. The bumper limiter was reshaped to be circular to ±0.5 mm at the midplane by means of mechanical measurements in order to better distribute the heat loads and eliminate hot spots. The ±0.5 mm accuracy is determined by the variation in individual tile thickness which is ±0.5 mm. This paper describes the methods used to mechanically align the limiter and presents evidence based on machine operation with plasma that the limiter is reasonably well aligned with the toroidal field. Future work dealing with the alignment of the total limiter to the toroidal field using mechanical and magnetic measurements and the replacement of a subset of the carbon tiles with carbon-carbon composite material is also discussed. 7 refs., 4 figs

  16. Structural Inference in the Art of Violin Making.

    Science.gov (United States)

    Morse-Fortier, Leonard Joseph

    The "secrets" of success of early Italian violins have long been sought. Among their many efforts to reproduce the results of Stradiveri, Guarneri, and Amati, luthiers have attempted to order and match natural resonant frequencies in the free violin plates. This tap-tone plate tuning technique is simply an eigenvalue extraction scheme. In the final stages of carving, the violin maker complements considerable intuitive knowledge of violin plate structure and of modal attributes with tap-tone frequency estimates to better understand plate structure and to inform decisions about plate carving and completeness. Examining the modal attributes of violin plates, this work develops and incorporates an impulse-response scheme for modal inference, measures resonant frequencies and modeshapes for a pair of violin plates, and presents modeshapes through a unique computer visualization scheme developed specifically for this purpose. The work explores, through simple examples questions of how plate modal attributes reflect underlying structure, and questions about the so -called evolution of modeshapes and frequencies through assembly of the violin. Separately, the work develops computer code for a carved, anisotropic, plate/shell finite element. Solutions are found to the static displacement and free-vibration eigenvalue problems for an orthotropic plate, and used to verify element accuracy. Finally, a violin back plate is modelled with full consideration of plate thickness and arching. Model estimates for modal attributes compare very well against experimentally acquired values. Finally, the modal synthesis technique is applied to predicting the modal attributes of the violin top plate with ribs attached from those of the top plate alone, and with an estimate of rib mass and stiffness. This last analysis serves to verify the modal synthesis method, and to quantify its limits of applicability in attempting to solve problems with severe structural modification. Conclusions

  17. Engineering design and thermal hydraulics of plasma facing components of SST-1

    International Nuclear Information System (INIS)

    Pragash, N. Ravi; Chaudhuri, P.; Santra, P.; Chenna Reddy, D.; Khirwadkar, S.; Saxena, Y.C.

    2001-01-01

    SST-1 is a medium size tokamak with super conducting magnetic field coils. All the subsystems of SST-1 are designed for quasi steady state (∼1000 s) operation. Plasma Facing Components (PFCs) of SST-1 consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be compatible for steady state operation. As SST-1 is designed to run double null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. All the PFC are made of copper alloys (CuCrZr and CuZr) on which graphite tiles are mechanically attached. These copper alloy back plates are actively cooled with water flowing in the channels grooved on them with the main consideration in the design of PFCs as the steady state heat removal of about 1.0 MW/m 2 . In addition to be able to remove high heat fluxes, the PFCs are also designed to be compatible for baking at 350 degree sign C. Extensive studies, involving different flow parameters and various cooling layouts, have been done to select the final cooling parameters and layout. Thermal response of the PFCs and vacuum vessel during baking, has been calculated using a FORTRAN code and a 2-D finite element analysis. The PFCs and their supports are also designed to withstand large electro-magnetic forces. Finite element analysis using ANSYS software package is used in this and other PFCs design. The engineering design including thermal hydraulics for cooling and baking of all the PFCs is completed. Poloidal limiters are being fabricated. The remaining PFCs, viz. divertors, stabilizers and baffles are likely to go for fabrication in the next few months. The detailed engineering design, the finite element calculations in the structural and thermal designs are presented in this paper

  18. Mechanical performance of SiC based MEMS capacitive microphone for ultrasonic detection in harsh environment

    Science.gov (United States)

    Zawawi, S. A.; Hamzah, A. A.; Mohd-Yasin, F.; Majlis, B. Y.

    2017-08-01

    In this project, SiC based MEMS capacitive microphone was developed for detecting leaked gas in extremely harsh environment such as coal mines and petroleum processing plants via ultrasonic detection. The MEMS capacitive microphone consists of two parallel plates; top plate (movable diaphragm) and bottom (fixed) plate, which separated by an air gap. While, the vent holes were fabricated on the back plate to release trapped air and reduce damping. In order to withstand high temperature and pressure, a 1.0 μm thick SiC diaphragm was utilized as the top membrane. The developed SiC could withstand a temperature up to 1400°C. Moreover, the 3 μm air gap is invented between the top membrane and the bottom plate via wafer bonding. COMSOL Multiphysics simulation software was used for design optimization. Various diaphragms with sizes of 600 μm2, 700 μm2, 800 μm2, 900 μm2 and 1000 μm2 are loaded with external pressure. From this analysis, it was observed that SiC microphone with diaphragm width of 1000 μm2 produced optimal surface vibrations, with first-mode resonant frequency of approximately 36 kHz. The maximum deflection value at resonant frequency is less than the air gap thickness of 8 mu;m, thus eliminating the possibility of shortage between plates during operation. As summary, the designed SiC capacitive microphone has high potential and it is suitable to be applied in ultrasonic gas leaking detection in harsh environment.

  19. 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

  20. A comparison of wood density between classical Cremonese and modern violins.

    Directory of Open Access Journals (Sweden)

    Berend C Stoel

    Full Text Available Classical violins created by Cremonese masters, such as Antonio Stradivari and Giuseppe Guarneri Del Gesu, have become the benchmark to which the sound of all violins are compared in terms of their abilities of expressiveness and projection. By general consensus, no luthier since that time has been able to replicate the sound quality of these classical instruments. The vibration and sound radiation characteristics of a violin are determined by an instrument's geometry and the material properties of the wood. New test methods allow the non-destructive examination of one of the key material properties, the wood density, at the growth ring level of detail. The densities of five classical and eight modern violins were compared, using computed tomography and specially developed image-processing software. No significant differences were found between the median densities of the modern and the antique violins, however the density difference between wood grains of early and late growth was significantly smaller in the classical Cremonese violins compared with modern violins, in both the top (Spruce and back (Maple plates (p = 0.028 and 0.008, respectively. The mean density differential (SE of the top plates of the modern and classical violins was 274 (26.6 and 183 (11.7 gram/liter. For the back plates, the values were 128 (2.6 and 115 (2.0 gram/liter. These differences in density differentials may reflect similar changes in stiffness distributions, which could directly impact vibrational efficacy or indirectly modify sound radiation via altered damping characteristics. Either of these mechanisms may help explain the acoustical differences between the classical and modern violins.

  1. Dual energy radiography using active detector technology

    International Nuclear Information System (INIS)

    Seibert, J.A.; Poage, T.F.; Alvarez, R.E.

    1996-01-01

    A new technology has been implemented using an open-quotes active-detectorclose quotes comprised of two computed radiography (CR) imaging plates in a sandwich geometry for dual-energy radiography. This detector allows excellent energy separation, short exposure time, and high signal to noise ratio (SNR) for clinically robust open-quotes bone-onlyclose quotes and open-quotes soft-tissue onlyclose quotes images with minimum patient motion. Energy separation is achieved by two separate exposures at widely different kVp's: the high energy (120 kVp + 1.5 mm Cu filter) exposure is initiated first, followed by a short burst of intense light to erase the latent image on the front plate, and then a 50 kVp (low energy) exposure. A personal computer interfaced to the x-ray generator, filter wheel, and active detector system orchestrates the acquisition sequence within a time period of 150 msec. The front and back plates are processed using a CR readout algorithm with fixed speed and wide dynamic range. open-quotes Bone-onlyclose quotes and open-quotes soft-tissue onlyclose quotes images are calculated by geometric alignment of the two images and application of dual energy decomposition algorithms on a pixel by pixel basis. Resultant images of a calibration phantom demonstrate an increase of SNR 2 / dose by ∼73 times when compared to a single exposure open-quotes passive-detectorclose quotes comprised of CR imaging plates, and an ∼8 fold increase compared to a screen-film dual-energy cassette comprised of different phosphor compounds. In conclusion, dual energy imaging with open-quotes active detectorclose quotes technology is clinically feasible and can provide substantial improvements over conventional methods for dual-energy radiography

  2. R and D activities for the design of the MITICA Plasma Driver Plate manufacturing process via explosion bonding technique

    International Nuclear Information System (INIS)

    Pavei, M.; Dal Bello, S.; Groeneveld, H.; Rizzolo, A.

    2013-01-01

    Highlights: ► The work is focused on the manufacturing process of the Plasma Driver Plate of MITICA. ► A clad plate of molybdenum and copper has been manufactured. ► Simulations have been carried out to improve the design geometry of the component. ► The driver-hole rim have been machined and hot formed. ► No delamination were found in the molybdenum. -- Abstract: The back plate of the MITICA plasma source, named Plasma Driver Plate (PDP), will be protected from the impact of the highly energetic back-streaming positive ions (BSI+), generated inside the accelerator, by a 1.0 mm thick molybdenum layer that will be joined by Explosion Bonding (EB) technique to the copper heat sink. This technology has been investigated and used for manufacturing prototypes, demonstrating very high strength of the obtained molybdenum–copper interface. The production of the shaped edge profile of the driver-hole, after the EB, is an open point. In order to demonstrate the possibility to produce the PDP by explosion bonding, the manufacturing of a full scale prototype of the area just around one of the PDP driver-holes was identified as the road to address most of the manufacturing issues. Elasto-plastic finite element analyses have been carried out to improve the hole rim geometry and the process parameters of all the manufacturing steps. A full scale prototype of the PDP driver-hole has been manufactured and tested. This contribution gives an overview of the R and D activities carried out to address the main open issues, to define the PDP component detailed geometry and its manufacturing processes, via EB technique

  3. On the NBI system for substantial current drive in a fusion power plant: Status and R and D needs for ion source and laser neutralizer

    International Nuclear Information System (INIS)

    Franzen, P.; Fantz, U.

    2014-01-01

    Highlights: • NBI is a candidate for a cw tokamak DEMO due to its high current drive efficiency. • The plug-in efficiency must be improved from the present 20–30% to more than 50%. • A suitable candidate is a photo neutralizer with almost 100% neutralization efficiency; basic feasibility studies are underway. • Cw operation with a large availability puts rather high demands on source operation with some safety margins, especially for the components with high power density loads (source back plate and extraction system). • Alternatives to the present use of cesium are under exploitations. - Abstract: The requirements for the heating and current drive systems of a fusion power plant will strongly depend on the DEMO scenario. The paper discusses the R and D needs for a neutral beam injection system — being a candidate due to the highest current drive efficiency — for the most demanding scenario, a steady state tokamak DEMO. Most important issues are the improvement of the wall-plug efficiency from the present ∼25% to the required 50–60% by improving the neutralization efficiency with a laser neutralizer system and the improvement of the reliability of the ion source operation. The demands on and the potential of decreasing the ion source operation pressure, as well as decreasing the amount of co-extracted electrons and backstreaming ions are discussed using the ITER requirements and solutions as basis. A further concern is the necessity of cesium for which either the cesium management must be improved or alternatives to cesium for the production of negative ions have to be identified

  4. Progress on DEMO blanket attachment concept with keys and pins

    International Nuclear Information System (INIS)

    Vizvary, Zsolt; Iglesias, Daniel; Cooper, David; Crowe, Robert; Riccardo, Valeria

    2015-01-01

    Highlights: • DEMO blanket attachment system with keys and pins (without using bolts). • Blanket segments are preloaded by progressively designed springs. • Blanket back plate flexibility has a major impact on spring design. • Mechanical analysis of other components indicates no unresolvable issues. • Thermal analysis indicates acceptable temperatures for the support system. - Abstract: The blanket attachment has to cope with gravity, thermal and electromagnetic loads, also it has to be installed and serviced by remote handling. Pre-stressed components suffer from stress relaxation in irradiated environments such as DEMO. To circumvent this problem pre-stressed component should be either avoided or shielded, and where possible keys and pins should be used. This strategy has been proposed for the DEMO multi-module segments (MMS). The blanket segments are held by two tapered keys each, designed to allow thermal expansions while providing contact with the vacuum vessel and to resist the poloidal and radial moments the latter being dominant at 9.1 MNm inboard and 15 MNm outboard. On the top of the blanket segment there is a pin which provides vertical support. At the bottom another vertical support has to lock them in position after installation and manage the pre-load on the segments. The pre-load is required to deal with the electromagnetic loads during disruption. This is provided by a set of springs, which require shielding as they are preloaded. These are sized to cope with the force (3 MN inboard, 1.4 MN outboard) due to halo currents and the toroidal moment which can reverse. Calculations show that the flexibility of the blanket segment itself plays a significant role in defining the required support system. The blanket segment acts as a preloaded spring and it has to be part of the attachment design as well.

  5. Fusion-related work at the Nuclear Energy Agency Data Bank

    International Nuclear Information System (INIS)

    Henriksson, H.; Mompean, F.J.; Kodeli, I.

    2007-01-01

    The OECD Nuclear Energy Agency (NEA) Data Bank is part of an international network of data centres in charge of the compilation and dissemination of basic nuclear reaction data. Through its activities in the reaction data field, the NEA participates in the preparation of data for the modelling of future nuclear facility concepts and the development of reactor installations. A working party at the NEA on international nuclear data evaluation cooperation (WPEC) is established to promote the exchange of nuclear data evaluations, measurements, nuclear model calculations and validation. WPEC provides a framework for co-operative activities, such as the high priority request list for experimental data of special interest for certain applications, such as IFMIF or ITER. The NEA Data Bank administrates the collection and validation as well as the distribution of the Joint Evaluated Fusion and Fission (JEFF) library, where the activities in the European Fusion and Activation File projects (EFF and EAF respectively) play an important role for new data evaluations. The topics cover verification of activation and transport data, calculation methods and validation via integral experiments. The EFF project brings together all available expertise in Europe related to the nuclear data requirements of existing and future fusion devices, and the project contributed greatly to the internationally recognised nuclear data library JEFF-3.1, released in May 2005. The NEA also provides tools for the EFF project, such as computer codes for nuclear energy and radiation physics applications. Of special interest for fusion applications are the integral experiments collected in the Shielding Integral Benchmark Archive Database (SINBAD) database. SINBAD is an internationally established set of radiation shielding and dosimetry data containing over 80 experiments relevant for reactor and accelerator shielding. About 30 of these experiments are dedicated to fusion blanket neutronics. Materials

  6. State-of-the-art 3-D neutronics analysis methods for fusion energy systems

    International Nuclear Information System (INIS)

    Wilson, P.P.H.; Feder, R.; Fischer, U.; Loughlin, M.; Petrizzi, L.; Wu, Y.

    2007-01-01

    -TBM. FZK has used the McCAD interface programme to generate models of the Electron Cyclotron Resonance Heating (ECRH) launcher for integration into the standard ITER MCNP model. UKAEA have carried out design analysis of the RF antenna systems using both Attila and MCNP to determine a number of nuclear responses. Other systems being studied include the ARIES Compact Stellarator, IFMIF, and EAST. The widespread use of these tools in the design and analysis of ITER and other fusion energy systems will enable a more accurate assessment of the nuclear response of individual components, leading to a reduction in design margins, improved overall performance, and new level of quality assurance (QA) since these new tools ensure that engineering and analysis models are consistent. (orig.)

  7. EURATOM strategy towards fusion energy

    International Nuclear Information System (INIS)

    Varandas, C.

    2007-01-01

    Research and development (Research and Development) activities in controlled thermonuclear fusion have been carried out since the 60's of the last century aiming at providing a new clean, powerful, practically inexhaustive, safe, environmentally friend and economically attractive energy source for the sustainable development of our society.The EURATOM Fusion Programme (EFP) has the leadership of the magnetic confinement Research and Development activities due to the excellent results obtained on JET and other specialized devices, such as ASDEX-Upgrade, TORE SUPRA, FTU, TCV, TEXTOR, CASTOR, ISTTOK, MAST, TJ-II, W7-X, RFX and EXTRAP. JET is the largest tokamak in operation and the single device that can use deuterium and tritium mixes. It has produced 16 MW of fusion power, during 3 seconds, with an energy amplification of 0.6. The next steps of the EFP strategy towards fusion energy are ITER complemented by a vigorous Accompanying Programme, DEMO and a prototype of a fusion power plant. ITER, the first experimental fusion reactor, is a large-scale project (35-year duration, 10000 MEuros budget), developed in the frame of a very broad international collaboration, involving EURATOM, Japan, Russia Federation, United States of America, Korea, China and India. ITER has two main objectives: (i) to prove the scientific and technical viability of fusion energy by producing 500 MW, during 300 seconds and a energy amplification between 10 and 20; and (ii) to test the simultaneous and integrated operation of the technologies needed for a fusion reactor. The Accompanying Programme aims to prepare the ITER scientific exploitation and the DEMO design, including the development of the International Fusion Materials Irradiation Facility (IFMIF). A substantial part of this programme will be carried out in the frame of the Broader Approach, an agreement signed by EURATOM and Japan. The main goal of DEMO is to produce electricity, during a long time, from nuclear fusion reactions. The

  8. Technology Programme

    Energy Technology Data Exchange (ETDEWEB)

    Batistoni, Paola; De Marco, Francesco; Pieroni, Leonardo [ed.

    2005-07-01

    The technology activities carried out by the Euratom-ENEA Association in the framework of the European Fusion Development Agreement concern the Next Step (International Thermonuclear Experimental Reactor - ITER), the Long-Term Programme (breeder blanket, materials, International Fusion Materials Irradiation Facility - IFMIF), Power Plant Conceptual Studies and Socio-Economic Studies. The Underlying Technology Programme was set up to complement the fusion activities as well to develop technologies with a wider range of interest. The Technology Programme mainly involves staff from the Frascati laboratories of the Fusion Technical and Scientific Unit and from the Brasimone laboratories of the Advanced Physics Technologies Unit. Other ENEA units also provide valuable contributions to the programme. ENEA is heavily engaged in component development/testing and in design and safety activities for the European Fusion Technology Programme. Although the work documented in the following covers a large range of topics that differ considerably because they concern the development of extremely complex systems, the high level of integration and coordination ensures the capability to cover the fusion system as a whole. In 2004 the most significant testing activities concerned the ITER primary beryllium-coated first wall. In the field of high-heat-flux components, an important achievement was the qualification of the process for depositing a copper liner on carbon fibre composite (CFC) hollow tiles. This new process, pre-brazed casting (PBC), allows the hot radial pressing (HRP) joining procedure to be used also for CFC-based armour monoblock divertor components. The PBC and HRP processes are candidates for the construction of the ITER divertor. In the materials field an important milestone was the commissioning of a new facility for chemical vapour infiltration/deposition, used for optimising silicon carbide composite (SiCf/SiC) components. Eight patents were deposited during 2004

  9. The nuclear fusion reactor. How close are we to its realisation

    International Nuclear Information System (INIS)

    Lackner, K.

    2001-01-01

    . Two, material related issues are, however, also on the critical path: development of heat and plasma particle flux resistant materials for contact with the plasma, and neutron fluence tolerant materials for structural functions and the breeding blanket. The further development of the former will proceed on ITER itself, as their performance tests require a plasma environment. ITER will, however, not have sufficient fluence to carry out conclusive nuclear tests, which require the availability of a dedicated test facility with a suitable neutron energy spectrum ( IFMIF). Based on these scenarios we should have the critical physics and technology information for a fusion power plant fully available in 2020

  10. Annual report of Naka Fusion Research Establishment from April 1, 2002 to March 31, 2003

    International Nuclear Information System (INIS)

    Tsuji, Hiroshi; Hamamatsu, Kiyotaka; Matsumoto, Hiroshi; Yoshida, Hidetoshi

    2003-11-01

    technological database to assure the design of fusion power demonstration plants, which include the development of Blanket Test Modules to be tested by ITER, reduced activation structural materials, and their neutron irradiation facility, now called the International Fusion Materials Irradiation Facility (IFMIF). In the ITER Program, Canada made the first site proposal to host ITER in June 2001 and three additional site offers including Japanese Rokkasho proposal were submitted in June 2002. Fourteen years after the inception of ITER, construction of ITER has come close to a reality. JAERI as the main implementation institute of the ITER program in Japan, has made major technical contributions in preparing the Japanese site proposal and licensing procedures. JAERI has also coordinated scientific and technical activities in support of ITER collaborating with universities and other research institutions in Japan. (J.P.N.)

  11. Modelling irradiation effect of EUROFER

    International Nuclear Information System (INIS)

    Boutard, J.-L.; Dudarev, S.; Victoria, M.

    2006-01-01

    irradiating with double (dpa, He) or triple (dpa, He, H) beam and characterising (TEM, AP-FIM, nano-identation) volume of materials of the same order as the ones that can be simulated. The strategy with materials with increasing complexity of chemical composition and initial microstructure will be presented. The resulting modelling tools and associated data base should be used to correlate experimental data from varied irradiation source and to optimize IFMIF testing programme and extrapolate the results with enhanced confidence. (author)

  12. Fusion is urgent needed for the developing countries

    International Nuclear Information System (INIS)

    Li Jiangang

    2005-01-01

    Energy is a global problem, as it is central to economic development, climate and environment, and international stability and sustainability. Energy need is expected to double in 40 years and an even larger increase is needed to lift the world out of poverty. 80% of world's energy is generated by burning fossil fuels, which is driving climate change and generating pollution. China will grow up to be a moderate developed country in 2050. The coal-centred energy structure will remain until 2050. Annual Energy Consumption per person will increase from near 1 TCE to no less than 3 TCE ( at present time, US: 11.5 TCE; West Europe: 5.6 TCE; Japan: 5.1 TCE) Estimated Energy Demand: increasing from near 1B TCE to over 4B TCE within next 3-4 decades. To realize the long-term sustainable development, it is necessary for China to explore reliable ways and develop thousands of GW non- fossil fuel power. The fission energy is a transit solution. To build hundreds of GW Fission Nuclear Power Plants in China - social problems, safety and environmental concerns, technical difficulties should be solved in near future. It is crucial and urgent for China to realize the controlled Nuclear Fusion Energy for our long-term development in the future as early as possible. Fusion shows environmentally responsible and intrinsically safe, the supplies of fuel are essentially limitless. JET has produced 16MW of fusion power and shown that fusion can be mastered on earth. Fusion has a long and successful history of international collaboration with obvious benefits to all partners for peaceful purpose. ITER is a device for us to bring the Sun to earth for the first time in the history. A properly organised and funded fusion development programme could lead to a proto-type fusion power plant to generate electricity to the grid within about 30 years (ITER+IFMIF). For developing countries, such as China and India, fusion is one of the very few options for large-scale sustainable energy generation

  13. On tungsten technologies and qualification for DEMO

    International Nuclear Information System (INIS)

    Laan, J. van der; Hegeman, H.; Wouters, O.; Luzginova, N.; Jonker, B.; Van der Marck, S.; Opschoor, J.; Wang, J.; Dowling, G.; Stuivenga, M.; Carton, E.

    2009-01-01

    products like Re and Os can be simulated using chemically tailored powders, and the evolution of microstructure and properties can be compared with the spectral tailoring. Such approach will allow speed up the tungsten development and qualification prior to the availability of IFMIF and knowledge gained from ITER operation. MTR's are also suited for component testing of primary wall modules, by providing representative loading pattern to joints under simultaneous heat flux and neutron loads. (author)

  14. State-of-the-art 3-D neutronics analysis methods for fusion energy systems

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, P.P.H. [Wisconsin-Madison Univ., Madison, WI (United States); Feder, R. [Princeton Plasma Physics Lab. (United States); Fischer, U. [Forschungszentrum Karlsruhe (Germany); Loughlin, M. [United Kingdom Atomic Energy Authority (United Kingdom); Petrizzi, L. [ENEA-Frascati (Italy); Wu, Y. [Academy of Sciences (China). Inst. of Plasma Physics; Youssef, M. [California Univ., Los Angeles, CA (United States)

    2007-07-01

    -TBM. FZK has used the McCAD interface programme to generate models of the Electron Cyclotron Resonance Heating (ECRH) launcher for integration into the standard ITER MCNP model. UKAEA have carried out design analysis of the RF antenna systems using both Attila and MCNP to determine a number of nuclear responses. Other systems being studied include the ARIES Compact Stellarator, IFMIF, and EAST. The widespread use of these tools in the design and analysis of ITER and other fusion energy systems will enable a more accurate assessment of the nuclear response of individual components, leading to a reduction in design margins, improved overall performance, and new level of quality assurance (QA) since these new tools ensure that engineering and analysis models are consistent. (orig.)

  15. Annual report on major results and progress of Fusion Research and Development Directorate of JAEA from April 1, 2006 to March 31, 2007

    International Nuclear Information System (INIS)

    Suzuki, Satoshi; Ishii, Yasutomo; Sukegawa, Atsuhiko; Iwai, Yasunori; Nakamura, Hiroo; Sugie, Tatsuo

    2008-08-01

    This annual report provides an overview of major results and progress on research and development (R and D) activities at Fusion Research and Development Directorate of Japan Atomic Energy Agency (JAEA) from April 1, 2006 to March 31, 2007, including those performed in collaboration with other research establishments of JAEA, research institutes, and universities. In JT-60, as a result of ferritic steel tiles (FSTs) installation to reduce the toroidal field ripple and the application of the real time current profile control, high boot strap current fraction (∼0.7) has successfully been sustained about 8 s. In addition, the conceptual design of JT-60SA, which was placed as a combined project of JA-EU Satellite Tokamak Programme under the Broader Approach Programme and JAEA's programme for national use, was progressed. In theoretical and analytical researches, studies on ITB events and their triggers, plasma shape effect on edge stability and driven magnetic island evolution in rotating plasmas were progressed. In the NEXT project, computer simulations of the plasma turbulence were progressed. In fusion reactor technologies, R and Ds for ITER and fusion DEMO plants have been carried out. For ITER, a steady state operation of the 170GHz gyrotron up to 10min with 0.82MW was demonstrated. Also current density of the neutral beam injector has been extended to 146A/m 2 at 0.84MeV. In the ITER Test Blanket Module (TBM), designs and R and Ds of Water and Helium Cooled Solid Breeder TBMs including tritium breeder/multiplier materials were progressed. Tritium processing technology for breeding blankets and neutronics integral experiments with a blanket mockup were also progressed. For ITER and DEMO blankets, studies on neutron irradiation effects and ion irradiation effects on F82H steel characteristics were continued using HFIR, TIARA and so on. In the IFMIF program, transitional activities to EVEDA were continued. In the ITER Program, under the framework of the ITER

  16. Nitrogen Recovery by Fe-Ti Alloy from Molten Lithium at High Temperatures

    International Nuclear Information System (INIS)

    Juro Yagi; Akihiro Suzuki; Takayuki Terai; Takeo Muroga

    2006-01-01

    Molten lithium will be used as a beam target of IFMIF (International Fusion Materials Irradiation Facility), and is also expected as a self-cooling and tritium breeding material in fusion reactors. Since tritium is generated in both cases, tritium recovery is required from viewpoints of safety and a feasible fuel cycle. Nitrogen impurity in the lithium, however, not only enhance corrosion to tubing materials, but also promote nitride contamination on a surface of yttrium, which is considered to be a tritium gettering candidate. In our previous study, nitrogen recovery by hot trap method with Fe + 5%Ti alloy as a gettering material showed a higher nitrogen reduction capacity than that with Ti or Cr metal. In this study, high temperature recovery of nitrogen with Fe-Ti alloy was examined to achieve more efficient recovery and higher recovery rate coefficient. Fe - 4%Ti alloy are fabricated by electron beam melting, and its thin plates (40 mm x 10 mm x 1 mm) are used in our experiments. The Fe - 4%Ti alloy plates were immersed into 25 g of liquid lithium in Mo crucible under Ar atmosphere. The crucible was put in a SUS316 stainless steel pot heated at 600, 700, or 800 o C up to 100 hours. A small portion of the liquid lithium in the crucible was sampled out with adequate time interval, and the nitrogen concentrations in the sampled lithium were observed by changing nitrogen to ammonia. Experiments using lithium containing about 100 wt. ppm of nitrogen at the beginning show that the nitrogen reduction became faster with temperature and the minimum achieved nitrogen concentration was less than 20 wppm in case of 800 C. SEM-EDS analysis on the plates after experiment shows a Ti-rich surface layer of tens of micrometers on the alloy immersed in lithium at 800 C, and XPS analysis indicates the surface layer is TiN, while no Ti-rich layer nor TiN were observed on the alloys immersed at 600 o C and 700 o C. By increasing temperature from 600 o C to 800 o C, the diffusion

  17. Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

    International Nuclear Information System (INIS)

    2005-01-01

    facility IFMIF with lithium target. Also, the results of research work performed with the purpose of supporting innovative technologies aimed at the use of liquid metals, e.g., heat pipe applied to nuclear power facilities, lyophobic-capillary porous systems, and others are presented here. As part of the technical program, meeting participants visited test facilities for thermal hydraulics research with liquid metal coolants (sodium, NaK, lead-bismuth, lithium) in the SSC RF IPPE

  18. Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    facility IFMIF with lithium target. Also, the results of research work performed with the purpose of supporting innovative technologies aimed at the use of liquid metals, e.g., heat pipe applied to nuclear power facilities, lyophobic-capillary porous systems, and others are presented here. As part of the technical program, meeting participants visited test facilities for thermal hydraulics research with liquid metal coolants (sodium, NaK, lead-bismuth, lithium) in the SSC RF IPPE.

  19. [An analysis of Spanish biomedical journals by the impact factor].

    Science.gov (United States)

    Baños, J E; Casanovas, L; Guardiola, E; Bosch, F

    1992-06-13

    One of the most frequently used parameters for evaluating scientific publications is that of impact factor (IF) published in the Science Citation Index-Journal Citation Reports (SCI-JCR) which evaluates the number of citations a journal receives on behalf of other journals. The present study analyzed the Spanish biomedical journals included in the SCI-JCR by the IF. The IF were obtained from the SCI-JCR (1980-89). The journals were evaluated by the IF and the weighted impact factor (WIF) calculated according to WIF = (IF/MIF) x 100 in which MIF = maximum IF of the considered area. Nine Spanish biomedical journals were included in the SCI-JCR, four being basic sciences (Histology and Histopathology, Inmunología, Methods and Findings in Experimental and Clinical Pharmacology, Revista Española de Fisiología) and five clinical journals (Allergologia et Immunopathologia, Medicina Clínica, Nefrología, Revista Española de las Enfermedades del Aparato Digestivo, Revista Clínica Española). Their IF were much lower than the most important journals in each area with the mean (+/- standard deviation) being 0.21 +/- 0.22 (range 0.016-0.627). The mean WIF was 2.88 +/- 4.07 (0.16-12.82). The journals of basic sciences had higher IF and WIF than the clinical journals (p less than 0.05). Only the four journals of basic sciences were included in the SCI. Four journals, those of basic sciences, are preferentially or exclusively published in English and other five are published in Spanish. The differences in IF among these groups were not significant (p = 0.06) while those of WIF were significant (p less than 0.05). The number of Spanish biomedical journals in the SCI-JCR has risen from 1 in 1980 to 9 in 1989 with IF which have evolved variably. In mind of impact factor, the contribution of Spanish journals is low, with that of biomedical sciences being higher than that of clinical journals. Language and inclusion in the Science Citation Index may explain, at least in part

  20. Technology Programme

    International Nuclear Information System (INIS)

    Batistoni, Paola; De Marco, Francesco; Pieroni, Leonardo

    2005-01-01

    The technology activities carried out by the Euratom-ENEA Association in the framework of the European Fusion Development Agreement concern the Next Step (International Thermonuclear Experimental Reactor - ITER), the Long-Term Programme (breeder blanket, materials, International Fusion Materials Irradiation Facility - IFMIF), Power Plant Conceptual Studies and Socio-Economic Studies. The Underlying Technology Programme was set up to complement the fusion activities as well to develop technologies with a wider range of interest. The Technology Programme mainly involves staff from the Frascati laboratories of the Fusion Technical and Scientific Unit and from the Brasimone laboratories of the Advanced Physics Technologies Unit. Other ENEA units also provide valuable contributions to the programme. ENEA is heavily engaged in component development/testing and in design and safety activities for the European Fusion Technology Programme. Although the work documented in the following covers a large range of topics that differ considerably because they concern the development of extremely complex systems, the high level of integration and coordination ensures the capability to cover the fusion system as a whole. In 2004 the most significant testing activities concerned the ITER primary beryllium-coated first wall. In the field of high-heat-flux components, an important achievement was the qualification of the process for depositing a copper liner on carbon fibre composite (CFC) hollow tiles. This new process, pre-brazed casting (PBC), allows the hot radial pressing (HRP) joining procedure to be used also for CFC-based armour monoblock divertor components. The PBC and HRP processes are candidates for the construction of the ITER divertor. In the materials field an important milestone was the commissioning of a new facility for chemical vapour infiltration/deposition, used for optimising silicon carbide composite (SiCf/SiC) components. Eight patents were deposited during 2004

  1. Fusion material development program in the broader approach activities

    Energy Technology Data Exchange (ETDEWEB)

    Nishitani, T. [Directorates of Fusion Energy Research: Naka, Ibaraki, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Tanigawa, H.; Jitsukawa, S. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Hayashi, K.; Takatsu, H. [Fusion Research and Development Directorate, Japan Momie Energy Agency, Ibaraki-ken (Japan); Yamanishi, T. [Tritium Process Laboratory, Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken (Japan); Tsuchiya, K. [Directorates of Fusion Energy Research, JAEA, Higashi-ibaraki-gun, Ibaraki-ken (Japan); MoIslang, A. [Forschungszentrum Karlsruhe GmbH, FZK, Karlsruhe (Germany); Baluc, N. [EPFL-Ecole Polytechnique Federale de Lausanne, Association Euratom-Confederation Suisse, UHD - CRPP, PPB, Lausanne (Switzerland); Pizzuto, A. [ENEA CR Frascat, Frascati (Italy); Hodgson, E.R. [CIEMAT-Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, Association Euratom-CIEMAT, Madrid (Spain); Lasser, R.; Gasparotto, M. [EFDA CSU Garching (Germany)

    2007-07-01

    Full text of publication follows: The world fusion community is now launching construction of ITER, the first nuclear-grade fusion machine in the world. In parallel to the ITER program, Broader Approach (BA) activities are initiated by EU and Japan, mainly at Rokkasho BA site in Japan. The BA activities include the International Fusion Materials Irradiation Facility-Engineering Validation and Engineering Design Activities (IFMIF-EVEDA), the International Fusion Energy Research Center (IFERC), and the Satellite Tokamak. IFERC consists of three sub project; a DEMO Design and R and D coordination Center, a Computational Simulation Center, and an ITER Remote Experimentation Center. Technical R and Ds mainly on fusion materials will be implemented as a part of the DEMO Design and R and D coordination Center. Based on the common interest of each party toward DEMO, R and Ds on a) reduced activation ferritic martensitic (RAFM) steels as a DEMO blanket structural material, SiCf/SiC composites, advanced tritium breeders and neutron multiplier for DEMO blankets, and Tritium Technology were selected and assessed by European and Japanese experts. In the R and D on the RAFM steels, the fabrication technology, techniques to incorporate the fracture/rupture properties of the irradiated materials, and methods to predict the deformation and fracture behaviors of structures under irradiation will be investigated. For SiCf/SiC composites, standard methods to evaluate high-temperature and life-time properties will be developed. Not only for SiCf/SiC but also related ceramics, physical and chemical properties such as He and H permeability and absorption will be investigated under irradiation. As the advanced tritium breeder R and D, Japan and EU plan to establish the production technique for advanced breeder pebbles of Li{sub 2}TiO{sub 3} and Li{sub 4}SiO{sub 4}, respectively. Also physical, chemical, and mechanical properties will be investigated for produced breeder pebbles. For the

  2. VNS: A volumetric neutron source for fusion nuclear technology testing and development

    International Nuclear Information System (INIS)

    Abdou, M.A.; Peng, Y.K.; Ying, A.Y.

    1994-01-01

    Recent progress in fusion plasma research and the initiation of the Engineering Design Activity for ITER provide incentives to seriously explore technically sound and logically consistent pathways toward development of fusion as a practical and attractive energy source. A critical goal is the successful construction and operation of a fusion power demonstration plant (DEMO). Major world program strategies call for DEMO operation by the year 2025. Such a date is important in order for fusion to play a significant role in the energy supply market in the second half of the twenty-first century. Without such a DEMO goal, it will be very hard to justify major financial commitments in the near term for major projects such as ITER. The major question is whether a DEMO goal by the year 2025 is attainable from a technical standpoint. This has been the central question being addressed in a study, called VENUS. Results to date show that a DEMO by the year 2025 can be realized if three major facilities begin operation in parallel by the year 2005. These facilities are: (1) ITER, (2) VNS, and (3) IFMIF. Results show that VNS is a necessary element toward DEMO in a strategy consistent with present world program plans. The key requirements to test and develop fusion nuclear components (e.g. blanket) are 1 MW/m 2 neutron wall load, >10 m 2 of test area at the first wall, steady state or long burn plasma operation, fluence of ∼6MWy/m 2 at the first wall in ∼10-12 year period, and duty cycle x availability factor of ∼0.3. Results of the study show that an attractive design envelope for VNS that satisfies the nuclear testing and development requirements exists. Within this design envelope, the most attractive design points for VNS appear to be driven plasma (Q∼1) in tokamak configuration with normal toroidal-field copper coils, major radius 1.5-2.0m, fusion power ∼100MW, and neutron wall load ∼1.5MW/m 2

  3. Torture and its sequel--a comparison between victims from six countries.

    Science.gov (United States)

    Moisander, Pia A; Edston, Erik

    2003-11-26

    from Bangladesh and Turkey. Suspension was common in all countries except for Uganda. Falaka, i.e. beating of the soles, and electric torture were common (>60%) in Bangladesh, Iran, Syria, and Turkey. Sharp injuries inflicted with knives and bayonets were often seen among the Bangladeshi and Ugandans. Burning injuries due to cigarettes were commonly seen only in patients from Bangladesh. Some methods were found to be almost exclusive for each country: "water treatment" (Bangladesh), the "tyre" (Syria), "telephono" and "submarino" (Peru). The sequel of torture differed in some respects between groups. Fractures were more common among Iranians. Patients from Uganda and Bangladesh had numerous scars. Subjectively reported symptoms were most frequent among Bangladeshi, especially joint pain and ear, nose, and throat symptoms and least frequent among Ugandans. PTSD diagnosed on the basis of a psychiatric interview and psychological tests was found in 69-92% of patients in all groups. The study shows significant differences between countries regarding circumstances, torture methods, and sequel to torture. This knowledge is of value to forensic specialists documenting alleged torture and essential for fair and valid forensic statements.

  4. Obituary: Sidney Edelson, 1916-2002

    Science.gov (United States)

    Yeomans, Donald Keith

    2003-12-01

    On 24 March 2002, the solar physicist Sidney Edelson died in Santa Barbara, California. Sidney was born in Brooklyn NY on 24 August 1916 to Benjamin and Sarah Edelson. His father worked in the garment industry. He obtained his BA from Brooklyn College (1938) and a MA from New York University (1949). He entered Georgetown University in 1950 and received both a MA (1953) and PhD (1961). His PhD thesis was entitled ``A Study of Long and Short Term Variations in Solar Radiation at Radio and Optical Wavelengths." When the U.S. entered World War II, Sidney enlisted as an apprentice seaman and retired as a Lt Commander. He was active in both the European and Pacific theaters. He was captain of a minesweeper in the northern Atlantic and commanded a LST vehicle landing troops at Normandy on Omaha beach. Later on, he was part of the amphibious forces that landed the 5th Marine Division at Okinawa. After the war, he commanded the USS Typhon repatriating Japanese POWs from China to Sasebo, Japan. For this, he was given a certificate of appreciation by Chiang Kaishek. After the war efforts, he served in 1946-47 as a port captain for the China Waterways Transport in Shanghai. It was at this time that he met and married Erny Margaret Anderson, a surgical nurse. They were unable to have children because of a bayonet wound suffered by Erny during a 1937 Japanese attack upon the Catholic Mission hospital where she was working. The majority of his research work was in the area of solar physics; he noted the time relationships between centimeter wavelength bursts and Halpha solar flares (1959), studied the short term variations in the solar radiation at radio and optical wavelengths (1961), observed the solar flux variations at mm and cm radio wavelengths (1973), and pointed out the close association of the emission features at 9 and 3.3 mm with the solar magnetic field structure (1973). Sidney's professional career included research work on solar physics at a number of institutions

  5. Neutron Diffraction Residual Strain Tensor Measurements Within The Phase IA Weld Mock-up Plate P-5

    Energy Technology Data Exchange (ETDEWEB)

    Hubbard, Camden R [ORNL

    2011-09-01

    Oak Ridge National Laboratory (ORNL) has worked with NRC and EPRI to apply neutron and X-ray diffraction methods to characterize the residual stresses in a number of dissimilar metal weld mockups and samples. The design of the Phase IA specimens aimed to enable stress measurements by several methods and computational modeling of the weld residual stresses. The partial groove in the 304L stainless steel plate was filled with weld beads of Alloy 82. A summary of the weld conditions for each plate is provided in Table 1. The plates were constrained along the long edges during and after welding by bolts with spring-loaded washers attached to the 1-inch thick Al backing plate. The purpose was to avoid stress relief due to bending of the welded stainless steel plate. The neutron diffraction method was one of the methods selected by EPRI for non-destructive through thickness strain and stress measurement. Four different plates (P-3 to P-6) were studied by neutron diffraction strain mapping, representing four different welding conditions. Through thickness neutron diffraction strain mappings at NRSF2 for the four plates and associated strain-free d-zero specimens involved measurement along seven lines across the weld and at six to seven depths. The mountings of each plate for neutron diffraction measurements were such that the diffraction vector was parallel to each of the three primary orthogonal directions of the plate: two in-plane directions, longitudinal and transverse, and the direction normal to the plate (shown in left figure within Table 1). From the three orthogonal strains for each location, the residual stresses along the three plate directions were calculated. The principal axes of the strain and stress tensors, however, need not necessarily align with the plate coordinate system. To explore this, plate P-5 was selected for examination of the possibility that the principal axes of strain are not along the sample coordinate system axes. If adequate data could

  6. DEMO maintenance scenarios: scheme for time estimations and preliminary estimates for blankets arranged in multi-module-segments

    International Nuclear Information System (INIS)

    Nagy, D.

    2007-01-01

    Previous conceptual studies made clear that the ITER blanket concept and segmentation is not suitable for the environment of a potential fusion power plant (DEMO). One promising concept to be used instead is the so-called Multi-Module-Segment (MMS) concept. Each MMS consists of a number of blankets arranged on a strong back plate thus forming ''banana'' shaped in-board (IB) and out-board (OB) segments. With respect to port size, weight, or other limiting aspects the IB and OB MMS are segmented in toroidal direction. The number of segments to be replaced would be below 100. For this segmentation concept a new maintenance scenario had to be worked out. The aim of this paper is to present a promising MMS maintenance scenario, a flexible scheme for time estimations under varying boundary conditions and preliminary time estimates. According to the proposed scenario two upper, vertical arranged maintenance ports have to be opened for blanket maintenance on opposite sides of the tokamak. Both ports are central to a 180 degree sector and the MMS are removed and inserted through both ports. In-vessel machines are operating to transport the elements in toroidal direction and also to insert and attach the MMS to the shield. Outside the vessel the elements have to be transported between the tokamak and the hot cell to be refurbished. Calculating the maintenance time for such a scenario is rather challenging due to the numerous parallel processes involved. For this reason a flexible, multi-level calculation scheme has been developed in which the operations are organized into three levels: At the lowest level the basic maintenance steps are determined. These are organized into maintenance sequences that take into account parallelisms in the system. Several maintenance sequences constitute the maintenance phases which correspond to a certain logistics scenario. By adding the required times of the maintenance phases the total maintenance time is obtained. The paper presents

  7. Line x-ray source for diffraction enhanced imaging in clinical and industrial applications

    Science.gov (United States)

    Wang, Xiaoqin

    Mammography is one type of imaging modalities that uses a low-dose x-ray or other radiation sources for examination of breasts. It plays a central role in early detection of breast cancers. The material similarity of tumor-cell and health cell, breast implants surgery and other factors, make the breast cancers hard to visualize and detect. Diffraction enhanced imaging (DEI), first proposed and investigated by D. Chapman is a new x-ray radiographic imaging modality using monochromatic x-rays from a synchrotron source, which produced images of thick absorbing objects that are almost completely free of scatter. It shows dramatically improved contrast over standard imaging when applied to the same phantom. The contrast is based not only on attenuation but also on the refraction and diffraction properties of the sample. This imaging method may improve image quality of mammography, other medical applications, industrial radiography for non-destructive testing and x-ray computed tomography. However, the size, and cost, of a synchrotron source limits the application of the new modality to be applicable at clinical levels. This research investigates the feasibility of a designed line x-ray source to produce intensity compatible to synchrotron sources. It is composed of a 2-cm in length tungsten filament, installed on a carbon steel filament cup (backing plate), as the cathode and a stationary oxygen-free copper anode with molybdenum coating on the front surface serves as the target. Characteristic properties of the line x-ray source were computationally studied and the prototype was experimentally investigated. SIMIION code was used to computationally study the electron trajectories emanating from the filament towards the molybdenum target. A Faraday cup on the prototype device, proof-of-principle, was used to measure the distribution of electrons on the target, which compares favorably to computational results. The intensities of characteristic x-ray for molybdenum

  8. Sample holder for studying temperature dependent particle guiding

    International Nuclear Information System (INIS)

    Bereczky, R.J.; Toekesi, K.; Kowarik, G.; Aumayr, F.

    2011-01-01

    Complete text of publication follows. The so called guiding effect is a complex process involving the interplay of a large number of charged particles with a solid. Although many research groups joined this field and carried out various experiments with insulator capillaries many details of the interactions are still unknown. We investigated the temperature dependence of the guiding since it opens new possibilities both for a fundamental understanding of the guiding phenomenon and for applications. For the temperature dependent guiding experiments a completely new heatable sample holder was designed. We developed and built such a heatable sample holder to make accurate and reproducible studies of the temperature dependence of the ion guiding effect possible. The target holder (for an exploded view see Fig. 1) consists of two main parts, the front and the back plates. The two plates of the sample holder, which function as an oven, are made of copper. These parts surround the capillary in order to guarantee a uniform temperature along the whole tube. The temperature of the copper parts is monitored by a K-Type thermocouple. Stainless steel coaxial heaters surrounding the oven are used for heating. The heating power up to a few watts is regulated by a PID controller. Cooling of the capillary is achieved by a copper feed-through connected to a liquid nitrogen bath outside the UHV chamber. This solution allows us to change the temperature of the sample from -30 deg C up to 90 deg C. Our experiments with this newly developed temperature regulated capillary holder show that the glass temperature (i.e. conductivity) can be used to control the guiding properties of the glass capillary and adjust the conditions from guiding at room temperature to simple geometrical transmission at elevated temperatures. This holds the promise to investigate the effect of conductivity on particle transport (build-up and removal of charge patches) through capillaries in more details

  9. NASA Tech Briefs, April 2013

    Science.gov (United States)

    2013-01-01

    Topics covered include: Fully Integrated, Miniature, High-Frequency Flow Probe Utilizing MEMS Leadless SOI Technology; Nanoscale Surface Plasmonics Sensor With Nanofluidic Control; Advanced Dispersed Fringe Sensing Algorithm for Coarse Phasing Segmented Mirror Telescopes; Neural Network Back-Propagation Algorithm for Sensing Hypergols; Bulk Moisture and Salinity Sensor; Change-Based Satellite Monitoring Using Broad Coverage and Targetable Sensing; Circularly Polarized Microwave Antenna Element with Very Low Off-Axis Cross-Polarization; Ultra-Low Heat-Leak, High-Temperature Superconducting Current Leads for Space Applications; Flash Cracking Reactor for Waste Plastic Processing; An Automated Safe-to-Mate (ASTM) Tester; Wireless Chalcogenide Nanoionic-Based Radio-Frequency Switch; Compute Element and Interface Box for the Hazard Detection System; DOT Transmit Module; Composite Aerogel Multifoil Protective Shielding; Li-Ion Electrolytes with Improved Safety and Tolerance to High-Voltage Systems; Polymer-Reinforced, Non-Brittle, Lightweight Cryogenic Insulation; Controlled, Site-Specific Functionalization of Carbon Nanotubes with Diazonium Salts; Regenerable Sorbent for CO2 Removal; Sprayable Aerogel Bead Compositions With High Shear Flow Resistance and High Thermal Insulation Value; Lexan Linear Shaped Charge Holder with Magnets and Backing Plate; Robotic Ankle for Omnidirectional Rock Anchors; Wind, Wave, and Tidal Energy Without Power Conditioning; An Active Heater Control Concept to Meet IXO Type Mirror Module Thermal-Structural Distortion Requirement; Waterless Clothes-Cleaning Machine; Integrated Electrical Wire Insulation Repair System; LVGEMS Time-of-Flight Mass Spectrometry on Satellites; Surface Inspection Tool for Optical Detection of Surface Defects; Per-Pixel, Dual-Counter Scheme for Optical Communications; Certification-Based Process Analysis; Surface Navigation Using Optimized Waypoints and Particle Swarm Optimization; Smart-Divert Powered Descent

  10. Initial results in SST-1 after up-gradation

    Science.gov (United States)

    Pradhan, S.; Khan, Z.; Tanna, V. L.; Prasad, U.; Paravastu, Y.; Raval, D. C.; Masand, H.; Kumar, Aveg; Dhongde, J. R.; Jana, S.; Kakati, B.; Patel, K. B.; Bhandarkar, M. K.; Shukla, B. K.; Ghosh, D.; Patel, H. S.; Parekh, T. J.; Mansuri, I. A.; Dhanani, K. R.; Varadharajulu, A.; Khristi, Y. S.; Biswas, P.; Gupta, C. N.; George, S.; Semwal, P.; Sharma, D. K.; Gulati, H. K.; Mahajan, K.; Praghi, B. R.; Banaudha, M.; Makwana, A. R.; Chudasma, H. H.; Kumar, M.; Manchanda, R.; Joisa, Y. S.; Asudani, K.; Pandya, S. N.; Pathak, S. K.; Banerjee, S.; Patel, P. J.; Santra, P.; Pathan, F. S.; Chauhan, P. K.; Khan, M. S.; Thankey, P. L.; Prakash, A.; Panchal, P. N.; Panchal, R. N.; Patel, R. J.; Mahsuria, G. I.; Sonara, D. P.; Patel, K. M.; Jayaswal, S. P.; Sharma, M.; Patel, J. C.; Varmora, P.; Srikanth, G. L. N.; Christian, D. R.; Garg, A.; Bairagi, N.; Babu, G. R.; Panchal, A. G.; Vora, M. M.; Singh, A. K.; Sharma, R.; Nimavat, H. D.; Shah, P. R.; Purwar, G.; Raval, T. Y.; Sharma, A. L.; Ojha, A.; Kumar, S.; Ramaiya, N. K.; Siju, V.; Gopalakrishna, M. V.; Kumar, A.; Sharma, P. K.; Atrey, P. K.; Kulkarni, SV; Ambulkar, K. K.; Parmar, P. R.; Thakur, A. L.; Raval, J. V.; Purohit, S.; Mishra, P. K.; Adhiya, A. N.; Nagora, U. C.; Thomas, J.; Chaudhari, V. K.; Patel, K. G.; Dalakoti, S.; Virani, C. G.; Gupta, S.; Kumar, Ajay; Chaudhari, B.; Kaur, R.; Srinivasan, R.; Raju, D.; Kanabar, D. H.; Jha, R.; Das, A.; Bora, D.

    2017-04-01

    SST-1 Tokamak has recently completed the 1st phase of up-gradation with successful installation and integration of all its First Wall components. The First Wall of SST-1 comprises of ∼ 3800 high heat flux compatible graphite tiles being assembled and installed on 132 CuCrZr heat sink back plates engraved with ∼ 4 km of leak tight baking and cooling channels in five major sub groups equipped with ∼ 400 sensors and weighing ∼ 6000 kg in total in thirteen isolated galvanic and six isolated hydraulic circuits. The phase-1 up-gradation spectrum also includes addition of Supersonic Molecular Beam Injection (SMBI) both on the in-board and out-board side, installation of fast reciprocating probes, adding some edge plasma probe diagnostics in the SOL region, installation and integration of segmented and up-down symmetric radial coils aiding/controlling plasma rotations, introduction of plasma position feedback and density controls etc. Post phase-I up-gradation spanning from Nov 2014 till June 2016, initial plasma experiments in up-graded SST-1 have begun since Aug 2016 after a brief engineering validation period in SST-1. The first experiments in SST-1 have revealed interesting aspects on the ‘eddy currents in the First Wall support structures’ influencing the ‘magnetic Null evolution dynamics’ and the subsequent plasma start-up characteristics after the ECH pre-ionization, the influence of the first walls on the ‘field errors’ and the resulting locked modes observed, the magnetic index influencing the evolution of the equilibrium of the plasma column, low density supra-thermal electron induced discharges and normal ohmic discharges etc. Presently; repeatable ohmic discharges regimes in SST-1 having plasma currents in excess of 65 KA (qa ∼ 3.8, BT = 1.5 T) with a current ramp rates ∼ 1.2 MA/s over a duration of ∼ 300 ms with line averaged densities ∼ 0.8 × 1019 and temperatures ∼ 200 eV with copious MHD signatures have been experimentally

  11. Process Model for Friction Stir Welding

    Science.gov (United States)

    Adams, Glynn

    1996-01-01

    Friction stir welding (FSW) is a relatively new process being applied for joining of metal alloys. The process was initially developed by The Welding Institute (TWI) in Cambridge, UK. The FSW process is being investigated at NASA/MSEC as a repair/initial weld procedure for fabrication of the super-light-weight aluminum-lithium shuttle external tank. The FSW investigations at MSFC were conducted on a horizontal mill to produce butt welds of flat plate material. The weldment plates are butted together and fixed to a backing plate on the mill bed. A pin tool is placed into the tool holder of the mill spindle and rotated at approximately 400 rpm. The pin tool is then plunged into the plates such that the center of the probe lies at, one end of the line of contact, between the plates and the shoulder of the pin tool penetrates the top surface of the weldment. The weld is produced by traversing the tool along the line of contact between the plates. A lead angle allows the leading edge of the shoulder to remain above the top surface of the plate. The work presented here is the first attempt at modeling a complex phenomenon. The mechanical aspects of conducting the weld process are easily defined and the process itself is controlled by relatively few input parameters. However, in the region of the weld, plasticizing and forging of the parent material occurs. These are difficult processes to model. The model presented here addresses only variations in the radial dimension outward from the pin tool axis. Examinations of the grain structure of the weld reveal that a considerable amount of material deformation also occurs in the direction parallel to the pin tool axis of rotation, through the material thickness. In addition, measurements of the axial load on the pin tool demonstrate that the forging affect of the pin tool shoulder is an important process phenomenon. Therefore, the model needs to be expanded to account for the deformations through the material thickness and the

  12. Mechanical and thermal resistance of multi-material components for ITER

    International Nuclear Information System (INIS)

    Burlet, H.

    2013-01-01

    The First Wall panels for ITER are complex parts composed of stainless steel, copper and beryllium [1]. These materials are joined using diffusion bonding technique. The stainless steel is a commonly used in nuclear reactors 316LN material and acts as a structural material. The copper alloy is a CuCrZr material which acts as a heat sink. The beryllium consisting in tiles and layer is used as the protective plasma facing material. The fabrication of these panels is performed through 2 main steps. The first step consists in welding all together a bi-metallic support structure made from a thick CuCrZr plate embedded with 316LN tubes and bonded to a thick 316LN backing plate with cooling channels. The bonding is performed in a HIP (Hot Isostatic Pressure) facility. The second step is performed at a lower temperature and aims at simultaneously welding by HIP Be onto CuCrZr and ageing the CuCrZr heat sink to obtain the correct mechanical resistance of this alloy reinforced by precipitates. The various joints 316LN/316LN, 316LN/CuCrZr, and CuCrZr/Be are then characterized [2] from a microstructural point of view and by mechanical tests. It is quite hard to characterize the strength of a diffusion bonded joints. Standard tests may be used for homogeneous joints whereas specific tests have been developed to characterize the heterogeneous bonds. To optimize the bond, we performed mainly impact and tensile bi-material tests (Fig 1). Once the manufacture parameters have been optimized, advanced mechanical tests are performed based on Bimetallic CT specimens, axisymmetric notched specimens, 4P bending specimens. Numerical simulations are required to analyse the mechanical response. In order to characterize the fatigue resistance of the joints, specific mock-ups have been designed by the European Fusion Development Agreement EFDA team (Fig 2). Results of heat flux testing will be reviewed for the various joints. The assembly of heterogeneous materials by Hipping is very complex

  13. Annual report of Fusion Research and Development Directorate of JAEA for FY2008 and FY2009

    International Nuclear Information System (INIS)

    Isei, Nobuaki

    2011-03-01

    Research Center, preliminary technological development on R and D issues related to each of low-activation structural materials, SiC/SiC composite materials, tritium technologies, advanced tritium breeder, and advanced neutron multiplier to create DEMO reactors was conducted. In addition, review of conceptual design of DEMO reactors through cooperation with universities and domestic research institutes, and that of selection of the models of super computers to be installed in the Fusion Computer Simulation Center were promoted. In engineering demonstration/design activities for the International Fusion Materials Irradiation Facility (IFMIF), the development of lithium test loop was promoted under cooperation with JAEA's Oarai Research and Development Center which has technologies for liquid metals, and the local construction was started. Besides, design of accelerators was promoted and fabrication of prototypes was started. Concerning activities related to Satellite Tokamak (JT-60SA), integrated design of JT-60SA was completed in Japan and Europe incorporating domestic opinions, and fabrication of superconducting conductors for poloidal field coils was started and procurement activities were promoted in the facilities for fabricating superconducting conductors constructed in Naka Fusion Institute. At the same time, the operation of JT-60 was completed in August 2008 aiming for the establishment of JT-60SA, and preparations for dismantling toward full-scale dismantling and removal are being promoted according to schedule. In addition, the development of the Rokkasho BA site, which will be the center of the BA activities, was also advanced according to schedule, and construction of the DEMO R and D Building, the Computer Simulation and Remote experiment Building, and the IFMIF/EVEDA Accelerator Building, and the central substation was completed as originally planned in March 2010. For research and development on fusion plasma, research on the realization of steady-state and

  14. Message from the Editor

    Science.gov (United States)

    Stambaugh, Ronald D.

    2014-01-01

    , General Atomics, USA A. Hassanein, Purdue University, USA Y.-M. Jeon, National Fusion Research Institute, Spain S. Kajita, Nagoya University, Japan T.P. Kiviniemi, Aalto University, Finland R.M. More, Lawrence Livermore National Laboratory, USA F. Sattin, Associazione Euratom-ENEA-CNR, Italy J.A. Snipes, ITER Organization, France W. Suttrop, Max Planck Institute for Plasma Physics-Garching, Germany F.L. Tabares, Energy Environment and Technology Research Centre, Spain Y. Ueda, Osaka University, Japan V.S. Voitsenya, Kharkov Institute of Physics and Technology, Ukraine G. Xu, Chinese Academy of Sciences-Hefei Institutes of Physical Sciences, People's Republic of China In addition, there is a group of several hundred referees who have helped us in the past year to maintain the high scientific standard of Nuclear Fusion . At the end of this issue we give the full list of all referees for 2013. Our thanks to them! We also wish to express our thanks to Paul Thomas, who served as Guest Editor for the special issue of the overview and summary reports from the 24th Fusion Energy Conference in San Diego, October 2012. This issue is of great value as a summary of the major developments worldwide in fusion research in the last two years. Authors The winner of the 2013 Nuclear Fusion Award is D.G. Whyte for the paper: I-mode: an H-mode energy confinement regime with L-mode particle transport in Alcator C-Mod [1], and we congratulate him and coauthors on this achievement. We also note special topic papers published in 2013: Technical challenges in the construction of the steady-state stellarator Wendestein 7-X by H.S. Bosch et al [2], Power requirements for electron cyclotron current drive and ion cyclotron resonance heating for sawtooth control in ITER by I.T. Chapman et al [3] and IFMIF: overview of the validation activities by J. Knaster et al [4]. The Board of Editors The Board of Editors has had a substantial turnover in members. For their great service to the journal, we wish to

  15. Annual report of Fusion Research and Development Directorate of JAEA

    International Nuclear Information System (INIS)

    Kubo, Hirotaka; Hoshino, Katsumichi; Isei, Nobuaki; Nakamura, Hiroo; Sato, Satoshi; Shimada, Katsuhiro; Sugie, Tatsuo

    2009-01-01

    the ITER TBM fabrication technology, a full-scale TBM first wall was fabricated with reduced-activation-ferritic-martensitic steel (F82H) by a hot isostatic press method. Tritium processing technology for breeding blankets and neutronics integral experiments with a blanket mockup were also progressed. For ITER and DEMO blankets, studies on neutron irradiation effects and ion irradiation effects on F82H steel characteristics were continued using HFIR, TIARA and so on. ITER Agreements entered into force on 24th October 2007. On the same day ITER Organization (IO) was established and JAEA was designated as the Japanese Domestic Agency (JADA) by Japanese Government. The Procurement Arrangement for the Toroidal Field (TF) Conductors was concluded between the IO and JADA in Nov. 2007, and then the contract with Japanese companies to fabricate TF conductors was launched in March 2008. The Quality Assurance Program of JADA inevitable to implement the procurement was approved by the IO. The Project Management (e.g. Schedule Management, Procurement Management, and QA) of JADA was executed. The preparation of the procurement was continued for the TF coil, Blanket First Wall, Divertor, Remote maintenance System of Blanket, EC and NB Heating System, and Diagnostics. The Agreement for the Broader Approach (BA) Activities entered into force and JAEA was assigned as the Implementing Agency for the BA Activities, on 1st June 2007. Contracts for the constructions of the buildings for the IFERC Project and the IFMIF/EVEDA Project as well as the preparation for the Rokkasho site were made in March 2008. For the IFERC Project, information exchange common view of DEMO design was carried out. Procurement Arrangement for the urgent tasks to be implemented for DEMO R and D was concluded. Preparation toward selection of a super computer for the Computational Simulation Centre started. For the IFMIF/EVEDA Project, detailed planning for the initial engineering research of the accelerator

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

    International Nuclear Information System (INIS)

    Diegele, E.

    2009-01-01

    environment and conditions. Materials and materials technologies (fabrication, welding, joining) have to be fully qualified in front of a rigorous licensing process within the next decade. Therefore, materials development for DEMO is based on present technologies and knowledge with some reasonable extrapolation. The 9% Cr steel EUROFER steel is the primary EU candidate structural material. For increased thermal efficiency the temperature window of the structural materials needs to be enlarged. Various ODS (Oxide Dispersion Strengthened) Fe-Cr-steels are candidates for higher temperature application. The large fraction of high energy neutrons in the fusion neutron spectrum results in gaseous transmutations (He and H) that are more than one order of magnitude higher than in fission. Even though fission based material test reactors are the essential and indispensable pillar of the current and future irradiation qualification programme, they can not provide sufficient data for a successful licensing process towards DEMO. For this reason, the construction and use of a facility called IFMIF, designed for simulating as closely a possible the fusion neutron spectrum, is mandatory. Meantime, and complimentary, an enhanced material science programme should increase knowledge and understanding of radiation effects. The focus of this programme for the next decade should be on the development and validation of predictive capabilities for modelling micro-structural evolution and mechanical properties of EUROFER-type steels under fusion reactor relevant conditions, addressing in particular the Helium issue. In a longer term perspective, this should result in the implementation of an integrated approach involving modelling and model-oriented experimental validation into a strategy of accelerated development and testing of candidate fusion materials, material systems and material technologies. (author)

  17. Radioprotection optimization in the electronuclear, industrial and medical domains

    International Nuclear Information System (INIS)

    Schieber, C.; Abela, G.; Ammerich, M.; Balduyck, S.; Batalla, A.; Drouet, F.; Fracas, P.; Gauron, Ch.; Le Guen, B.; Lombard, J.; Mougnard, Ph.; Murith, Ch.; Rannou, A.; Rodde, S.; Selva, M.; Tranchant, Ph.; Schieber, C.; Solaire, T.; Le Tonqueze, Y.; Jolivet, P.; Chauveau, D.; Mathevet, L.; Juhel, T.; Mertz, L.; Bochud, F.O.; Desmaris, G.; Turquet de Beauregard, G.; Roy, C.; Delacroix, S.; Sevilla, A.; Rehel, J.L.; Bernhard, S.; Palut-Laurent, O.; Lochard, J.; Lebaron-Jacobs, L.; Wack, G.; Barange, K.; Delabre, H.

    2011-01-01

    This document gathers the slides of the available presentations given during these conference days. Thirty one presentations are assembled in the document and deal with: 1 - implementation of the ALARA principle in the nuclear, industrial and medical domains: status and challenges (C. Schieber); 2 - image quality and scanner irradiation: what ingredients to chose? (T. Solaire); 3 - radioprotection stakes and implementation of the ALARA approach during the IFMIF design (Y. Le Tonqueze); 4 - ALARA at the design stage of the EPR (P. Jolivet); 5 - alternative techniques to iridium 192 gamma-graphy for welds control: results and recommendations from the ALTER-X project (D. Chauveau); 6 - alternative techniques to ionizing radiations use in the medical domain: implementation of navigation strategies (L. Mathevet); 7 - justification of ionizing radiations use in non-medical imaging: overview of the French situation and perspectives status (S. Rodde); 8 - ISOE: task scheduling for radioprotection optimization in nuclear power plants (G. Abela); 9 - Practices and ALARA prospects among big nuclear operators (T. Juhel); 10 - experience feedback on the use of diagnostic reference levels (DRLs) in diagnostic imaging optimization (L. Mertz); 11 - DRLs: Swiss strategy and concept limits (F.O. Bochud); 12 - external dosimetry tools: the existing, the developing and the remaining problems (A. Rannou); 13 - is the optimization principle applicable to the aircraft personnel's exposure to cosmic radiation? (G. Desmaris); 14-15 - experience feedback of the ALARA approach concerning an operation with strong dosimetric stakes (P. Mougnard and N. Fontaine); 16 - optimization of reactor pool decontaminations ((P. Tranchant); 17 - radiopharmaceuticals transport - ALARA principle related stakes (G. Turquet de Beauregard); 18 - ALARA in vet radio-diagnosis activity: good practices guide (C. Roy); 19 - implementation of the ALARA approach at the Proton-therapy centre of Orsay's Curie Institute

  18. Radioprotection optimization in the electronuclear, industrial and medical domains; Optimisation de la radioprotection dans les domaines electronucleaire, industriel et medical

    Energy Technology Data Exchange (ETDEWEB)

    Schieber, C.; Abela, G.; Ammerich, M.; Balduyck, S.; Batalla, A.; Drouet, F.; Fracas, P.; Gauron, Ch.; Le Guen, B.; Lombard, J.; Mougnard, Ph.; Murith, Ch.; Rannou, A.; Rodde, S.; Selva, M.; Tranchant, Ph.; Schieber, C.; Solaire, T.; Le Tonqueze, Y.; Jolivet, P.; Chauveau, D.; Mathevet, L.; Juhel, T.; Mertz, L.; Bochud, F.O.; Desmaris, G.; Turquet de Beauregard, G.; Roy, C.; Delacroix, S.; Sevilla, A.; Rehel, J.L.; Bernhard, S.; Palut-Laurent, O.; Lochard, J.; Lebaron-Jacobs, L.; Wack, G.; Barange, K.; Delabre, H.

    2011-07-01

    This document gathers the slides of the available presentations given during these conference days. Thirty one presentations are assembled in the document and deal with: 1 - implementation of the ALARA principle in the nuclear, industrial and medical domains: status and challenges (C. Schieber); 2 - image quality and scanner irradiation: what ingredients to chose? (T. Solaire); 3 - radioprotection stakes and implementation of the ALARA approach during the IFMIF design (Y. Le Tonqueze); 4 - ALARA at the design stage of the EPR (P. Jolivet); 5 - alternative techniques to iridium 192 gamma-graphy for welds control: results and recommendations from the ALTER-X project (D. Chauveau); 6 - alternative techniques to ionizing radiations use in the medical domain: implementation of navigation strategies (L. Mathevet); 7 - justification of ionizing radiations use in non-medical imaging: overview of the French situation and perspectives status (S. Rodde); 8 - ISOE: task scheduling for radioprotection optimization in nuclear power plants (G. Abela); 9 - Practices and ALARA prospects among big nuclear operators (T. Juhel); 10 - experience feedback on the use of diagnostic reference levels (DRLs) in diagnostic imaging optimization (L. Mertz); 11 - DRLs: Swiss strategy and concept limits (F.O. Bochud); 12 - external dosimetry tools: the existing, the developing and the remaining problems (A. Rannou); 13 - is the optimization principle applicable to the aircraft personnel's exposure to cosmic radiation? (G. Desmaris); 14-15 - experience feedback of the ALARA approach concerning an operation with strong dosimetric stakes (P. Mougnard and N. Fontaine); 16 - optimization of reactor pool decontaminations ((P. Tranchant); 17 - radiopharmaceuticals transport - ALARA principle related stakes (G. Turquet de Beauregard); 18 - ALARA in vet radio-diagnosis activity: good practices guide (C. Roy); 19 - implementation of the ALARA approach at the Proton-therapy centre of Orsay's Curie

  19. Application of Galerkin meshfree methods to nonlinear thermo-mechanical simulation of solids under extremely high pulsed loading

    International Nuclear Information System (INIS)

    Ibáñez, Daniel Iglesias; García Orden, Juan C.; Brañas, B.; Carmona, J.M.; Molla, J.

    2013-01-01

    Highlights: • The paper presents a novel application of meshfree methods, valid for its implementation on a multibody framework. • Coupled nonlinear thermo-mechanical formulation is detailed and described in the reference configuration, as this allows to compute the shape functions only once. • We show the conditions in which future information induces inefficiency. • Beam parameters are the only information needed to apply the thermal load. • The solution procedure takes charge of updating the volumetric heat rate as the body moves and deforms. -- Abstract: Beam facing elements of the International Fusion Materials Irradiation Facility (IFMIF) Linear Particle Accelerator prototype (LIPAc) must stop 5–40 MeV D + ions with a peak current of 125 mA. The duty cycle of the beam loading varies from 0.1% to 100% (CW), depending on the device, with the ions being stopped in the first hundreds microns of the beam facing material. For intermediate duty cycles up to CW, the thermal load can be considered a heat flux load on the boundary, but this approximation gets too conservative as the duty cycle is reduced because the thermal diffusion becomes more important. Instant heat flux produced by the beam can reach up to 3 GW/m 2 in elements such as the beam dump and slits during short times of hundredths of microseconds. In these cases, the accuracy of the volumetric heat generation is critical for obtaining realistic results. Meshfree Galerkin methods discretize a continuum using scattered nodes. As opposed to FEM, no predefined connectivity is needed between the nodes, so C ∞ (infinitely differentiable) locally supported shape functions can be used to approximate both the trial and the test functions. This feature makes these type of methods well suited for those problems where the domain experiences very large deformations or has high gradients of the state variables. Radial basis (RBF) and moving least squares (MLS) functions have been applied to the

  20. Annual report on major results and progress of Naka Fusion Research Establishment of JAERI from April 1 to September 30, 2005 and Fusion Research and Development Directorate of JAEA from October 1, 2005 to March 31, 2006

    International Nuclear Information System (INIS)

    Yoshida, Hidetoshi; Oasa, Kazumi; Hayashi, Takao; Nakamura, Hiroo; Ogawa, Hiroaki

    2006-09-01

    components, critical heat flux of a screw tube has been examined. Neutronics integral experiments with a blanket mockup were also progressed. For ITER TBMs and DEMO blankets, irradiation effects on F82H characteristics were progressed using HFIR, JMTR and so on. In the IFMIF program, transitional activities were also progressed. Vacuum technology and its application to industries have been examined. In the ITER Program, under the framework of the ITER Transitional Arrangements, the Design and R and D Tasks have been carried out by the Participant Teams along the work plan approved on September 2005. In FY 2005, JAERI/JAEA has performed sixty-six Design Tasks and has completed thirty-four Tasks that make the implementation of preparing the procurement documents for facilities and equipments that are scheduled to be ordered at an early stage of ITER construction. The work plan for the Broader Approach' Program has been continuously discussed through the bilateral negotiation meetings between Japan and the EU, and JAERI/JAEA provided the technical support for the meetings. Finally, in fusion reactor design studies, a reactor concept of SlimCS was proposed to demonstrate an electric power generation of 1GW level, self-sufficiency of tritium fuel and year-long continuous operation. (author)

  1. Ferritic steels for the first generation of breeder blankets

    International Nuclear Information System (INIS)

    Diegele, E.

    2009-01-01

    environment and conditions. Materials and materials technologies (fabrication, welding, joining) have to be fully qualified in front of a rigorous licensing process within the next decade. Therefore, materials development for DEMO is based on present technologies and knowledge with some reasonable extrapolation. The 9% Cr steel EUROFER steel is the primary EU candidate structural material. For increased thermal efficiency the temperature window of the structural materials needs to be enlarged. Various ODS (Oxide Dispersion Strengthened) Fe-Cr-steels are candidates for higher temperature application. The large fraction of high energy neutrons in the fusion neutron spectrum results in gaseous transmutations (He and H) that are more than one order of magnitude higher than in fission. Even though fission based material test reactors are the essential and indispensable pillar of the current and future irradiation qualification programme, they can not provide sufficient data for a successful licensing process towards DEMO. For this reason, the construction and use of a facility called IFMIF, designed for simulating as closely a possible the fusion neutron spectrum, is mandatory. Meantime, and complimentary, an enhanced material science programme should increase knowledge and understanding of radiation effects. The focus of this programme for the next decade should be on the development and validation of predictive capabilities for modelling micro-structural evolution and mechanical properties of EUROFER-type steels under fusion reactor relevant conditions, addressing in particular the Helium issue. In a longer term perspective, this should result in the implementation of an integrated approach involving modelling and model-oriented experimental validation into a strategy of accelerated development and testing of candidate fusion materials, material systems and material technologies. (author)

  2. Performance of candidate gas turbine abradeable seal materials in high temperature combustion atmospheres

    Energy Technology Data Exchange (ETDEWEB)

    Simms, N.J. [Cranfield University, Power Generation Technology Centre, Cranfield, Beds, MK43 0AL (United Kingdom); Norton, J.F. [Cranfield University, Power Generation Technology Centre, Cranfield, Beds, MK43 0AL (United Kingdom); Consultant in Corrosion Science and Technology, Hemel Hempstead, Herts HP1 1SR (United Kingdom); McColvin, G. [Siemens Industrial Turbines Ltd., Lincoln, LN5 7FD (United Kingdom)

    2005-11-01

    The development of abradeable gas turbine seals for higher temperature duties has been the target of an EU-funded R and D project, ADSEALS, with the aim of moving towards seals that can withstand surface temperatures as high as {proportional_to} 1100 C for periods of at least 24,000 h. The ADSEALS project has investigated the manufacturing and performance of a number of alternative materials for the traditional honeycomb seal design and novel alternative designs. This paper reports results from two series of exposure tests carried out to evaluate the oxidation performance of the seal structures in combustion gases and under thermal cycling conditions. These investigations formed one part of the evaluation of seal materials that has been carried out within the ADSEALS project. The first series of three tests, carried out for screening purposes, exposed candidate abradeable seal materials to a simulated natural gas combustion environment at temperatures within the range 1050-1150 C in controlled atmosphere furnaces for periods of up to {proportional_to} 2,500 h with fifteen thermal cycles. The samples were thermally cycled to room temperature on a weekly basis to enable the progress of the degradation to be monitored by mass change and visual observation, as well as allowing samples to be exchanged at planned intervals. The honeycombs were manufactured from PM2000 and Haynes 214. The backing plates for the seal constructions were manufactured from Haynes 214. Some seals contained fillers or had been surface treated (e.g. aluminised). The second series of three tests were carried out in a natural gas fired ribbon furnace facility that allowed up to sixty samples of candidate seal structures (including honeycombs, hollow sphere structures and porous ceramics manufactured from an extended range of materials including Aluchrom YHf, PM2Hf, Haynes 230, IN738LC and MarM247) to be exposed simultaneously to a stream of hot combustion gas. In this case the samples were cooled

  3. HyGenSys: a Flexible Process for Hydrogen and Power Production with Reduction of CO2 Emission HyGenSys : un procédé flexible de production d’hydrogène et d’électricité avec réduction des émissions de CO2

    Directory of Open Access Journals (Sweden)

    Giroudière F.

    2010-09-01

    Full Text Available This paper presents the latest development of HyGenSys, a new sustainable process and technology for the conversion of natural gas to hydrogen and power. The concept combines a specific steam reforming reactor-exchanger with a gas turbine. The heat necessary for the steam reforming reaction comes from hot pressurized flue gases produced in a gas turbine instead of a conventional furnace. Thanks to this high level of heat integration, the overall efficiency is improved and the natural gas consumption is reduced which represents an advantage with regard to economics and CO2 emission reduction. In addition to the efficient HyGenSys process scheme itself, the technology of the reactorexchanger also offers a high level of heat integration for even more energy saving. Two main alternatives are examined in order to meet two different requirements. The first one, named HyGenSys-0, focuses on the hydrogen production for the refining and petrochemical application. The second one named HyGenSys-1, concerns the centralized power production with pre-combustion CO2capture. In that case, the produced hydrogen is fully used to fuel a power gas turbine. HyGenSys-1 has been developed and optimised in CACHET, a European Community funded project. The CACHET electrical power objective was 400 MW at the minimum. HyGenSys-0 and HyGenSys-1 are described in detail with challenges and advantages compared to existing technologies. For both alternatives, the heart of the technology is the reactor-exchanger. The reactor-exchanger design relies on an innovative arrangement of bayonet tubes that allows, at large scale, multiple heat exchanges between hot pressurized flue gas, natural gas feed and hydrogen rich stream produced. Cet article présente les développements récents d’HyGenSys, nouvel éco-procédé de conversion du gaz naturel en hydrogène et é