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Sample records for blanket mockup experiment

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

    International Nuclear Information System (INIS)

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

    2006-01-01

    A neutronics experiment has been performed in the frame of European Fusion Technology Program on a mock-up of the EU Test Blanket Module (TBM), Helium Cooled Pebble Bed (HCPB) concept, with the objective to validate the capability of nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. The experiment has been carried out at the FNG 14-MeV neutron source in collaboration between ENEA, Technische Universitaet Dresden, Forschungszentrum Karlsruhe, J. Stefan Institute Ljubljana and with the participation of JAEA. The mock-up, designed in such a way to replicate all relevant nuclear features of the TBM-HCPB, consisted of a steel box containing beryllium block and two intermediate steel cassettes, filled with of Li 2 CO 3 powder, replicating the breeder insert main characteristics: radial thickness, distance between ceramic layers, thickness of ceramic layers and of steel walls. In the experiment, the TPR has been measured using Li 2 CO 3 pellets at various depths at two symmetrical positions at each depth, one in the upper and one in the lower cassette. Twelve pellets were used at each position to determine the TPR profile through the cassette. Three independent measurements were performed by ENEA, TUD/VKTA and JAEA. The neutron flux in the beryllium layer was measured as well using activation foils. The measured tritium production in the TBM (E) was compared with the same quantity (C) calculated by the MCNP.4c using a very detailed model of the experimental set up, and using neutron cross sections from the European Fusion File (EFF ver.3.1) and from the Fusion Evaluated Nuclear Data Library (FENDL ver. 2.1, ITER reference neutron library). C/E ratios were obtained with a total uncertainty on the C/E comparison less than 9% (2 s). A sensitivity and uncertainty analysis has also been performed to evaluate the calculation uncertainty due to the uncertainty on neutron cross sections. The results of such

  2. Re-analysis of HCPB/HCLL Blanket Mock-up Experiments Using Recent Nuclear Data Libraries

    International Nuclear Information System (INIS)

    Kondo, K.; Fischer, U.; Klix, A.; Pereslavtsev, P.; Serikov, A.; Villari, R.

    2014-01-01

    We have re-analysed the two breeding blankets experiments performed previously in the frame of the European fusion program on two mock-ups of the European Helium-Cooled-Lithiium Lead (HCLL) and Helium-Cooled-Pebble-Bed (HCPB) test blanket modules for ITER. The tritium production rate and the neutron and photon spectra measured in these mock-ups were compared with calculations using FENDL-3 Starter Library, release 4 and state-of-the-art nuclear data evaluations, JEFF-3.1.2, JENDL-4.0 and ENDF/B-VII.0. The tritium production calculated for the HCPB mock-up underestimates the experimental result by about 10%. The result calculated with FENDL-3/SLIB4 gives slightly smaller tritium production by 2% than the one with FENDL-2.1. The difference attributes to the slight modification of the total and elastic scattering cross section of Be. For the HCLL experiment, all libraries reproduce the experimental results well. FENDL-3/SLIB4 gives better result both for the measured spectra and the tritium production compared to FENDL-2.1

  3. Conceptual design of a First Wall mock-up experiment in preparation for the qualification of breeding blanket technologies in the Helium Loop Karlsruhe (HELOKA) facility

    Energy Technology Data Exchange (ETDEWEB)

    Zeile, C., E-mail: christian.zeile@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Abou-Sena, A.; Boccaccini, L.V.; Ghidersa, B.E.; Kang, Q.; Kunze, A. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Lamberti, L. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Dipartimento Energia, Politecnico di Torino (Italy); Maione, I.A.; Rey, J.; Weth, A. von der [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-11-01

    Highlights: • Experiment in preparation for the qualification of Breeding Blanket technologies in HELOKA facility is proposed. • Experimental capabilities, instrumentation of the mock-up and experimental program are presented. • Design and manufacturing of the mock-up is described. • Design of modular attachment system to obtain different stress levels and distributions on the mock-up is discussed. - Abstract: An experimental program based on a First Wall mock-up is presented as preparation for the qualification of breeding blanket mock-ups at high heat flux in the Helium Loop Karlsruhe (HELOKA) facility. Two objectives of the experimental program have been defined: testing of the experimental setup and a first validation of FE models. The design and manufacturing of mock-up representing about 1/3 of the heated zone of an ITER Test Blanket Module (TBM) First Wall is discussed. A modular attachment system concept has been developed for the fixation of the mock-up in order to be able to generate different stress distributions and levels on the plate, which is confirmed by thermo-mechanical analyses. The HELOKA facility is able to provide a TBM relevant helium cooling system and to generate the required surface heat flux by an electron beam gun. An installed IR camera can be used to measure the temperature distribution on the surface.

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  5. Experiment and analysis of hypervapotron mock-ups for preparing the 2nd qualification of the ITER blanket first wall

    International Nuclear Information System (INIS)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Bang, In Cheol

    2010-01-01

    According to the increased heat flux condition up to 5 MW/m 2 in the International Thermonuclear Experimental Reactor (ITER), new design of the blanket first wall (FW) has been considered and the analysis was performed with ANSYS-CFX for checking its temperature with the ITER operation conditions. And a semi-prototype of the FW was proposed to be tested with the similar heat flux conditions under the second qualification for the FW procurement. In order to investigate the fabrication procedure and analysis capability of the code, two types of mock-up were fabricated according to the current semi-prototype design except for bending shape; one with hypervapotron and another without it. They were tested with KoHLT-2 (Korea Heat Load Test) facility and the results were compared with the ones by CFX code. The mass flow rate of inlet coolant was the same as the ITER condition and heat flux was loaded up to 0.48 MW/m 2 heat flux. The results show that the temperature of the mock-up can be predicted using the CFX code even with the complex geometry and the hypervapotron shows its function to increase the cooling.

  6. Neutronics experiments for uncertainty assessment of tritium breeding in HCPB and HCLL blanket mock-ups irradiated with 14 MeV neutrons

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Pillon, M.; Villari, R.; Fischer, U.; Klix, A.; Leichtle, D.; Kodeli, I.; Pohorecki, W.

    2012-01-01

    Two neutronics experiments have been carried out at 14 MeV neutron sources on mock-ups of the helium cooled pebble bed (HCBP) and the helium cooled lithium lead (HCLL) variants of ITER test blanket modules (TBMs). These experiments have provided an experimental validation of the calculations of the tritium production rate (TPR) in the two blanket concepts and an assessment of the uncertainties due to the uncertainties on nuclear data. This paper provides a brief summary of the HCPB experiment and then focuses in particular on the final results of the HCLL experiment. The TPR has been measured in the HCLL mock-up irradiated for long times at the Frascati 14 MeV Neutron Generator (FNG). Redundant and well-assessed experimental techniques have been used to measure the TPR by different teams for inter-comparison. Measurements of the neutron and gamma-ray spectra have also been performed. The analysis of the experiment, carried out by the MCNP code with FENDL-2.1 and JEFF-3.1.1 nuclear data libraries, and also including sensitivity/uncertainty analysis, shows good agreement between measurements and calculations, within the total uncertainty of 5.9% at 1σ level. (paper)

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

    International Nuclear Information System (INIS)

    Kodeli, I.

    2007-01-01

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

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

    Directory of Open Access Journals (Sweden)

    I. Kodeli

    2008-01-01

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

  9. Development of ITER shielding blanket prototype mockup by HIP bonding

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi; Furuya, Kazuyuki; Hatano, Toshihisa; Kuroda, Toshimasa; Enoeda, Mikio; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Takatsu, Hideyuki [Japan Atomic Energy Research Inst., Office of ITER Project Promotion, Tokyo (Japan)

    2000-07-01

    A prototype ({approx}900{sup H} x 1700{sup W} x 350{sup T} mm) of the ITER shielding blanket module has been fabricated following the previous successful fabrication of a small-scale ({approx}500{sup H} x 400{sup W} x 150{sup T} mm) and mid-scale ({approx}800{sup H} x 500{sup W} x 350{sup T} mm) mock-ups. This prototype incorporates most of key design features essential to the fabrication of the ITER shielding blanket module such as 1) the first wall heat sink made of Al{sub 2}O{sub 3} dispersion strengthened Cu (DSCu) with built-in SS316L coolant tubes bonded to a massive SS316LN shield block, 2) toroidally curved first wall with a radius of 5106 mm while straight in poloidal direction, 3) coolant channels oriented in poloidal direction in the first wall and in toroidal direction in the shield block, 4) the first wall coolant channel routing to avoid the interference with the front access holes, 5) coolant channels drilled through the forged SS316LN-IG shield block, and 6) four front access holes of 30 mm in diameter penetrated through the first wall and the shield block. For the joining method, especially for the first wall/side wall parts and the shield block, the solid HIP (Hot Isostatic Pressing) process was applied. It is difficult to apply conventional joining methods such as field welding, brazing, explosion bonding and mechanical one-axial diffusion bonding to a wide area bonding because sufficient mechanical strengths can not be obtained and excessive deformations occurs. In order to solve these fabrication issues, HIP bonding was applied. The first wall stainless steel (SS) coolant tubes of 10 mm in inner diameter and l mm in thickness were sandwiched by semi-circular grooved DSCu plates at the first wall and the front region of the side wall, and by semi-circular grooved SS plates at the back region of the side wall. After assembling of these first wall/side wall parts with the shield block, they were simultaneously bonded by single step HIP in order to

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

    Science.gov (United States)

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

    2018-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-01

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

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

    International Nuclear Information System (INIS)

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

    1998-12-01

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

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

    International Nuclear Information System (INIS)

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

    1993-03-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-03-01

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

  17. Neutronic and thermal estimation of blanket in-pile mockup with Li2TiO3 pebbles

    International Nuclear Information System (INIS)

    Nagao, Y.; Nakamichi, M.; Tsuchiya, K.; Kawamura, H.

    2001-01-01

    To evaluate exactly temperature distribution in large volume of tritium breeding materials during the blanket in-pile tests with the JMTR, neutronic and thermal calculations were conducted by using Monte Carlo code 'MCNP' with nuclear cross section library of 'FSXLIBJ3R2' and the transient and steady-state distribution code 'TRUMP'. From the results of preliminary estimation of temperature distribution in the blanket in-pile mockup, the calculated values were 24-28% higher than the measured values. One of the reasons is due to overestimation of calculated thermal neutron flux

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  19. Tritium breeding experiments with lithium titanate in thermal-type mockups

    International Nuclear Information System (INIS)

    Klix, Axel; Takahashi, Akito; Ochiai, Kentaro; Nishitani, Takeo

    2004-01-01

    Lithium titanate, an advanced tritium breeding material, is currently investigated in integral mock-up experiments at FNS. A method was developed which allows to measure low tritium concentrations directly in this material. The local tritium production rate was obtained by small lithium titanate pellet detectors inserted into the breeding layers which are dissolved after irradiation of the assemblies, and the accumulated tritium was counted by liquid scintillation techniques. The measurement method was applied in mock0-up experiments with candidate materials for the future DEMO reactor breeding blanket. Experimental assemblies consisted of sheets of low activation ferritic steel F82H, lithium titanate, and berylium. Tritium production rate profiles were obtained and compared with results from calculations with the Monte Carlo neutron transport code MCNP-4C. In case of the mock-ups with 95% enriched lithium titanate, the C/E ratios were within the error estimate while larger discrepancies were observed in case of 40% enriched lithium titanate. (author)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hernández, F., E-mail: francisco.hernandez@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany); Kolb, M. [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-WPT) (Germany); Ilić, M.; Kunze, A. [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany); Németh, J. [KFKI Research Institute for Particle and Nuclear Physics (Hungary); Weth, A. von der [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany)

    2013-10-15

    Highlights: ► As preparation for the HCPB-TBM Breeder Unit out-of-pile testing campaign, a pre-test experiment (PREMUX) has been prepared and described. ► A new heater system based on a wire heater matrix has been developed for imitating the neutronic volumetric heating and it is compared with the conventional plate heaters. ► The test section is described and preliminary thermal results with the available models are presented and are to be benchmarked with PREMUX. ► The PREMUX integration in the air cooling loop L-STAR/LL in the Karlsruhe Institute for Technology is shown and future steps are discussed. -- Abstract: The complexity of the experimental set-up for testing a full-scaled Breeder Unit (BU) mock-up for the European Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) has motivated to build a pre-test mock-up experiment (PREMUX) consisting of a slice of the BU in the Li{sub 4}SiO{sub 4} region. This pre-test aims at verifying the feasibility of the methods to be used for the subsequent testing of the full-scaled BU mock-up. Key parameters needed for the modeling of the breeder material is also to be determined by the Hot Wire Method (HWM). The modeling tools for the thermo-mechanics of the pebble beds and for the mock-up structure are to be calibrated and validated as well. This paper presents the setting-up of PREMUX in the L-STAR/LL facility at the Karlsruhe Institute of Technology. A key requirement of the experiments is to mimic the neutronic volumetric heating. A new heater concept is discussed and compared to several conventional heater configurations with respect to the estimated temperature distribution in the pebble beds. The design and integration of the thermocouple system in the heater matrix and pebble beds is also described, as well as other key aspects of the mock-up (dimensions, layout, cooling system, purge gas line, boundary conditions and integration in the test facility). The adequacy of these methods for the full-scaled BU

  3. Fabrication of prototype mockups of ITER shielding blanket with separable first wall

    International Nuclear Information System (INIS)

    Kosaku, Yasuo; Kuroda, Toshimasa; Enoeda, Mikio; Hatano, Toshihisa; Sato, Satoshi; Akiba, Masato

    2002-07-01

    Design of shielding blanket for ITER-FEAT applies the first wall which has the separable structure from the shield block for the purpose of radio-active waste reduction in the maintenance work and cost reduction in fabrication process. Also, it is required to have various types of slots in both of the first wall and the shield block, to reduce the eddy current for reduction of electro-magnetic force in disruption events. This report summarizes the demonstrative fabrication of the ITER shielding blanket with separable first wall performed for the shielding blanket fabrication technology development, under the task agreement of G 16 TT 108 FJ (T420-2) in ITER Engineering Design Activity Extension Period. The objectives of the demonstrative fabrication are: to demonstrate the comprehensive fabrication technique in a large scale component (e.g the joining techniques for the beryllium armor/copper alloy and copper alloy/SS, and the slotting method of the FW and shield block); to develop an improved fabrication method for the shielding blanket based on the ITER-FEAT updated design. In this work, the fabrication technique of full scale separable first wall shield blanket was confirmed by fabricating full width Be armored first wall panel, full scale of 1/2 shield block with poloidal cooling channels. As the R and D for updated cooling channel configuration, the fabrication technique of the radial channel shield block was also demonstrated. Concluding to the all R and D results, it was demonstrated successfully that the fabrication technique and optimized conditions in the results obtained under the task agreement of G 16 TT 95 FJ (T420-1) was applicable to the prototype of the separable first wall blanket module. Additionally, basic echo data of ultra-sonic test method (UT) was obtained to show the applicability of UT method for in tube access detection of defect on the Cu alloy/SS tube interface. (author)

  4. Sensitivity and uncertainty analyses of the HCLL mock-up experiment

    International Nuclear Information System (INIS)

    Leichtle, D.; Fischer, U.; Kodeli, I.; Perel, R.L.; Klix, A.; Batistoni, P.; Villari, R.

    2010-01-01

    Within the European Fusion Technology Programme dedicated computational methods, tools and data have been developed and validated for sensitivity and uncertainty analyses of fusion neutronics experiments. The present paper is devoted to this kind of analyses on the recent neutronics experiment on a mock-up of the Helium-Cooled Lithium Lead Test Blanket Module for ITER at the Frascati neutron generator. They comprise both probabilistic and deterministic methodologies for the assessment of uncertainties of nuclear responses due to nuclear data uncertainties and their sensitivities to the involved reaction cross-section data. We have used MCNP and MCSEN codes in the Monte Carlo approach and DORT and SUSD3D in the deterministic approach for transport and sensitivity calculations, respectively. In both cases JEFF-3.1 and FENDL-2.1 libraries for the transport data and mainly ENDF/B-VI.8 and SCALE6.0 libraries for the relevant covariance data have been used. With a few exceptions, the two different methodological approaches were shown to provide consistent results. A total nuclear data related uncertainty in the range of 1-2% (1σ confidence level) was assessed for the tritium production in the HCLL mock-up experiment.

  5. Results of the mock-up experiment on partial LOCA

    International Nuclear Information System (INIS)

    Dreier, J.; Winkler, H.

    1985-01-01

    A mockup experiment has been performed to verify the heat transfer model for a partial loss of coolant accident in the swimming pool reactor SAPHIR. Three coolant channels with the same dimensions as in a SAPHIR fuel element were simulated using four electrically heated plates. For a water level such that the heated plates are partially submerged, plate temperatures remain below 160 deg. C for plate powers of up to 650 W. For water levels low enough to just block the channels, plate temperatures of 400 deg. C are reached for plate powers as low as 60 W. Details of the experiment and further results are discussed. (author)

  6. Results of the mockup experiment on partial LOCA

    International Nuclear Information System (INIS)

    Dreier, J.; Winkler, H.

    1985-01-01

    A mockup experiment has been performed to verify the heat transfer model for a partial loss of coolant accident in the swimming pool reactor SAPHIR. Three coolant channels with the same dimensions as in a SAPHIR fuel element were simulated using four electrically heated plates. For a water level such that the heated plates are partially submerged, plate temperatures remain below 160 0 C for plate powers of up to 650 W. For water levels low enough to just block the channels, plate temperatures of 400 0 C are reached for plate powers as low as 60 W. Details of the experiment and further results are discussed

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-08-15

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

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

    International Nuclear Information System (INIS)

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

    2004-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-08-01

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

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

    International Nuclear Information System (INIS)

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

    2004-08-01

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

  12. Blanket Cooling Plates Mock-ups Manufactured in different Diffusion Weld Setup

    International Nuclear Information System (INIS)

    Von Der Weth, A.; Aktaa, J.

    2007-01-01

    Full text of publication follows: The breeding blanket box is considered as one of the most important components of a future fusion power plant. It will be assembled by so called cooling plates (CP) with a system of internal cooling channels. Such a CP is produced by two symmetric half pieces with half milled-in channels. Both pieces will be joined by a diffusion weld (DW) process. Within recent years a two step DW process for different EUROFER batches has been developed. It has been first applied to small laboratory scaled samples with dimensions of 25 mm x 30 mm x 40 mm. Then the DW process had then been successfully transferred to so called compact mock ups which are small CPs with dimensions of 67 mm x 70 mm x 50 mm. As third step this process has been used to manufacture a CP (465 mm x 205 mm x 50 mm) of a breeder unit in an industrial uniaxial diffusion weld setup. This paper treats the manufacturing sequence of a cooling plate and a first wall mock up in an industrial hot isostatic pressing (HIP) setup. The firstly laboratory specimens scaled diffusion weld process has been adjusted to different cooling channel dimensions and a different DW setup. The weld quality is investigated by tensile and Charpy impact testing. This allows comparison of the weld quality of mock ups welded in different DW setups. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-11-15

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

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

    International Nuclear Information System (INIS)

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

    2004-08-01

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

  15. Neutronics experiments for DEMO blanket at JAERI/FNS

    International Nuclear Information System (INIS)

    Sato, Satoshi; Ochiai, K.; Hori, J.; Verzilov, Y.; Klix, A.; Wada, M.; Terada, Y.; Yamauchi, M.; Morimoto, Y.; Nishitani, T.

    2003-01-01

    In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li 2 TiO 3 blocks with a 6 Li enrichment of 40 and 95%, and beryllium blocks. Sample pellets of Li 2 TiO 3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test speciments simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steels F82H were irradiated as typical fusion materials. The effective cross-sections for incident neutron flux to calculate the radioactive nuclei ( 56 Co, 184 Re, 48 V, 206 Bi, 65 Zn and 51 Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections remarkably increases with coming closer to polyethylene board that was a substitute of water. As a result of the present study, it has become clear that the sequential reaction rates are important factors to accurately evaluate the induced activity in fusion reactors design. (author)

  16. Neutronics experiments for DEMO blanket at JAERI/FNS

    International Nuclear Information System (INIS)

    Sato, S.; Ochiai, K.; Hori, J.; Verzilov, Y.; Klix, A.; Wada, M.; Terada, Y.; Yamauchi, M.; Morimoto, Y.; Nishitani, T.

    2003-01-01

    In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li 2 TiO 3 blocks with a 6 Li enrichment of 40 and 95%, and beryllium blocks. Sample pellets of Li 2 TiO 3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test specimens simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steel F82H were irradiated as typical fusion materials. The effective cross- sections for incident neutron flux to calculate the radioactive nuclei ( 56 Co, 184 Re, 48 V, 206 Bi, 65 Zn and 51 Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections remarkably increases with coming closer to polyethylene board that was a substitute of water. As a result of the present study, it has become clear that the sequential reaction rates are important factors to accurately evaluate the induced activity in fusion reactors design. (author)

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

    International Nuclear Information System (INIS)

    Kodeli, I.

    2006-01-01

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

  18. Electron beam irradiation experiments of monoblock divertor mock-up

    International Nuclear Information System (INIS)

    Satoh, Kazuyoshi; Akiba, Masato; Araki, Masanori; Suzuki, Satoshi; Yokoyama, Kenji; Smid, I.; Cardella, A.; Duwe, R.; Di Pietro, E.

    1993-03-01

    It is one of the key issues for ITER to develop the divertor plate. Electron beam irradiation tests were carried out on a NET divertor mock-up using JEBIS at JAERI under a collaboration between The NET team, JAERI and KFA Juelich. Screening tests (maximum heat flux of 23 MW/m 2 ) and thermal cycling tests (18 MW/m 2 , 30s, 1000cycle) were carried out. As a result of the screening tests, the erosion caused by sublimation of C/C was observed on the surface of armor tile. No serious damage such as cracks or detachments, however, were found. As a result of the thermal cycling tests, no major damage was detected on the C/C surface. However cooling time constant of the divertor mock-up increased over 600cycle. Therefore it implies that some defects would occur at the brazing interface of the divertor mock-up. (author)

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  20. Design and Analysis of the Korean Small Semi-prototype Mock-up for the 2nd Qualification of the ITER Blanket First Wall

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Yoon, Jae Sung; Lee, Eo Hwak; Lee, Seung Jae; Choi, Bo Guen; Park, Jeong Yong; Jung, Yang Il; Choi, Byung Kwon; Kim, Byoung Yoon

    2011-01-01

    Since the blanket First Wall (FW) of the International Thermonuclear Experimental Reactor (ITER) is subjected to a high heat and high neutron loads, it is one of the most important components. It composed of a beryllium (Be) layer as a plasma facing material, a copper alloy (CuCrZr) layer as a heat sink and type 316L authentic stainless steel (SS316L) as a structure material. The joining of the three different metals is the key issue to be solved. And more, the peak heat load was assumed to be 0.5 MW/m 2 in the initial design of the FW, but it was changed to be up to 5 MW/m 2 In Korea, the joining method has developed and it was proved through the several mock-up fabrication and high heat flux tests for confirming the joining integrity. Some of them were tested in the foreign facilities such as JEBIS at JAEA in Japan, TSEFEY at Efremov in Russia, and JUDITH at FZJ in Germany, and others were tested in our own facilities such as KoHLT-1 and -2. And finally, the 1 st , recently. Therefore, the FW panel design has been changed for enhancing the cooling and ITER Organization will provide the proposed design. Qualification was passed, in which two 80x80x3 Be/Cu/SS mock-ups were tested under 0.625 and 0.875 MW/m 2 heat fluxes for 12,000 cycles and then tested under 1.75 and 1.40 MW/m 2 Currently, the 2 heat fluxes for 1,000 cycles at FZJ and SNL, respectively. Currently, the 2 nd qualification program was started and the semi-prototype should be fabricated by the end of 2011 for testing under 5.0 MW/m 2 heat flux for certain number of cycles. In order to prepare the semi-prototype, several fabrication methods should be developed through the fabrication and test with the several mock-ups. In the present study, small Be mock-up was fabricated as the first step for the preparation. It was fabricated according to the designs considering the currently modified design of the FW. In the present paper, the fabrication objectives, methods, results and related tests were

  1. Adequacy of the analysis of mock-up control rod experiment with FCA

    International Nuclear Information System (INIS)

    Mizoo, Nobutatsu; Nakano, Masafumi

    1977-07-01

    A method of numerical analysis has been investigated for the mock-up control rod experiment of FCA VII-1 assembly constructed as the engineering mock-up of prototype fast breeder reactor MONJU. The results of criticality and B 4 C mock-up control rod worths analysis for the assembly are described in comparison with the experimental ones. The tendency of the C/E value with 10 B enrichment and the interaction effect of the multiple rods array was also examined. Reactivities and the mock-up rods worths were obtained with the X-Y geometry six groups diffusion theory. Twelve kinds of the mock-up rods with different 10 B contents and/or enrichments were used in the experiment; effective cross-sections are provided for each rod by calculation using the collision probability method. Criticality of VII-1 90Z assembly is underestimated for 3 reference critical configurations, ranging from -0.65%Δk/k to -0.77%Δk/k. The C/E values at core center for 12 kinds of B 4 C mock-up rods range from 1.03 to 1.09. The overestimate of the rod worth increases with macroscopic absorption cross-section of the rod region. The C/E values for 24 different arrays of the mock-up rods ranging from single rod to five rods lie between 1.04 and 1.08. The C/E value tends to decrease with increase in the number of rods inserted, the values for five rods arrays being about 4% lower than those for single rod arrays. The calculated interaction effects of the multiple rods arrays are slightly more negative than the experimental ones. (auth.)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  3. Mock-up critical experiments for prototype fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    Zukeran, Atsushi; Inoue, Teruji; Suzuki, Takeo; Kawashima, Kanau

    1976-01-01

    The mock-up criticality experiments for Monju are roughly divided into the full mock-up test using the ZEBRA of Winfrith Institute, UK AEA, and the partial mock-up experiment with FCA of JAERI. The former test has been carried out over 18 months from September 1971 as the Japan-UK cooperative research project MOZART. With the FCA, the experiment complementing the MOZART has been carried out, focusing on the nuclear characteristics of Monju which can be simulated with a relatively small core, and the experiment on highly enriched control rods and shielding is being continued now with the FCA 7 core. The experimental data of the MOZART and the ZPPR series in USA were exchanged at the international symposium in Tokyo, thus the prediction and the accuracy evaluation of the nuclear characteristics of Monju became possible, and the highly reliable core design was able to be accomplished. The simulated criticality experiment is necessary for directly grasping the reliability of calculated values in comparison with the experimental values, and also for the experimental prediction of the nuclear characteristics. The outline and the analysis of the simulated criticality experiment such as reactivity factor, control rod value, reaction rate distribution and sodium void reactivity are described, and the reflection of the results to the design of the core of Monju is explained. (Kako, I.)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  5. Validation results of satellite mock-up capturing experiment using nets

    Science.gov (United States)

    Medina, Alberto; Cercós, Lorenzo; Stefanescu, Raluca M.; Benvenuto, Riccardo; Pesce, Vincenzo; Marcon, Marco; Lavagna, Michèle; González, Iván; Rodríguez López, Nuria; Wormnes, Kjetil

    2017-05-01

    The PATENDER activity (Net parametric characterization and parabolic flight), funded by the European Space Agency (ESA) via its Clean Space initiative, was aiming to validate a simulation tool for designing nets for capturing space debris. This validation has been performed through a set of different experiments under microgravity conditions where a net was launched capturing and wrapping a satellite mock-up. This paper presents the architecture of the thrown-net dynamics simulator together with the set-up of the deployment experiment and its trajectory reconstruction results on a parabolic flight (Novespace A-310, June 2015). The simulator has been implemented within the Blender framework in order to provide a highly configurable tool, able to reproduce different scenarios for Active Debris Removal missions. The experiment has been performed over thirty parabolas offering around 22 s of zero-g conditions. Flexible meshed fabric structure (the net) ejected from a container and propelled by corner masses (the bullets) arranged around its circumference have been launched at different initial velocities and launching angles using a pneumatic-based dedicated mechanism (representing the chaser satellite) against a target mock-up (the target satellite). High-speed motion cameras were recording the experiment allowing 3D reconstruction of the net motion. The net knots have been coloured to allow the images post-process using colour segmentation, stereo matching and iterative closest point (ICP) for knots tracking. The final objective of the activity was the validation of the net deployment and wrapping simulator using images recorded during the parabolic flight. The high-resolution images acquired have been post-processed to determine accurately the initial conditions and generate the reference data (position and velocity of all knots of the net along its deployment and wrapping of the target mock-up) for the simulator validation. The simulator has been properly

  6. Mock-up experiment and analysis for the primary shield of the N.S. MUTSU

    International Nuclear Information System (INIS)

    Miyasaka, S.; Asaoka, T.; Taji, Y.; Ise, T.; Koyama, K.; Tsutsui, T.; Takeuchi, M.; Fuse, T.; Miura, T.; Yamaji, Y.

    1977-01-01

    A series of shielding mock-up experiments was performed at JRR-4, a swimming pool type reactor, of Japan Atomic Energy Research Institute (JAERI) to obtain the necessary experimental data and the sufficiently accurate method of calculation adopted for the modification of the MUTSU primary shield. Analyses for the experiments were carried out by using of the Ssub(n) codes, ANISN and TWOTRAN. The two dimensional calculations were performed with the P 1 -S 8 approximation. The neutron streaming through the annular gap between the pressure vessel and the primary shield has been confirmed to be estimated from the present method of calculation. The agreement between the calculated and the measured values is generally in about a factor of 2 to 4. (orig.) [de

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

    Science.gov (United States)

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

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  9. Dismantling Experiment of Mock-up Tube Bundle of Steam Generator

    International Nuclear Information System (INIS)

    Kim, Sung Kyun; Lee, Kune Woo

    2010-01-01

    A SG (steam generator) is one of the biggest decommissioning components in nuclear power plants and one has been replaced 2∼6 times during the whole operation of a nuclear power plant. The old SG should be decommissioned for the purpose of the volume reduction of radioactive waste. Among the components of SG, the tube bundle is one of the most difficult items to be dismantled due to the fact that it is very hard to cut since it is made of Inconel 600 which has high resistance of corrosion and abrasion. Moreover, All cutting process should be performed by remotely since radioactive contamination of the internal surface of SG tubes is very high (about 150,000∼300,000 Bq/cm 2 ). Therefore, it is necessary to choose the appropriate cutting methods by the pros and cons analysis for candidate dismantling technologies and to do experiment study for the validation. In this study, the results of cutting experiment for a mock-up bundle by using band saw cutting method are described herein

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

    International Nuclear Information System (INIS)

    Hashimoto, Toshiyuki; Takatsu, Hideyuki; Sato, Satoshi

    1994-07-01

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

  11. Mock-up experiments for the project of high dose irradiation on the RPV concrete

    International Nuclear Information System (INIS)

    Zdarek, J.; Brabec, P.; Frybort, O.; Lahodova, Z.; Vit, J.; Stemberk, P.

    2015-01-01

    Aging of NPP's concrete structures comes into growing interest in connection with solution of life extension programmes of operated units. Securing continued safe operation of NPPs calls for additional proofs of suitable long term behaviour of loaded reinforced concrete structures. An irradiation test of concrete samples was performed in the core of the LVR-15 reactor. The irradiation capsule was hung in the irradiation channel and the cooling of the capsule was ensured through direct contact of the capsule wall with the primary circuit water. Cylindrical, serpentine concrete samples (50 mm in diameter and 100 mm in length), representing composition of WWER RPV cavity, was chosen as a compromise of mechanical properties testing needs and dimension limitations of reactor irradiation channel. Heating during irradiation test was maintained under 93 Celsius degrees by cooling and was controlled by embedded thermocouple. Design of the cooling management was supported by computational analysis. The dependencies of heated concrete samples to the neutron fluence and the gamma heating were obtained by changing the thermal power of the reactor and by changing the vertical position of the sample in the irradiation channel. The irradiation capsule was filled with inert gas (helium) to allow the measurement of generated gas. The determination of concrete samples activity for long-term irradiation was performed on the principles of the Neutron Activation Analysis. Preliminary mock-up tests have proved the ability to fulfill technical needs for planned high dose irradiation experiment

  12. Design/build/mockup of the Waste Isolation Pilot Plant gas generation experiment glovebox

    International Nuclear Information System (INIS)

    Rosenberg, K.E.; Benjamin, W.W.; Knight, C.J.; Michelbacher, J.A.

    1996-01-01

    A glovebox was designed, fabricated, and mocked-up for the WIPP Gas Generation Experiments (GGE) being conducted at ANL-W. GGE will determine the gas generation rates from materials in contact handled transuranic waste at likely long term repository temperature and pressure conditions. Since the customer's schedule did not permit time for performing R ampersand D of the support systems, designing the glovebox, and fabricating the glovebox in a serial fashion, a parallel approach was undertaken. As R ampersand D of the sampling system and other support systems was initiated, a specification was written concurrently for contracting a manufacturer to design and build the glovebox and support equipment. The contractor understood that the R ampersand D being performed at ANL-W would add additional functional requirements to the glovebox design. Initially, the contractor had sufficient information to design the glovebox shell. Once the shell design was approved, ANL-W built a full scale mockup of the shell out of plywood and metal framing; support systems were mocked up and resultant information was forwarded to the glovebox contractor to incorporate into the design. This approach resulted in a glovebox being delivered to ANL-W on schedule and within budget

  13. Mock-up experiment at Birmingham University for BNCT project of Osaka University – Neutron flux measurement with gold foil

    International Nuclear Information System (INIS)

    Tamaki, S.; Sakai, M.; Yoshihashi, S.; Manabe, M.; Zushi, N.; Murata, I.; Hoashi, E.; Kato, I.; Kuri, S.; Oshiro, S.; Nagasaki, M.; Horiike, H.

    2015-01-01

    Mock-up experiment for development of accelerator based neutron source for Osaka University BNCT project was carried out at Birmingham University, UK. In this paper, spatial distribution of neutron flux intensity was evaluated by foil activation method. Validity of the design code system was confirmed by comparing measured gold foil activities with calculations. As a result, it was found that the epi-thermal neutron beam was well collimated by our neutron moderator assembly. Also, the design accuracy was evaluated to have less than 20% error. - Highlights: • Accelerator based neutron source for BNCT is being developed in Osaka University. • Mock-up experiment was carried out at Birmingham University, UK. • Neutronics performance of our assembly was evaluated from gold foil activation. • Gold foil activation was determined by using HPGe detectors. • Validity of the neutronics design code system was confirmed.

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  15. Required momentum, heat, and mass transport experiments for liquid-metal blankets

    International Nuclear Information System (INIS)

    Tillack, M.S.; Sze, D.K.; Abdou, M.A.

    1986-01-01

    Through the effects on fluid flow, many aspects of blanket behavior are affected by magnetohydrodynamic (MHD) effects, including pressure drop, heat transfer, mass transfer, and structural behavior. In this paper, a set of experiments is examined that could be performed in order to reduce the uncertainties in the highly related set of issues dealing with momentum, heat, and mass transport under the influence of a strong magnetic field (i.e., magnetic transport phenomena). By improving our basic understanding and by providing direct experimental data on blanket behavior, these experiments will lead to improved designs and an accurate assessment of the attractiveness of liquid-metal blankets

  16. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    Directory of Open Access Journals (Sweden)

    Krása Antonín

    2017-01-01

    Full Text Available VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector. Discrepancies between experiments and Monte Carlo calculations (MCNP5 of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2 are presented.

  17. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    Science.gov (United States)

    Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

    2017-09-01

    VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

  18. Pyroprocess Deployment Analysis and Remote Accessibility Experiment using Digital Mockup and Simulation

    International Nuclear Information System (INIS)

    Kim, K. H.; Park, H. S.; Kim, S. H.; Choi, C. H.; Lee, H. J.; Park, B. S.; Yoon, G. S.; Kim, K. H.; Kim, H. D.

    2009-11-01

    Nuclear fuel cycle facility that treats with spent fuel must be designed and manufactured a Pyroprcess facility and process with considering a speciality as every process have to be processed remotely. To prevent an unexpected accident under a circumstance that must operate with a remote manipulator after done the Pyroprocess facility, an procedure related Pyroprocess operation and maintenance need to establish it in the early design stage. To develop the simulator that is mixed by 3D modelling and simulation, a system architecture was designed. A full-scale digital mockup with a real pyroprocess facility was designed and manufactured. An inverse kinematics algorithm of remote manipulator was created in order to simulate an accident and repair that could happen in pyroprocess operation and maintenance under a virtual digital mockup environment. Deployment analysis of process devices through a workspace analysis was carried out and Accessibility analysis by using haptic device was examined

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

    International Nuclear Information System (INIS)

    Weth, A. von der; Aktaa, J.

    2008-01-01

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

  20. Engineering test station for TFTR blanket module experiments

    International Nuclear Information System (INIS)

    Jassby, D.L.; Leinoff, S.

    1979-12-01

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

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

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1992-12-01

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

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

    International Nuclear Information System (INIS)

    Woodruff, G.L.

    1987-01-01

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

  3. A new blanket tritium recovery experiment with intense DT neutron source at JAEA/FNS

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Kentaro, E-mail: ochiai.kentaro@jaea.go.jp [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Edao, Yuki [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Hoshino, Tsuyoshi [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Kawamura, Yoshinori [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Ohta, Masayuki; Kwon, Saerom; Konno, Chikara [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan)

    2016-11-01

    Highlights: • For detail investigation of the tritium recovery performance on the fusion reactor blanket, we have started a new blanket tritium recovery experiment with ionization chamber at JAEA/FNS. • A new improved container was provided for the appropriate tritium measurement by IC and also utilized for the enhancement of TPR in the new container. The TPR was calculated with a calculation code MCNP5 and some typical nuclear data libraries and then the radioactivity of the tritium recovery with LSC corresponded with that of calculation. • The tritium release curves by the IC outputs are similar to those by the LSC output. However, it was indicated that the quantitative measurement by IC needed further improvement for the tritium recovery. - Abstract: We have performed the tritium release experiment on the fusion reactor blanket at JAEA/FNS since 2009, and then clarified the ratio of tritium release and the recovered tritium chemical form. In order to acquire the detailed tritium recovery performances, we have started a new blanket tritium recovery experiment with ionization chamber (IC) at JAEA/FNS. For the appropriate tritium measurement with IC, we improved the experimental container and carried out with an intense DT neutron source at JAEA/FNS. From our new experiment, the tritium recovery radioactivity from the LSC measurement corresponds with the calculation within 6%. However, it was pointed out that further improvement in the quantitative tritium measurement by IC method was needed.

  4. LMFBR blanket physics project progress report No. 4

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Lanning, D.D.; Kaplan, I.; Supple, A.T.

    1973-01-01

    During the period covered by the report, July 1, 1972, through June 30, 1973, work was devoted to completion of experimental measurements and data analysis on Blanket Mockup No. 3, a graphite-reflected blanket, and to initiation of experimental work on Blanket Mockup No. 4, a steel-reflected assembly designed to mock up a demonstration plant blanket. Work was also carried out on the analysis of a number of advanced blanket concepts, including the use of high-albedo reflectors, the use of thorium in place of uranium in the blanket region, and the ''parfait'' or completely internal blanket concept. Finally, methods development work was initiated to develop the capability for making gamma heating measurements in the blanket mockups. (U.S.)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  7. Analysis of the impacts of the J-TEXT TBM mock-up on the equilibrium magnetic field

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Zhengqing [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, and College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Rao, Bo, E-mail: borao@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, and College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Zhang, Ming; Zhang, Jun [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, and College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Wang, Weihua [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); New Star Institute of Applied Technology, Hefei 230031 (China); Liu, Sumei [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); School of Engineering,Anhui Agricultural University, Hefei 230036 (China)

    2016-11-01

    Highlights: • J-TEXT TBM mock-up was designed and fabricated to test and study the distribution of eddy current, electromagnetic and thermal load on the TBM during plasma disruption. • This paper focuses on evaluating the influence of the TBM structural material (RAMF steel) to tokamak discharge and security. The simulation data presents a relatively complete assessment of impacts of the J-TEXT TBM mock-up on the equilibrium magnetic field. • The conclusion of the simulation will offer the guidance for installation interface design of the TBM mock-up. - Abstract: The Test Blanket Module (TBM) will be used in the test port of ITER to demonstrate tritium self-sufficiency and the extraction of high grade heat for electricity production. J-TEXT TBM mock-up using reduced activation ferritic/martensitic (RAFM) steel as structural material was designed and fabricated to perform and validate relevant electromagnetic and thermal technologies of the China Helium-Cooled Ceramic Breeder Test Blanket Module (CN HCCB-TBM) on the J-TEXT. Its size is one third of the CN HCCB-TBM. By using the finite element analysis technology, this paper analyzed the impacts on the equilibrium magnetic field over the plasma region after introducing the structure material RAFM steel. The distribution of toroidal field (TF) ripple and the magnitude of the error field with the mock-up at different positions were given. Simulation shows the distribution of the null field region formed by poloidal field (PF). The influence to tokamak discharge has been evaluated by drawing the magnetic field lines. Based on the results above, we have optimized and finished the installation of the mock-up to J-TEXT which meets the needs of the experiments and to ensure the normal discharge.

  8. Qualification test for ITER HCCR-TBS mockups with high heat flux test facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon, E-mail: skkim93@kaeri.re.kr [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Seong Dae; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated. • A thermo-hydraulic analysis was performed using a high heat flux test facility by using electron beam. • The plan for qualification tests was developed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization. - Abstract: The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated, and an integrity and thermo-hydraulic performance test should be completed under the same or similar operation conditions of ITER. The test plan for a thermo-hydraulic analysis was developed by using a high heat flux test facility, called the Korean heat load test facility by using electron beam (KoHLT-EB). This facility is utilized for a qualification test of the plasma facing component (PFC) for the ITER first wall and DEMO divertor, and for the thermo-hydraulic experiments. In this work, KoHLT-EB will be used for the plan of the performance qualification test of the ITER HCCR-TBS mockups. This qualification tests should be performed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization (IO), which describe the specifications and qualifications of the heat flux test facility and test procedure for ITER PFC.

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

    International Nuclear Information System (INIS)

    Youssef, M.Z.

    1992-12-01

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

  10. Fission blanket benchmark experiment on spherical assembly of uranium and PE with PE reflector

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Tonghua; Lu, Xinxin; Wang, Mei; Han, Zijie, E-mail: neutron_integral@aliyun.com; Jiang, Li; Wen, Zhongwei; Liu, Rong

    2016-04-15

    Highlights: • The fission rate distribution on two depleted uranium assemblies was measured with plate fission chambers. • We do calculations using MCNP code and ENDF/B-V.0 library. • The overestimation of calculations to the measured fission rates was found. • The observed discrepancy are discussed. - Abstract: New concept of fusion-fission hybrid for energy generation has been proposed. To validate the nuclear performance of fission blanket of hybrid, as part of series of validation experiment, two types of fission blanket assemblies were setup in this work and measurements were made of the reaction rate distribution for uranium fission in the spherical assembly of depleted uranium and polyethylene by Plate Fission Chamber (PFC). There are two PFCs in experiment, one is depleted uranium chamber and the other is enriched uranium chamber. The Monte-Carlo transport code MCNP5 and continuous energy cross sections library ENDF/BV.0 were used for the analysis of fission rate distribution in the two types of assemblies. The calculated results were compared with the experimental ones. The overestimation of fission rate for depleted uranium and enriched uranium were found in the inner boundary of the two assemblies. However, the C/E ratio tends to decrease for the distance from the core slightly and the results for enriched uranium are better than that for depleted uranium.

  11. The design, fabrication and installation of cable routing mockups in support of Spacelab 2

    Science.gov (United States)

    1981-01-01

    From flight and mockup drawings of Spacelab 2 (SL 2) experiments and hardware, shop ready mockup drawings were produced. Floor panels were the first items considered for fabrication. Cold plate and orthogrid mockups were designed and fabricated. Experiment and other hardware mockups were fabricated of aluminum or plywood, depending on size and configuration. Eighty-three cable routing bracket mockups were fabricated of aluminum and delivered for painting.

  12. Experimental studies on tungsten-armour impact on nuclear responses of solid breeding blanket

    International Nuclear Information System (INIS)

    Sato, Satoshi; Nakao, Makoto; Verzilov, Yury; Ochiai, Kentaro; Wada, Masayuki; Kubota, Naoyoshi; Kondo, Keitaro; Yamauchi, Michinori; Nishitani, Takeo

    2005-01-01

    In order to experimentally evaluate the tungsten armour impact on tritium production of the solid breeding blanket being developed by JAERI for tokamak-type DEMO reactors, neutronics integral experiments have been performed using DT neutrons at the Fusion Neutron Source facility of JAERI. Solid breeding blanket mockups relevant to the DEMO blanket have been applied in this study. The mockups are made of a set of layers consisting of 0-25.2 mm thick tungsten, 16 mm thick F82H, 12 mm thick Li 2 TiO 3 and 100-200 mm thick beryllium with a cross-section of 660 x 660 mm in maximum. Pellets of Li 2 CO 3 are embedded in the Li 2 TiO 3 layers to measure the tritium production rate. By installing the 5 mm, 12.6 mm and 25.2 mm thick tungsten armours, the sum of the integrated tritium productions at the pellets are reduced by about 2.1%, 2.5% and 6.1% relative to the case without the armour, respectively. Numerical calculations have been conducted using the Monte Carlo code. In the case of the mockups with the tungsten armour, calculation results for the sum of the integrated tritium productions agree well with the experimental data within 4% and 19% in the experiments without and with a neutron reflector, respectively

  13. Experimental studies on tungsten-armour impact on nuclear responses of solid breeding blanket

    Science.gov (United States)

    Sato, Satoshi; Nakao, Makoto; Verzilov, Yury; Ochiai, Kentaro; Wada, Masayuki; Kubota, Naoyoshi; Kondo, Keitaro; Yamauchi, Michinori; Nishitani, Takeo

    2005-07-01

    In order to experimentally evaluate the tungsten armour impact on tritium production of the solid breeding blanket being developed by JAERI for tokamak-type DEMO reactors, neutronics integral experiments have been performed using DT neutrons at the Fusion Neutron Source facility of JAERI. Solid breeding blanket mockups relevant to the DEMO blanket have been applied in this study. The mockups are made of a set of layers consisting of 0-25.2 mm thick tungsten, 16 mm thick F82H, 12 mm thick Li2TiO3 and 100-200 mm thick beryllium with a cross-section of 660 × 660 mm in maximum. Pellets of Li2CO3 are embedded in the Li2TiO3 layers to measure the tritium production rate. By installing the 5 mm, 12.6 mm and 25.2 mm thick tungsten armours, the sum of the integrated tritium productions at the pellets are reduced by about 2.1%, 2.5% and 6.1% relative to the case without the armour, respectively. Numerical calculations have been conducted using the Monte Carlo code. In the case of the mockups with the tungsten armour, calculation results for the sum of the integrated tritium productions agree well with the experimental data within 4% and 19% in the experiments without and with a neutron reflector, respectively.

  14. Measurements of electron density and temperature profiles in a gas blanket experiment

    International Nuclear Information System (INIS)

    Kuthy, A.

    1979-02-01

    Radial profiles of electron density, temperature and H sub(β) intensity are presented for the rotating plasma device F-1. The hydrogen filling pressure, the average magnetic field strength at the midplane, and the power input to the discharge have been varied in the ranges 10-100 mTorr, 0.25-0.5 Tesla, and 0.1 to 1.5 MW, respectively. These experiments have been performed with the main purpose of studying the gas blanket (cold-mantle) state of the plasma. It is shown, that a simple spectroscopic method can be used to derive the radial distribution of the electron temperature in such plasmas. The observed peak temperatures and densities are in agreement with earlier theoretical estimates. (author)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  16. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  17. CANDU 9 Control Centre Mockup

    International Nuclear Information System (INIS)

    Webster, A.; Macbeth, M.J.

    1996-01-01

    This paper provides a summary of the design process being followed, the benefits of applying a systematic design using human factors engineering, presents an overview of the CANDU 9 control centre mockup facility, illustrates the control centre mockup with photographs of the 3D CADD model and the full scale mockup, and provides an update on the current status of the project. (author)

  18. Fusion fuel blanket technology

    International Nuclear Information System (INIS)

    Hastings, I.J.; Gierszewski, P.

    1987-05-01

    The fusion blanket surrounds the burning hydrogen core of a fusion reactor. It is in this blanket that most of the energy released by the nuclear fusion of deuterium-tritium is converted into useful product, and where tritium fuel is produced to enable further operation of the reactor. As fusion research turns from present short-pulse physics experiments to long-burn engineering tests in the 1990's, energy removal and tritium production capabilities become important. This technology will involve new materials, conditions and processes with applications both to fusion and beyond. In this paper, we introduce features of proposed blanket designs and update and status of international research. In focusing on the Canadian blanket technology program, we discuss the aqueous lithium salt blanket concept, and the in-reactor tritium recovery test program

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

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1989-12-01

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

  20. Proposal for a blanket purchase agreement for the supply and repair of subracks for the LHC experiments

    CERN Document Server

    2002-01-01

    This document concerns the award of a blanket purchase agreement for the supply and repair of subracks for the LHC experiments. Following a market survey carried out among 27 firms in seven Member States and one firm in a non-Member State, a call for tenders (IT-2916/EP) was sent on 9 November 2001 to 16 firms in five Member States. By the closing date, CERN had received six tenders. The Finance Committee is invited to agree to the negotiation of a blanket purchase agreement with WIENER, PLEIN & BAUS (DE) for the supply of subracks for a period of four years and a repair service for a period of ten years after expiry of the initial two year guarantee period, for a total amount not exceeding 5 600 000 euros, subject to revision for inflation from 1 January 2003. At the present rate of exchange, the total amount of the blanket purchase agreement is equivalent to approximately 8 300 000 Swiss francs. This requirement will be financed by the collaborating institutes of the LHC experiments and by CERN. CERN's ...

  1. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

    Forbes, I.A.; Driscoll, M.J.; Rasmussen, N.C.; Lanning, D.D.; Kaplan, I.

    1971-01-01

    This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238 U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics

  2. Experimental studies on tungsten-armor impact on nuclear responses of solid breeding blanket

    International Nuclear Information System (INIS)

    Sato, S.; Nakao, M.; Verzilov, Y.; Ochiai, K.; Wada, M.; Kubota, N.; Kondo, K.; Yamauchi, M.; Enoeda, M.; Nishitani, T.

    2005-01-01

    In order to experimentally evaluate the tungsten armor impact on tritium production of the solid breeding blanket being developed by JAERI for tokamak-type DEMO reactors, neutronics integral experiments have been performed by using DT neutrons at Fusion Neutron Source (FNS) facility of JAERI. Solid breeding blanket mockups relevant to the DEMO blanket have been applied in this study. The mockups are constructed by a set of layers consisting of 0 - 25.2 mm thick tungsten, 16 mm thick F82H, 12 mm thick Li 2 TiO 3 and 100 - 200mm thick beryllium with cross-section of 660 x 660 mm in maximum. Pellets of Li 2 CO 3 are embedded inside the Li 2 TiO 3 layers to measure the tritium production rate. By installing the 5, 12.6 and 25.2 mm thick tungsten armors, sum of the integrated tritium productions at the pellets are reduced by about 2, 3 and 6 % relative to the case without the armor, respectively. Numerical calculations have been conducted using the Monte Carlo code. Calculation results for sum of the integrated tritium productions in the case with the tungsten armor agree well with the experiment data within 4% and 19% under condition without and with a neutron reflector, respectively. (author)

  3. Role of irradiation reactor mock-ups

    International Nuclear Information System (INIS)

    Casali, F.; Cerles, J.M.; Debrue, J.

    1977-01-01

    A survey is given of the utilization of low power facilities in support to irradiation reactor experiments. The BRO2, ISIS and RB3 facilities are described as neutronic mock-ups of the BR2, OSIRIS and ESSOR reactors respectively

  4. Non destructive testing of concrete nuclear containment plants with surface waves: Lab experiment on decimeter slabs and on the VeRCoRs mock-up

    Science.gov (United States)

    Abraham, Odile; Legland, Jean-Baptiste; Durand, Olivier; Hénault, Jean-Marie; Garnier, Vincent

    2018-04-01

    The maintenance and evaluation of concrete nuclear containment walls is a major concern as they must, in case of an accident, ensure the confinement of the nuclear radiations and resist to the loads. A homemade multi-receiver multi-source dry contact linear probe to record ultrasonic surface waves on concrete in the frequency range [60 kHz - 200 kHz] has been used in this context. The measurement protocol includes the summation of up to 50 spatially distributed seismograms and the determination of the surface waves phase velocity dispersion curve. The probe has been tested against several concrete states under no loading (water saturation level, temperature damage). Then, the same measurements have been performed on sound and fire damaged slabs submitted to uniaxial loading (stress up to 30 % of the concrete compression resistance). It is shown that the robustness and precision of the surface waves measurement protocol make it possible to follow the stress level. In March 2017 a first experiment with this surface wave probe has been conducted on a reduced 1:3 scale nuclear containment plant (EDF VeRCoRs mock-up) under loading conditions that replicates that of decennial inspection. The surface wave phase velocity dispersion curves of each state are compared and cross-validated with other NDT results.

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

    International Nuclear Information System (INIS)

    Kodeli, I.

    2005-01-01

    In order to validate new nuclear cross-section evaluations, method development and design of the helium-cooled pebble bed (HCPB) test blanket module of ITER a benchmark experiment was performed this year at the Frascati Neutron Generator (FNG) in the scope of the EFF (European Fusion File) project in Europe. The objective of this experiment is to study the tritium breeding ratio and other nuclear quantities in a breeder blanket in order to establish and improve the quality of related JEFF nuclear data. The experiment consists of a metallic beryllium set-up with two double layers of breeder material (Li 2 CO 3 powder). The reaction rate measurements include the Li 2 CO 3 pellets (tritium breeding ratio), activation foils, and neutron and gamma spectrometers inserted at several axial and lateral locations in the block. Our task is to perform the deterministic transport, and cross section sensitivity and uncertainty analysis. The role of the cross-section sensitivity and uncertainty analysis is to optimise the design of the benchmark, and to assist in the interpretation of the measurement results. The paper presents the pre- and post- analysis of the benchmark experiment. (author)

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

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2004-07-01

    This report is the Proceedings of 'the Eleventh International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. This workshop was held in Tokyo, Japan on December 15-17, 2003. About thirty experts from China, EU, Japan, Korea, Latvia, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of Li 2 TiO 3 and so on, fabrication technology developments and characterization of the Li 2 TiO 3 and Li 4 SiO 4 pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li 2 TiO 3 and Li 4 SiO 4 pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system. The 26 of the presented papers are indexed individually. (J.P.N.)

  7. Spectrum unfolding from activation measurements in a CTR-model blanket experiment

    International Nuclear Information System (INIS)

    Kuijpers, L.J.M.

    1977-07-01

    Neutron spectra in a lithium fusion reactor model blanket are determined experimentally by performing SAND II unfolding runs from measured activities. The principles of the iterative SAND II method are given and characteristics of the output are described. The spectra are calculated from available data with the aid of a Monte Carlo program, of which procedure numerical results are given. Both kinds of spectra are compared; when number of input data is varied or different cross section data sets are chosen, inconsistencies in activities or cross section data may be detected. (orig./WL) [de

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  9. ANL ITER high-heat-flux blanket-module heat transfer experiments

    International Nuclear Information System (INIS)

    Kasza, K.E.

    1992-02-01

    An Argonne National Laboratory facility for conducting tests on multilayered slab models of fusion blanket designs is being developed; some of its features are described. This facility will allow testing under prototypic high heat fluxes, high temperatures, thermal gradients, and variable mechanical loadings in a helium gas environment. Steady and transient heat flux tests are possible. Electrical heating by a two-sided, thin stainless steel (SS) plate electrical resistance heater and SS water-cooled cold panels placed symmetrically on both sides of the heater allow achievement of global one-dimensional heat transfer across blanket specimen layers sandwiched between the hot and cold plates. The heat transfer characteristics at interfaces, as well as macroscale and microscale thermomechanical interactions between layers, can be studied in support of the ITER engineering design effort. The engineering design of the test apparatus has shown that it is important to use multidimensional thermomechanical analysis of sandwich-type composites to adequately analyze heat transfer. This fact will also be true for the engineering design of ITER

  10. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  11. Proposal to negotiate an amendment to an existing blanket purchase contract for the supply of Burndy connectors for the LHC project and LHC experiments

    CERN Document Server

    2006-01-01

    This document concerns the proposal to negotiate an amendment to an existing blanket purchase contract for the supply of Burndy connectors for the LHC project and LHC experiments. For the reasons explained in this document, the Finance Committee is invited to agree to the negotiation of an amendment to the blanket purchase contract for the supply of Burndy connectors for the LHC project and LHC experiments with the company SOURIAU (CH), for the next three years for up to 600 000 euros (954 600 Swiss francs), subject to revision for inflation from January 2007, bringing the total amount of the blanket purchase contract to a maximum amount of 1 200 000 euros (1 909 200 Swiss francs), subject to revision for inflation from January 2007. The amounts in Swiss francs have been calculated using the present rate of exchange. 2006/60/5/e

  12. Flow split, pressure drop, and mixing experiments in a 61-pin shaved-wire blanket assembly

    International Nuclear Information System (INIS)

    Wang, S.F.; Rohsenow, W.M.; Todreas, N.E.

    1981-10-01

    The experimental results of a series of hydraulic tests on a shaved-wire 61-pin blanket assembly are presented and compared to previously published MIT results for full-wire tests. The reduction of the wall subchannel flow area by incorporating the shaved-wire reduced the flow in edge and corner subchannels by 4.5 to 16.5% and 1.5%, respectively. The swirl velocity in the wall subchannels was reduced by 50%. The bundle average friction factor, the turbulent subchannel friction factors, and the mixing parameter and laminar/transition friction factors in the interior subchannels remained unchanged. However, the laminar/transition friction factors in the edge subchannels were significantly increased. These effects generally cannot be predicted by the current flow split, friction factor, and mixing parameter correlations

  13. Siloette, Siloe mock-up

    International Nuclear Information System (INIS)

    Delcroix, V.; Jeanne, G.; Mitault, G.; Schulhof, P.

    1964-01-01

    Siloette is the Siloe mock-up. The main installations are described: various tanks, building, auxiliaries, control systems... Precis ions are given about precautions taken for using spent fuel elements. (authors) [fr

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  15. Collection of Summaries of reports on result of research at basic experiment device for nuclear fusion reactor blanket design, 1994

    International Nuclear Information System (INIS)

    1995-07-01

    The development of nuclear fusion reactors reached such stage that the generation of fusion power output comparable with the input power into core plasma is possible. At present, the engineering design of the international thermonuclear fusion experimental reactor, ITER, is advanced by the cooperation of Japan, USA, Europe and Russia, aiming at the start of operation at the beginning of 21st century. This meeting for reporting the results has been held every year, and this time, it was held on May 19, 1995 at University of Tokyo with the theme ''The interface properties of fusion reactor materials and the control of particle transport''. About 50 participants from academic, governmental and industrial circles discussed actively on the theme. Three lectures on the topics of fusion reactor engineering and materials and seven lectures on the basic experiment of fusion reactor blanket design related to the next period project were given at the meeting. (K.I.)

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

    International Nuclear Information System (INIS)

    1978-01-01

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

  17. 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.

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

    International Nuclear Information System (INIS)

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

    1988-03-01

    We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs

  19. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1983-08-01

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

  20. Proposal to negotiate an amendment to an existing blanket purchase contract for the supply and repair of subracks for the LHC experiments

    CERN Document Server

    2006-01-01

    This document concerns the proposal to negotiate an amendment to an existing blanket purchase contract for the supply and repair of subracks for the LHC experiments. For the reasons explained in this document, the Finance Committee is invited to agree to the negotiation of an amendment to the blanket purchase contract for the supply and repair of subracks for the LHC experiments with the company WIENER, PLEIN & BAUS (DE), for an extension of the period of supply from four to six years for an amount exceeding the previously authorised amount of 5 600 000 euros, subject to revision for inflation from January 2003, by up to 1 880 000 euros, subject to revision for inflation, bringing the total amount of the blanket purchase contract to a maximum amount of 7 480 000 euros, subject to revision for inflation. At the present rate of exchange, the total amended amount of the blanket purchase contract is equivalent to approximately 11 800 000 Swiss francs. CERN's total financial contribution to the funding of the ...

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

    Science.gov (United States)

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

    2003-12-01

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

  2. Carbon tiles as spectral-shifter for long-life liquid blanket in LHD-type reactor FFHR

    International Nuclear Information System (INIS)

    Sagara, A.; Imagawa, S.; Tanaka, T.; Muroga, T.; Kubota, Y.; Dolan, T.; Hashizume, H.; Kunugi, T.; Fukada, S.; Shimizu, A.; Terai, T.; Mitarai, O.

    2006-01-01

    In terms of engineering feasibility for long-life Flibe blanket in LHD-type reactor FFHR, the Spectral-shifter and Tritium breeder Blanket (STB) concept is evaluated by taking neutron irradiation effects into account under system integration such as Flibe cooling and components replacement. FEM calculations for the neutron wall loading of 1.5 MW/m 2 show that the temperature of the STB armor tile can be kept below 2000 K by optimizing the first metal wall thickness. The heat load experiment on the STB armor mockup confirms feasibility of the temperature control and mechanical joining. Degradation of STB armor tiles due to neutron irradiation requires replacement of them every few years by means of remote handling 'screw coasters' using helical winding, where the replaced tiles are low level wastes. Although the STB concept is feasible within nuclear and thermal properties, more detailed structural optimization is needed including the mechanical and chemical properties

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

    International Nuclear Information System (INIS)

    1996-07-01

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

  4. Feasibility study of the thermo-siphon mock-up test

    International Nuclear Information System (INIS)

    Choi, Jung Woon; Kim, Young Jin; Lee, Kye Hong; Kim, Young Ki; Jeong, Sang Kwon

    2004-09-01

    Described is the feasibility of the thermo-siphon mock-up test for the HANARO-CNS facility. The purposes of the mock-up tests are discussed in detail as the three concepts: for the detailed design, for the operation of the CNS facility, for the safety assurance of itself. This report considers the two stages of mock-up tests in terms of the experimental schedule and plan. As the first stage, the small-size mock-up test using Argon will be implemented to obtain the experience in the cryogenic fluid and to understand the basic concept of the CNS thermo-siphon. In the second stage, two kinds of mock-up tests are discussed: the full-scale mock-up test using liquid hydrogen or the integrated final test using hydrogen outside the reactor after the full-scale mock-up test using Freon gas. The contents discussed in this report will be the basis or the guide lines for the mock-up test. In addition, the results of the mock-up test will be the foundation for the safe operation of the HANARO-CNS facility

  5. EU contribution to the procurement of the ITER blanket first wall

    International Nuclear Information System (INIS)

    Lorenzetto, Patrick; Banetta, Stefano; Bellin, Boris; Boireau, Bruno; Bucci, Philippe; Cicero, Tindaro; Conchon, Denis; Dellopoulos, Georges; Hardaker, Stephen; Marshall, Paul; Nogué, Patrice; Pérez, Marcos; Gutierrez, Leticia Ruiz; Samaniego, Fernando; Sherlock, Paul; Zacchia, Francesco

    2016-01-01

    Highlights: • Presentation of the blanket first wall design concept to be procured by Europe. • Presentation of the main outcome of the R&D programme with the resulting FW fabrication route. • Presentation of the ITER first wall pre-qualification programme with the results achieved so far. • Presentation of the on-going irradiation experiments. • Presentation of the EU procurement strategy. - Abstract: Fusion for Energy (F4E), the European Union’s Domestic Agency for ITER, is responsible for the procurement of about 50% of the ITER blanket first wall (FW), called normal heat flux FW. A procurement strategy has been implemented by the In-Vessel Project Team at F4E aimed at mitigating technical and commercial risks for the procurement of ITER blanket FW panels, promoting as far as possible competition among industrial partners. This procurement strategy has been supported by an extensive Research and Development (R&D) programme, implemented over more than 15 years in Europe, to develop various fabrication technologies. It includes in particular the manufacture and testing of small-scale, medium-scale mock-ups and full-scale prototypes of blanket FW panels. In this R&D programme, significant efforts have been devoted to the development of a reliable materials joining technique. Hot Isostatic Pressing was selected for the manufacture of the FW panels made from beryllium, copper–chromium–zirconium alloy and 316L(N)-IG austenitic stainless steel. This paper presents the main outcome of the on-going R&D programme, the latest results of the FW qualification programme together with the procurement strategy implemented by F4E for the supply of the European contribution to the procurement of the ITER blanket FW.

  6. EU contribution to the procurement of the ITER blanket first wall

    Energy Technology Data Exchange (ETDEWEB)

    Lorenzetto, Patrick, E-mail: Patrick.Lorenzetto@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Banetta, Stefano; Bellin, Boris [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Boireau, Bruno [AREVA NP, Centre Technique, 71200 Le Creusot (France); Bucci, Philippe [Atmostat, rue René Hamon 31, 94815 Villejuif Cedex (France); Cicero, Tindaro [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Conchon, Denis [Atmostat, rue René Hamon 31, 94815 Villejuif Cedex (France); Dellopoulos, Georges [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Hardaker, Stephen [Amec Foster Wheeler plc, Booths Park, Chelford Road, Knutsford WA16 8QZ (United Kingdom); Marshall, Paul [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain); Nogué, Patrice [AREVA NP, Centre Technique, 71200 Le Creusot (France); Pérez, Marcos [Leading Enterprises SL, Pasaje de La Agüera, 39409 San Felices de Buelna (Spain); Gutierrez, Leticia Ruiz [Iberdrola Ingeniería y Construcción S.A.U., Avenida Manoteras 20, 28050 Madrid (Spain); Samaniego, Fernando [Leading Enterprises SL, Pasaje de La Agüera, 39409 San Felices de Buelna (Spain); Sherlock, Paul [Amec Foster Wheeler plc, Booths Park, Chelford Road, Knutsford WA16 8QZ (United Kingdom); Zacchia, Francesco [Fusion for Energy, Torres Diagonal Litoral B3, Carrer Josep Plà 2, B-08019 Barcelona (Spain)

    2016-11-01

    Highlights: • Presentation of the blanket first wall design concept to be procured by Europe. • Presentation of the main outcome of the R&D programme with the resulting FW fabrication route. • Presentation of the ITER first wall pre-qualification programme with the results achieved so far. • Presentation of the on-going irradiation experiments. • Presentation of the EU procurement strategy. - Abstract: Fusion for Energy (F4E), the European Union’s Domestic Agency for ITER, is responsible for the procurement of about 50% of the ITER blanket first wall (FW), called normal heat flux FW. A procurement strategy has been implemented by the In-Vessel Project Team at F4E aimed at mitigating technical and commercial risks for the procurement of ITER blanket FW panels, promoting as far as possible competition among industrial partners. This procurement strategy has been supported by an extensive Research and Development (R&D) programme, implemented over more than 15 years in Europe, to develop various fabrication technologies. It includes in particular the manufacture and testing of small-scale, medium-scale mock-ups and full-scale prototypes of blanket FW panels. In this R&D programme, significant efforts have been devoted to the development of a reliable materials joining technique. Hot Isostatic Pressing was selected for the manufacture of the FW panels made from beryllium, copper–chromium–zirconium alloy and 316L(N)-IG austenitic stainless steel. This paper presents the main outcome of the on-going R&D programme, the latest results of the FW qualification programme together with the procurement strategy implemented by F4E for the supply of the European contribution to the procurement of the ITER blanket FW.

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  8. Mockup experiments to investigate the leak rate correlation between mercury and helium for the mercury target system of J-PARC

    International Nuclear Information System (INIS)

    Haga, Katsuhiro; Naoe, Takashi; Kogawa, Hiroyuki; Wakui, Takashi; Futakawa, Masatoshi

    2009-01-01

    Checking the seal performance of the mercury piping network is very important for the mercury target system operation of J-PARC, and the test method for leaks using the pressure change measurement is preferable for this purpose because it can be carried out easily and precisely by measuring the pressure change, and it is free from the risk of mercury contamination. The piping network is pressurized by helium gas. Thus, the correlation between the helium leak rate and mercury leak flow rate was investigated experimentally by carrying out leak tests for helium and mercury with an identical mockup flange model. The results showed that the mercury leak flow rates of the experimental data were lower than those of the estimated value by 64% on average. It was also found that the threshold of the helium leak rate at which good seal performance for mercury can be obtained exists between 2.18 x 10 -4 and 1.01 x 10 -2 Pa.m 3 /s. This fact confirmed the sufficient safety margin of the mercury target system against the mercury leak, where 1 x 10 -6 Pa.m 3 /s was adopted as the seal performance criterion. (author)

  9. In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions

    Science.gov (United States)

    Litunovsky, N.; Gervash, A.; Lorenzetto, P.; Mazul, I.; Melder, R.

    2009-04-01

    The paper describes the experimental technique and preliminary results of thermal fatigue testing of ITER first wall (FW) water-cooled mock-ups inside the core of the RBT-6 experimental fission reactor (RIAR, Dimitrovgrad, Russia). This experiment has provided simultaneous effect of neutron fluence and thermal cycling damages on the mock-ups. A PC-controlled high-temperature graphite ohmic heater was applied to provide cyclic thermal load onto the mock-ups surface. This experiment lasted for 309 effective irradiation days with a final damage level (CuCrZr) of 1 dpa in the mock-ups. About 3700 thermal cycles with a heat flux of 0.4-0.5 MW/m 2 onto the mock-ups were realized before the heater fails. Then, irradiation was continued in a non-cycling mode.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hernández, F., E-mail: francisco.hernandez@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany); Kolb, M. [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-WPT) (Germany); Annabattula, R. [Indian Institute of Technology Madras (IITM), Department of Mechanical Engineering (India); Weth, A. von der [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR) (Germany)

    2014-10-15

    Highlights: • PREMUX has been constructed as preparation for a future out-of-pile thermo-mechanical qualification of a HCPB breeder unit mock-up. • The rationale and constructive details of PREMUX are reported in this paper. • PREMUX serves as a test rig for the new heater system developed for the HCPB-BU mock-up. • PREMUX will be used as benchmark for the thermal and thermo-mechanical models developed in ANSYS for the pebble beds of the HCPB-BU. • Preliminary results show the functionality of PREMUX and the good agreement of the measured temperatures with the thermal model developed in ANSYS. - Abstract: One of the European blanket designs for ITER is the Helium Cooled Pebble Bed (HCPB) blanket. The core of the HCPB-TBM consists of so-called breeder units (BUs), which encloses beryllium as neutron multiplier and lithium orthosilicate (Li{sub 4}SiO{sub 4}) as tritium breeder in form of pebble beds. After the design phase of the HCPB-BU, a non-nuclear thermal and thermo-mechanical qualification program for this device is running at the Karlsruhe Institute of Technology. Before the complex full scale BU testing, a pre-test mock-up experiment (PREMUX) has been constructed, which consists of a slice of the BU containing the Li{sub 4}SiO{sub 4} pebble bed. PREMUX is going to be operated under highly ITER-relevant conditions and has the following goals: (1) as a testing rig of new heater concept based on a matrix of wire heaters, (2) as benchmark for the existing finite element method (FEM) codes used for the thermo-mechanical assessment of the Li{sub 4}SiO{sub 4} pebble bed, and (3) in situ measurement of thermal conductivity of the Li{sub 4}SiO{sub 4} pebble bed during the tests. This paper describes the construction of PREMUX, its rationale and the experimental campaign planned with the device. Preliminary results testing the algorithm used for the temperature reconstruction of the pebble bed are reported and compared qualitatively with first analyses

  11. Using mockups for hands-on training

    International Nuclear Information System (INIS)

    Morris, A.R.

    1991-01-01

    The presentation of Using Mockups for Hands-on Training will be a slide presentation showing slides of mockups that are used by the Westinghouse Hanford Company in Maintenance Training activities. This presentation will compare mockups to actual plant equipment. It will explain the advantages and disadvantages of using mockups. The presentation will show students using the mockups in the classroom environment and slides of the actual plant equipment. The presentation will discuss performance-based training. This part of the presentation will show slides of students doing hands-on training on aerial lifts, fork trucks, and crane and rigging applications. Also shown are mockups that are used for basic hydraulics; hydraulic torquing; refrigeration and air conditioning; valve seat repair; safety relief valve training; and others. The presentation will discuss functional duplicate equipment and simulated nonfunctional equipment. The presentation will discuss the acquisition of mockups from spare parts inventory or from excess parts inventory. The presentation will show attendees how the mockups are used to enhance the training of the Hanford Site employees and how similar mockups could be used throughout the nuclear industry

  12. Breeding blanket for Demo

    International Nuclear Information System (INIS)

    Proust, E.; Giancarli, L.

    1992-01-01

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

  13. High 240Pu FTR/EMC experiments and analysis: Carbide fuel and UO2 blanket subassembly worths

    International Nuclear Information System (INIS)

    Ombrellaro, P.A.

    1977-06-01

    Carbide-plutonium fuel and UO 2 blanket subassembly worth measurements performed at ANL in the EMC/LWR were analyzed. Composition exchange worth calculations were performed for: (a) the replacement of high- 240 Pu fuel composition for low- 240 Pu fuel composition and carbide-plutonium fuel composition, successively, in the center subassembly of the core; (b) the replacement of low- 240 Pu fuel composition for carbide--plutonium fuel composition in one outer driver subassembly; and (c) the replacement of the radial reflector composition with UO 2 blanket composition in one subassembly of the radial reflector. The composition exchange worth calculations were performed in two-dimensional x,y geometry, using diffusion theory and perturbation theory. Each method produces about the same calculated-to-experimental bias factors

  14. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  15. Materials for breeding blankets

    International Nuclear Information System (INIS)

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

    1995-09-01

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

  16. Materials for breeding blankets

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. Criticality safety analysis for mockup facility

    International Nuclear Information System (INIS)

    Shin, Young Joon; Shin, Hee Sung; Kim, Ik Soo; Oh, Seung Chul; Ro, Seung Gy; Bae, Kang Mok

    2000-03-01

    Benchmark calculations for SCALE4.4 CSAS6 module have been performed for 31 UO 2 fuel, 15MOX fuel and 10 metal material criticality experiments and then calculation biases of the SCALE 4.4 CSAS6 module have been revealed to be 0.00982, 0.00579 and 0.02347, respectively. When CSAS6 is applied to the criticality safety analysis for the mockup facility in which several kinds of nuclear material components are included, the calculation bias of CSAS6 is conservatively taken to be 0.02347. With the aid of this benchmarked code system, criticality safety analyses for the mockup facility at normal and hypothetical accidental conditions have been carried out. It appears that the maximum K eff is 0.28356 well below than the critical limit, K eff =0.95 at normal condition. In a hypothetical accidental condition, the maximum K eff is found to be 0.73527 much lower than the subcritical limit. For another hypothetical accidental condition the nuclear material leaks out of container and spread or lump in the floor, it was assumed that the nuclear material is shaped into a slab and water exists in the empty space of the nuclear material. K eff has been calculated as function of slab thickness and the volume ratio of water to nuclear material. The result shows that the K eff increases as the water volume ratio increases. It is also revealed that the K eff reaches to the maximum value when water if filled in the empty space of nuclear material. The maximum K eff value is 0.93960 lower than the subcritical limit

  18. Proposal for the award of a blanket contract for automatic air-sampling systems for fire and gas detection in the LHC experiments

    CERN Document Server

    2004-01-01

    This document concerns the award of a blanket contract for automatic air-sampling systems for fire and gas detection in the LHC experiments. Following a market survey carried out among 119 firms in ten Member States, a call for tenders (IT-2891/ST) was sent on 1 August 2003 to four firms, in three Member States. By the closing date, CERN had received two tenders from one firm and one consortium, in three Member States. The Finance Committee is invited to agree to the negotiation of a blanket contract with ICARE (FR), the lowest bidder, for the supply of automatic air-sampling systems for fire and gas detection in the LHC experiments for a total amount not exceeding 1 750 000 euros (2 714 000 Swiss francs), subject to revision for inflation from 1 January 2007 with options for air-sampling smoke detection systems for electrical racks, for an additional amount of 400 000 euros (620 000 Swiss francs), subject to revision for inflation from 1 January 2007, bringing the total amount to a maximum of 2 150 000 euros...

  19. High Heat Flux Test of the ITER FW Semi-prototype Mockups

    International Nuclear Information System (INIS)

    Bae, Young Dug; Lee, Dong Won; Kim, Suk Kwon; Yoon, Jae Sung; Hong, Bong Guen

    2010-01-01

    As a procurement party of the ITER (International Thermonuclear Experimental Reactor) blanket project, we have to pass the qualification program which comprises two distinct phases; Phases 1 is about qualifying joining techniques, and phase 2 covers production scaled up component which is so called 'semi-prototype' and the enhanced heat flux qualification. We have constructed the FWQMs (First Wall Qualification Mockups) and already passed the qualification thermal fatigue tests organized by the IO (ITER Organization). The tests were performed at the EB-1200 in the USA and at the BESTH /JUDITH-2 in the EU. For the phase 2 qualification, the semi-prototype design will be provided by the IO. Even though the design has not fixed because the generic design of the first wall is not fixed, it shall comprise a minimum of 3 full length pairs of first wall fingers. We designed and fabricated two semi-prototype mockups which have a full length pair of first wall finger. The two mockups were designed to have different types of the cooling channels; one has a plain rectangular cross-section and the other has a hypervapotron type cooling channel. In order to compare the cooling capability of the two types, we performed high heat flux tests of the two mockups at the KoHLT-2

  20. Fabrication of a 1/6-scale mock-up and manifolds for the Korea first wall in the ITER

    International Nuclear Information System (INIS)

    Yoon, Jae Sung; Kim, Suk Kwon; Lee, Eo Hwak; Lee, Dong Won

    2012-01-01

    Korea has developed and participated in the Test Blanket Module (TBM) program of the International Thermo-nuclear Experimental Reactor (ITER). The first wall (FW) of the TBM is an important component that faces the plasma directly and therefore it is subjected to high heat and neutron loads. To fabricate the TBM FW, the Hot Isostatic Pressing (HIP) bonding method has been investigated. In the present study, the manufacturing method of the TBM FW is introduced through the fabrication and testing of a 1/6-scale mockup. To distribute fluid uniformly in the mock-up, a manifold was designed and fabricated using the ANSYS-CFX analysis. After the mock-up was fabricated and its fluid distribution tests performed, we compared the results of tests with the simulated results

  1. Measurement and control system for the ITER remote handling mock-up test

    International Nuclear Information System (INIS)

    Oka, K.; Kakudate, S.; Takiguchi, Y.; Ako, K.; Taguchi, K.; Tada, E.; Ozaki, F.; Shibanuma, K.

    1998-01-01

    The mock-up test platforms composed of full-scale remote handling (RH) equipment were developed for demonstrating remote replacement of the ITER blanket and divertor. In parallel, the measurement and control system for operating these RH equipment were constructed on the basis of open architecture with object oriented feature, aiming at realization of fully-remoted automatic operation required for ITER. This paper describes the design concept of the measurement and control system for the remote handling equipment of ITER, and outlines the measured performances of the fabricated measurement system for the remote handling mock-up tests, which includes Data Acquisition System (DAS), Visual Monitoring System (VMS) and Virtual Reality System (VRS). (authors)

  2. MIT LMFBR blanket physics project progress report No. 7, July 1, 1975--September 30, 1976

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1976-01-01

    Work during the period was devoted primarily to a range of analytical/numerical investigations, including evaluation of means to improve external blanket designs, beneficial attributes of the use of internal blankets, improved methods for the calculation of heterogeneous self-shielding and parametric studies of calculated spectral indices. Experimental work included measurements of the ratio of U-238 captures to U-235 fissions in a standard blanket mockup, and completion of development work on the radiophotoluminescent readout of LiF thermoluminescent detectors. The most significant findings were that there is very little prospect for substantial improvement in the breeding performance of external blankets, but internal blankets continue to show promise, particularly if they are used in such a way as to increase the volume fraction of fuel inside the core envelope. An improved equivalence theorem was developed which may allow use of fast reactor methods to calculate heterogeneously self-shielded cross sections in both fast and thermal reactors

  3. Modelling and computer simulation for the manufacture by powder HIPing of Blanket Shield components for ITER

    International Nuclear Information System (INIS)

    Gillia, O.; Bucci, Ph.; Vidotto, F.; Leibold, J.-M.; Boireau, B.; Boudot, C.; Cottin, A.; Lorenzetto, P.; Jacquinot, F.

    2006-01-01

    In components of blanket modules for ITER, intricate cooling networks are needed in order to evacuate all heat coming from the plasma. Hot Isostatic Pressing (HIPing) technology is a very convenient method to produce near net shape components with complex cooling network through massive stainless steel parts by bonding together tubes inserted in grooves machined in bulk stainless steel. Powder is often included in the process so as to release difficulties arising with gaps closure between tube and solid part or between several solid parts. In the mean time, it releases the machining precision needed on the parts to assemble before HIP. However, inserting powder in the assembly means densification, i.e. volume change of powder during the HIP cycle. This leads to global and local shape changes of HIPed parts. In order to control the deformations, modelling and computer simulation are used. This modelling and computer simulation work has been done in support to the fabrication of a shield prototype for the ITER blanket. Problems such as global bending of the whole part and deformations of tubes in their powder bed are addressed. It is important that the part does not bend too much. It is important as well to have circular tube shape after HIP, firstly in order to avoid their rupture during HIP but also because non destructive ultrasonic examination is needed to check the quality of the densification and bonding between tube and powder or solid parts; the insertions of a probe in the tubes requires a minimal circular tube shape. For simulation purposes, the behaviour of the different materials has to be modelled. Although the modelling of the massive stainless steel behaviour is not neglected, the most critical modelling is about power. For this study, a thorough investigation on the powder behaviour has been performed with some in-situ HIP dilatometry experiments and some interrupted HIP cycles on trial parts. These experiments have allowed the identification of a

  4. Strength of Mock-up Trial Grout

    DEFF Research Database (Denmark)

    Sørensen, Eigil V.

    The present report describes tests carried out on samples taken and cast during the execution of a mock-up trial placement of the high performance grout MASTERFLOW 9500 on January 21, 2009.......The present report describes tests carried out on samples taken and cast during the execution of a mock-up trial placement of the high performance grout MASTERFLOW 9500 on January 21, 2009....

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

    International Nuclear Information System (INIS)

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

    1995-08-01

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

  6. Mirror reactor blankets

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  7. Blanket testing in NET

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  8. Drilling Experiments of Dummy Fuel Rods Using a Mock-up Drilling Device and Detail Design of Device for Drilling of Irradiated Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Yong; Lee, H. K.; Chun, Y. B.; Park, S. J.; Kim, B. G

    2007-07-15

    KAERI are developing the safety evaluation method and the analysis technology for high burn-up nuclear fuel rod that is the project, re-irradiation for re-instrumented fuel rod. That project includes insertion of a thermocouple in the center hole of PWR nuclear fuel rod with standard burn-up, 3,500{approx}4,000MWD/tU and then inspection of the nuclear fuel rod's heat performance during re-irradiation. To re-fabricate fuel rod, two devices are needed such as a drilling machine and a welding machine. The drilling machine performs grinding a center hole, 2.5 mm in diameter and 50 mm in depth, for inserting a thermocouple. And the welding machine is used to fasten a end plug on a fuel rod. Because these two equipment handle irradiated fuel rods, they are operated in hot cell blocked radioactive rays. Before inserting any device into hot cell, many tests with that machine have to be conducted. This report shows preliminary experiments for drilling a center hole on dummy of fuel rods and optimized drilling parameters to lessen operation time and damage of diamond dills. And the design method of a drilling machine for irradiated nuclear fuel rods and detail design drawings are attached.

  9. Nuclear Analyses of Indian LLCB Test Blanket System in ITER

    Science.gov (United States)

    Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

    2017-04-01

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

  10. 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)

  11. Integration mockup and process material management system

    Science.gov (United States)

    Verble, Adas James, Jr.

    1992-01-01

    Work to define and develop a full scale Space Station Freedom (SSF) mockup with the flexibility to evolve into future designs, to validate techniques for maintenance and logistics and verify human task allocations and support trade studies is described. This work began in early 1985 and ended in August, 1991. The mockups are presently being used at MSFC in Building 4755 as a technology and design testbed, as well as for public display. Micro Craft also began work on the Process Material Management System (PMMS) under this contract. The PMMS simulator was a sealed enclosure for testing to identify liquids, gaseous, particulate samples, and specimen including, urine, waste water, condensate, hazardous gases, surrogate gasses, liquids, and solids. The SSF would require many trade studies to validate techniques for maintenance and logistics and verify system task allocations; it was necessary to develop a full scale mockup which would be representative of current SSF design with the ease of changing those designs as the SSF design evolved and changed. The tasks defined for Micro Craft were to provide the personnel, services, tools, and materials for the SSF mockup which would consist of four modules, nodes, interior components, and part task mockups of MSFC responsible engineering systems. This included the Engineering Control and Life Support Systems (ECLSS) testbed. For the initial study, the mockups were low fidelity, soft mockups of graphics art bottle, and other low cost materials, which evolved into higher fidelity mockups as the R&D design evolved, by modifying or rebuilding, an important cost saving factor in the design process. We designed, fabricated, and maintained the full size mockup shells and support stands. The shells consisted of cylinders, end cones, rings, longerons, docking ports, crew airlocks, and windows. The ECLSS required a heavier cylinder to support the ECLSS systems test program. Details of this activity will be covered. Support stands were

  12. Limitations on blanket performance

    International Nuclear Information System (INIS)

    Malang, S.

    1999-01-01

    The limitations on the performance of breeding blankets in a fusion power plant are evaluated. The breeding blankets will be key components of a plant and their limitations with regard to power density, thermal efficiency and lifetime could determine to a large degree the attractiveness of a power plant. The performance of two rather well known blanket concepts under development in the frame of the European Blanket Programme is assessed and their limitations are compared with more advanced (and more speculative) concepts. An important issue is the question of which material (structure, breeder, multiplier, coatings) will limit the performance and what improvement would be possible with a 'better' structural material. This evaluation is based on the premise that the performance of the power plant will be limited by the blankets (including first wall) and not by other components, e.g. divertors, or the plasma itself. However, the justness of this premise remains to be seen. It is shown that the different blanket concepts cover a large range of allowable power densities and achievable thermal efficiencies, and it is concluded that there is a high incentive to go for better performance in spite of possibly higher blanket cost. However, such high performance blankets are usually based on materials and technologies not yet developed and there is a rather high risk that the development could fail. Therefore, it is explained that a part of the development effort should be devoted to concepts where the materials and technologies are more or less in hand in order to ensure that blankets for a DEMO reactor can be developed and tested in a given time frame. (orig.)

  13. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

  14. Tritium breeding blanket

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  15. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

    A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

  16. Blankets for thermonuclear device

    International Nuclear Information System (INIS)

    Maki, Koichi; Fukumoto, Hideshi.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

    This report presents the results of a study of fusion breeder blanket technology. It reviews the role of the breeder blanket, the current understanding of the scientific and engineering bases of liquid metal and solid breeder blankets and the programs now underway internationally to resolve the uncertainities in current knowledge. In view of existing national expertise and experience, a solid breeder R and D program for Canada is recommended

  18. Destructive analysis on the ITER FW small scale mock-ups

    International Nuclear Information System (INIS)

    Wang, Pinghuai; Chen, Jiming; Liu, Danhua; Jin, Fanya; Yang, Bo

    2015-01-01

    As one of the core components of ITER, the first wall (FW) panel of shield blanket defines a physical boundary for the plasma transients and exhausts the majority of the plasma heat flux. China will undertake 12.64% of FW manufacturing tasks, and all of them are enhanced heat flux (EHF) components which will suffer surface heat flux of 4 - 5 MW/m 2 . The FW will be manufactured by a combination technology of explosion bonding CuCrZr alloy/316L (N) stainless steel plate and hot iso-static pressing (HIP) joining of beryllium tiles/CuCrZr alloy. The Be/Cu joint qualities is the key issue for the manufacturing of the FW panels. Several small scale mock-ups were manufactured for the qualification of the HIP technology for the FW. To avoid the brittle Be-Cu phase formed during the HIPing process, different thick Ti and pure Cu were coated on the beryllium tiles before HIPing to CuCrZr alloy. Ultrasonic testing was conducted on the mock-ups and destructive analysis was carried out on the mock-ups. For the failed ones, the results show that in the UT indication area brittle fracture occurs at the Be/Ti interface and then Ti/Cu interface in other areas. Based on these results, the manufacturing technology was improved mainly on the beryllium tiles quality, coating process and canister design. (author)

  19. ITER blanket designs

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  20. How tangible mock-ups support design collaboration

    DEFF Research Database (Denmark)

    Brandt, Eva

    2007-01-01

    This paper is a contribution to a more conscious use of tangible mock-ups in collaborative design processes. It describes a design team's use of mock-ups in a series of workshops involving potential customers and users. Focus is primarily on the use of three-dimensional design mock-ups and how...... differences in these affected the dialogue. Reflective conversations were established by using tangible mock-ups as 'things-to-think with'. They served as boundary objects that spanned the gap between the different competencies and interests of participants in design. The design mock-ups evoked different...... things for different participants whereas the challenge for the design team was to find boundaries upon which everybody could agree. The level of details represented in a mock-up affected the communication so that a mock-up with few details evoked different issues whereas a very detailed mock-up evoked...

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

    Science.gov (United States)

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

    2006-02-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  4. The making of a mock-up

    DEFF Research Database (Denmark)

    Rosenqvist, Tanja Schultz; Heimdal, Elisabeth Jacobsen

    2011-01-01

    As part of a research project about user involvement in textile design we have carried out two Design:Labs (Binder & Brandt 2008) engaging different stakeholders in designing textile products for Danish hospital environments. In this paper we follow a mock-up session done as part of the second...

  5. Fabrication and integrity test preparation of HIP-joined W and ferritic-martensitic steel mockups for fusion reactor development

    International Nuclear Information System (INIS)

    Lee, Dong Won; Shin, Kyu In; Kim, Suk Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Choi, Bo Guen; Moon, Se Youn; Hong, Bong Guen

    2014-01-01

    Tungsten (W) and ferritic-martensitic steel (FMS) as armor and structural materials, respectively, are the major candidates for plasma-facing components (PFCs) such as the blanket first wall (BFW) and the divertor, in a fusion reactor. In the present study, three W/FMS mockups were successfully fabricated using a hot isostatic pressing (HIP, 900 .deg. C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 .deg. C, 70 MPa, 2 hrs), and the W/FMS joining method was developed based on the ITER BFW and the test blanket module (TBM) development project from 2004 to the present. Using a 10-MHz-frequency flat-type probe to ultrasonically test of the joint, we found no defects in the fabricated mockups. For confirmation of the joint integrity, a high heat flux test will be performed up to the thermal lifetime of the mockup under the proper test conditions. These conditions were determined through a preliminary analysis with conventional codes such as ANSYS-CFX for thermal-hydraulic conditions considering the test facility, the Korea heat load test facility with an electron beam (KoHLT-EB), and its water coolant system at the Korea Atomic Energy Research Institute (KAERI)

  6. Fabrication and integrity test preparation of HIP-joined W and ferritic-martensitic steel mockups for fusion reactor development

    Science.gov (United States)

    Lee, Dong Won; Shin, Kyu In; Kim, Suk Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Choi, Bo Guen; Moon, Se Youn; Hong, Bong Guen

    2014-10-01

    Tungsten (W) and ferritic-martensitic steel (FMS) as armor and structural materials, respectively, are the major candidates for plasma-facing components (PFCs) such as the blanket first wall (BFW) and the divertor, in a fusion reactor. In the present study, three W/FMS mockups were successfully fabricated using a hot isostatic pressing (HIP, 900 °C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 °C, 70 MPa, 2 hrs), and the W/FMS joining method was developed based on the ITER BFW and the test blanket module (TBM) development project from 2004 to the present. Using a 10-MHz-frequency flat-type probe to ultrasonically test of the joint, we found no defects in the fabricated mockups. For confirmation of the joint integrity, a high heat flux test will be performed up to the thermal lifetime of the mockup under the proper test conditions. These conditions were determined through a preliminary analysis with conventional codes such as ANSYS-CFX for thermal-hydraulic conditions considering the test facility, the Korea heat load test facility with an electron beam (KoHLT-EB), and its water coolant system at the Korea Atomic Energy Research Institute (KAERI).

  7. Fabrication and integrity test preparation of HIP-joined W and ferritic-martensitic steel mockups for fusion reactor development

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Shin, Kyu In; Kim, Suk Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Choi, Bo Guen [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Moon, Se Youn; Hong, Bong Guen [Chonbuk National University, Jeonju (Korea, Republic of)

    2014-10-15

    Tungsten (W) and ferritic-martensitic steel (FMS) as armor and structural materials, respectively, are the major candidates for plasma-facing components (PFCs) such as the blanket first wall (BFW) and the divertor, in a fusion reactor. In the present study, three W/FMS mockups were successfully fabricated using a hot isostatic pressing (HIP, 900 .deg. C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 .deg. C, 70 MPa, 2 hrs), and the W/FMS joining method was developed based on the ITER BFW and the test blanket module (TBM) development project from 2004 to the present. Using a 10-MHz-frequency flat-type probe to ultrasonically test of the joint, we found no defects in the fabricated mockups. For confirmation of the joint integrity, a high heat flux test will be performed up to the thermal lifetime of the mockup under the proper test conditions. These conditions were determined through a preliminary analysis with conventional codes such as ANSYS-CFX for thermal-hydraulic conditions considering the test facility, the Korea heat load test facility with an electron beam (KoHLT-EB), and its water coolant system at the Korea Atomic Energy Research Institute (KAERI)

  8. The MARINE experiment: Irradiation of sphere-pac fuel and pellets of UO{sub 2−x} for americium breeding blanket concept

    Energy Technology Data Exchange (ETDEWEB)

    D' Agata, E., E-mail: elio.dagata@ec.europa.eu [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Hania, P.R. [Nuclear Research and Consultancy Group, P.O. Box 25, NL-1755 ZG Petten (Netherlands); Freis, D.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Bejaoui, S. [Commissariat à l’Energie Atomique et aux Energies Alternatives, DEN/DEC, F-13108 St. Paul lez Durance Cedex (France); Charpin, F.F.; Baas, P.J.; Okel, R.A.F.; Til, S. van [Nuclear Research and Consultancy Group, P.O. Box 25, NL-1755 ZG Petten (Netherlands); Lapetite, J.-M. [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Delage, F. [Commissariat à l’Energie Atomique et aux Energies Alternatives, DEN/DEC, F-13108 St. Paul lez Durance Cedex (France)

    2017-01-15

    Highlights: • MARINE is designed to check the behaviour of MABB sphere-pac concept. • MABB sphere-pac are compared with MABB pellet. • Swelling and helium release behaviour will be the main output of the experiment. • An experiment to check sphere-pac MADF fuel behaviour has been already performed. - Abstract: Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like {sup 241}Am is therefore an option for the reduction of radiotoxicity and heat production of waste packages to be stored in a repository. The MARINE irradiation experiment is the latest of a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS, SPHERE) performed in the High Flux Reactor (HFR). The MARINE experiment is developed and carried out in the framework of the collaborative research project PELGRIMM of the EURATOM 7th Framework Programme (FP7). During the past years of experimental works in the field of transmutation and tests of innovative nuclear fuels, the release or trapping of helium as well as swelling have been shown to be the key issues for the design of such kind of fuel both as drivers and even more for Am-bearing blanket targets (due to the higher Am contents). The main objective of the MARINE experiment is to study the in-pile behaviour of uranium oxide fuel containing 13% of americium and to compare the behaviour of sphere-pac versus pellet fuel, in particular the role of microstructure and temperature on fission gas release and He on fuel swelling. The MARINE experiment will be irradiated in 2016 in the HFR in Petten (The Netherlands) and is expected to be completed in spring 2017. This paper discusses the rationale and objective of the MARINE experiment and provides a general description of its design for which some innovative features have been adopted.

  9. Novel blanket design for ICTR's

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  10. TFTR lithium blanket module program. Final design report. Volume VII. LBM instrumentation and Program of Experiments and Analysis

    International Nuclear Information System (INIS)

    Harker, Y.D.; Tsang, F.Y.; Jassby, D.L.

    1983-07-01

    The Program of Experiments and Analysis comprises 3 spheres of activity: (1) measurements of neutron fluence and flux spectra inside and around the LBM, and of tritium production in the LBM central zone; (2) neutronic-code modeling and analysis of the TFTR/LBM system to predict the quantities measured in (1); (3) comparisons of the predicted and measured quantities, and improvements of the code modeling and analysis and the experimental techniques, in order to resolve any discrepancy between prediction and measurement. The measurement techniques are discussed. Section 5 of this volume discusses the strategy for carrying out the measurement program, for making comparisons with the neutronics code predictions, and for resolving discrepancies

  11. First Wall, Blanket, Shield Engineering Technology Program

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1982-01-01

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

  12. Magnetoconvection in HCLL blankets

    International Nuclear Information System (INIS)

    Mistrangelo, C.; Buehler, L.

    2014-01-01

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

  13. Neutronic design for the TFTR lithium blanket module

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  14. Progress on pebble bed experimental activity for the HE-FUS3 mock-ups

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Sansone, L.; Simoncini, M.; Zito, D.

    2002-01-01

    The EU Long Term for DEMO Programme foresees the qualification of the reference design of the helium cooled pebble bed (HCPB) - test blanket module (TBM) to be tested in ITER Reactor. In this frame, FZK and ENEA have launched many experimental activities for the evaluation of the interactions between the Tritium breeder and neutron multiplier pebble beds and the steel containment walls. Main aim of these activities is the measuring the pebble bed effective thermal conductivity, the wall heat transfer coefficient as well as their dependency from the mechanical constraints. The paper presents the progress of the testing activity and results of the tests on two mock-up, called Tazza and Helichetta, carried out on the HE-FUS3 facility at ENEA Brasimone. (orig.)

  15. IBL Thermal Mockup Bake-Out Tests

    CERN Document Server

    Nuiry, FX

    2014-01-01

    This note summarizes different bake-out tests that have been performed with the ATLAS Insertable B-Layer (IBL) mockup. Two beam pipe configurations have been tested: one with the aerogel insulation layer all along the pipe and one without insulation over 622 mm around Z0. These tests have been crucial for decisions about aerogel removal, choice of heaters for the LHC beam pipe bake-out, and choice of temperature setpoints for the cooling system during nominal IBL operation. They also revealed very useful information on integration issues and the thermo-mechanical behaviour of the IBL detector.

  16. High heat load tests on W/Cu mock-ups and evaluation of their application to EAST device

    Energy Technology Data Exchange (ETDEWEB)

    Li, H. [Institute of Plasma Physics, Chinese Academy of Science, Hefei, Anhui 230031 (China); Hefei Electronic Engineering Institute, Hefei, Anhui 230037 (China)], E-mail: lih72@hotmail.com; Chen, J.L.; Li, J.G. [Institute of Plasma Physics, Chinese Academy of Science, Hefei, Anhui 230031 (China); Sun, X.J. [Hefei Electronic Engineering Institute, Hefei, Anhui 230037 (China)

    2009-01-15

    Tungsten has been considered as the primary candidate plasma-facing materials (PFM) for the EAST device. Three actively cooled W/Cu mock-ups with an interlayer made of tungsten-copper alloy (1.5 mm) were designed and manufactured. The tungsten armors, pure sintered tungsten plate (1 mm) and plasma-sprayed tungsten coatings (0.3 and 0.9 mm), were bonded to the interlayer by brazing and depositing respectively. All mock-ups can withstand high heat flux up to 5 MW/m{sup 2} and no obvious failure was found after tests. The thermal performance experiments and microstructure analyses indicated the structure of mock-ups possess good thermal contact and high heat transfer capability. WCu alloy as an interlayer can largely reduce the stress due to the mismatch and improve the reliability. The mock-up with 0.9 mm coating had the highest surface temperature than the other two mock-ups, delaminations of this mock-up were found in the near surface by SEM. The primary results show that pure sintered tungsten brazed to WCu alloy is a possible way, and thick plasma-sprayed coating technique still need to be improved.

  17. In-pile thermocycling testing and post-test analysis of beryllium divertor mockups

    Energy Technology Data Exchange (ETDEWEB)

    Giniatulin, R.; Mazul, I. [Efremov Inst., St. Petersburg (Russian Federation); Melder, R.; Pokrovsky, A.; Sandakov, V.; Shiuchkin, A.

    1998-01-01

    The main damaging factors which impact the ITER divertor components are neutron irradiation, cyclic surface heat loads and hydrogen environment. One of the important questions in divertor mockups development is the reliability of beryllium/copper joints and the beryllium resistance under neutron irradiation and thermal cycling. This work presents the experiment, where neutron irradiation and thermocyclic heat loads were applied simultaneously for two beryllium/copper divertor mockups in a nuclear reactor channel to simulate divertor operational conditions. Two mockups with different beryllium grades were mounted facing each other with the tantalum heater placed between them. This device was installed in the active zone of the nuclear reactor SM-2 (Dimitrovgrad, Russia) and the tantalum block was heated by neutron irradiation up to a high temperature. The main part of the heat flux from the tantalum surface was transported to the beryllium surface through hydrogen, as a result the heat flux loaded two mockups simultaneously. The mockups were cooled by reactor water. The device was lowered to the active zone so as to obtain the heating regime and to provide cooling lifted. This experiment was performed under the following conditions: tantalum heater temperature - 1950degC; hydrogen environment -1000 Pa; surface heat flux density -3.2 MW/m{sup 2}; number of thermal cycles (lowering and lifting) -101; load time in each cycle - 200-5000 s; dwell time (no heat flux, no neutrons) - 300-2000 s; cooling water parameters: v - 1 m/s, Tin - 50degC, Pin - 5 MPa; neutron fluence -2.5 x 10{sup 20} cm{sup -2} ({approx}8 years of ITER divertor operation from the start up). The metallographic analysis was performed after experiment to investigate the beryllium and beryllium/copper joint structures, the results are presented in the paper. (author)

  18. ITER Blanket First Wall (WBS 1.6{sub 1}A)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; Kim, H. G.; Kim, J. H. (and others)

    2008-03-15

    International Thermonuclear Experimental Reactor (ITER) project is the international collaboration one for the commercialization of nuclear fusion energy through the technical and engineering verification. In ITER project, we plan to procure the blanket systems which has the risk of technology and cost when it is newly developed. We are developing the manufacturing process and joining technology for the ITER blanket to complete the procurement with qualified blanket system. To evaluate the soundness of manufacturing process, specimen and mock-up tests are being prepared. Finally, we can obtain the key technology of nuclear fusion reactor especially on the blanket design, joining and manufacturing technology through the present project and these technologies will help the construction of Korea fusion DEMO reactor and the development of commercial nuclear fusion reactor in Korea. In 1st year, through the fabrication of the Cu/SS and Be/Cu joint specimen, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The optimized HIP conditions (1050 .deg. C, 150 MPa, 2 hr for Cu/SS and 580 - 620 .deg. C, 100-150 MPa, 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint and NDT such as UT (10 MHz, 0.25 inch D, flat type) and ECT. Several mock-ups were fabricated for confirming the joint integrity and NDT. specimens fabricated with these mock-ups were used in mechanical tests including microstructure observation. The mock-ups were used in the HHF test after the developed NDT. In 2nd year, PHHT of Cu was investigated in order to recover its mechanical properties, and the pre-qualification mock-up were fabricated against the Qualification Program and sent to RF for HHF testing in TSEFEY. FW fabrication and joining procedure were documented in the form of the TSD. Qualification mock-up

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

    International Nuclear Information System (INIS)

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

    2006-07-01

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

  20. Development of a control system for a heavy object handling manipulator. Application to a remote maintenance system for ITER blanket module

    International Nuclear Information System (INIS)

    Yoshimi, Takashi; Tsuji, Kouichi; Miyagawa, Shinichi; Kubo, Tomomi; Kakudate, Satoshi; Tada, Eisuke

    2001-01-01

    This paper describes a control system for the heavy object handling manipulator. It has been developed for the blanket module remote maintenance system of ITER (International Thermonuclear Fusion Experimental Reactor). A rail-mounted vehicle-type manipulator is proposed for the precise handling of a blanket module which is about 4 tons in weight. Basically, this manipulator is controlled by teaching-playback technique. When grasping or releasing the module, the manipulator sags and the position of the end-effector changes about 50 [mm]. Applying only the usual teaching-playback control makes the smooth operation of setting/removing modules to/from the vacuum vessel wall difficult due to this position change. To solve this proper problem of heavy object handling manipulator, we have developed a system which uses motion patterns generated from two kinds of teaching points. These motion patterns for setting/removing heavy objects are generated by combining teaching points for positioning the manipulator with and without grasping the object. When these motion patterns are applied, the manipulator can transfer the object's weight smoothly at the setting/removing point. This developed system has been applied to the real-scale mock-up of the vehicle manipulator and through the actual module setting/removing experiments, we have verified its effectiveness and realized smooth maintenance operation. (author)

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

    International Nuclear Information System (INIS)

    Wu, Y.

    2007-01-01

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

  2. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

    Ozawa, Yoshihiro; Uda, Tatsuhiko; Maki, Koichi.

    1993-01-01

    The present invention provides a blanket of a thermonuclear device which produces tritium fuels consumed in plasmas while converting neutrons generated in the plasmas into heat energy. That is, zirconium is coated to at least one of neutron breeder pebbles and breeder pebbles, to suppress reaction between them by being in direct contact with each other at a high temperature. Further, fins are attached to a cooling pipe at a pitch smaller than the diameter of both of the pebbles, to prevent direct contact at whole surface of the pebbles and the cooling pipe, which would lower a temperature excessively. The length of the fin is controlled to control the thickness of a helium gas gap. With such constitution, direct contact of neutron breeder pebbles and the breeder pebble which are to be filled and mixed, and tend to react at a high temperature, can be prevented. The temperature of the breeding blanket is reliably prevented from lowering below a tritium emitting temperature. The structure is simplified and the production is facilitated. (I.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio, E-mail: enoeda.mikio@jaea.go.jp; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Yokoyama, Kenji

    2014-10-15

    Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li{sub 2}TiO{sub 3} pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  5. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    1983-10-01

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

  6. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, Masataka; Takeda, Nobukazu; Onozuka, Masanori [Japan Atomic Energy Agency (Japan); Kakudate, Satoshi [Mitsubishi Heavy Industries, Ltd. (Japan)

    2007-07-01

    The structure and fabrication methods of the ITER vacuum vessel have been investigated and defined by the ITER international team. However, some of the current specifications are very difficult to be achieved from the manufacturing point of view and will lead to cost increase. In the mock-up fabrication, it is planned to conduct the following items: 1. Feasibility of the Japanese proposed VV structure and fabrication methods and the applicability to the ITER are to be confirmed; 2. Assembly procedure and inspection procedure are to be confirmed; 3. Manufacturing tolerances are to be assessed; 4. Manufacturing schedule is to be assessed. This report summarizes the Japanese proposed specification of the VV mock-up describing differences between the ITER supplied design. General scope of the mock-up fabrication and the detailed dimensions are also shown. In the VV fabrication, several types of weld joint configuration will be used. This report shows the joint configurations proposed by Japan to be used for the inner shell connection, the rib-to-shell connection and outer shell connection, and the housing-to-shell connection, respectively. Non-destructive testing considered to be applied to each joint configuration is also presented. A series of the fabrication and assembly procedures for the mock-up are presented in this report, together with candidates of welding configurations. Finally, the report summarizes the results of mock-up fabrication, including results of nondestructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. (orig.)

  7. Liquid lithium blanket processing studies

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  8. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

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

  9. Finite element analysis and measurement for residual stress of dissimilar metal weld in pressurizer safety nozzle mockup

    International Nuclear Information System (INIS)

    Lee, Kyoung Soo; Kim, W.; Lee, Jeong Geun; Park, Chi Yong; Yang, Jun Seok; Kim, Tae Ryong; Park, Jai Hak

    2009-01-01

    Finite element (FE) analysis and experiment for weld residual stress (WRS) in the pressurizer safety nozzle mockup is described in various processes and results. Foremost of which is the dissimilar simulation metal welding (DMW) between carbon steel and austenitic stainless steel. Thermal and structural analyses were compared with actual residual stress, and actual measurements of. Magnitude and distribution of WRS in the nozzle mockup were assessed. Two measurement methods were used: hole-drilling method (HDM) with strain gauge for residual stress on the surface of the mockup, and block removal and splitting layer (BRSL) method for through-thickness. FE analysis and measurement data showed good agreement. In conclusion, the characteristics of weld residual stress of DMW could be well understood and the simplified FE analysis was verified as acceptable for estimating WRS

  10. Finite element analysis and measurement for residual stress of dissimilar metal weld in pressurizer safety nozzle mockup

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyoung Soo; Kim, W.; Lee, Jeong Geun; Park, Chi Yong; Yang, Jun Seok; Kim, Tae Ryong [Korea Electric Power Research Institute, Daejeon (Korea, Republic of); Park, Jai Hak [Chungbuk University, Cheongju (Korea, Republic of)

    2009-11-15

    Finite element (FE) analysis and experiment for weld residual stress (WRS) in the pressurizer safety nozzle mockup is described in various processes and results. Foremost of which is the dissimilar simulation metal welding (DMW) between carbon steel and austenitic stainless steel. Thermal and structural analyses were compared with actual residual stress, and actual measurements of. Magnitude and distribution of WRS in the nozzle mockup were assessed. Two measurement methods were used: hole-drilling method (HDM) with strain gauge for residual stress on the surface of the mockup, and block removal and splitting layer (BRSL) method for through-thickness. FE analysis and measurement data showed good agreement. In conclusion, the characteristics of weld residual stress of DMW could be well understood and the simplified FE analysis was verified as acceptable for estimating WRS

  11. Vacuum tests of a beamline front-end mock-up at the Advanced Photon Source

    International Nuclear Information System (INIS)

    Liu, C.; Nielsen, R.W.; Kruy, T.L.; Shu, D.; Kuzay, T.M.

    1994-01-01

    A-mock-up has been constructed to test the functioning and performance of the Advanced Photon Source (APS) front ends. The mock-up consists of all components of the APS insertion-device beamline front end with a differential pumping system. Primary vacuum tests have been performed and compared with finite element vacuum calculations. Pressure distribution measurements using controlled leaks demonstrate a better than four decades of pressure difference between the two ends of the mock-up. The measured pressure profiles are consistent with results of finite element analyses of the system. The safety-control systems are also being tested. A closing time of ∼20 ms for the photon shutter and ∼7 ms for the fast closing valve have been obtained. Experiments on vacuum protection systems indicate that the front end is well protected in case of a vacuum breach

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-01

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

  13. Fusion blanket design and optimization techniques

    International Nuclear Information System (INIS)

    Gohar, Y.

    2005-01-01

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

  14. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

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

    1986-02-01

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

  15. Thermal fatigue tests with actively cooled divertor mock-ups for ITER

    International Nuclear Information System (INIS)

    Roedig, M.; Duwe, R.; Linke, J.; Schuster, A.; Wiechers, B.; Ibbott, C.; Jacobson, D.; Le Marois, G.; Lind, A.; Lorenzetto, P.; Vieider, G.; Peacock, A.; Ploechl, L.; Severi, Y.; Visca, E.

    1998-01-01

    Mock-ups for high heat flux components with beryllium and CFC armour materials have been tested by means of the electron beam facility JUDITH. The experiments concerned screening tests to evaluate heat removal efficiency and thermal fatigue tests. CFC monoblocks attached to DS-Cu (Glidcop Al25) and CuCrZr tubes by active metal casting and Ti brazing showed the best thermal fatigue behaviour. They survived more than 1000 cycles at heat loads up to 25 MW m -2 without any indication of failure. Operational limits are given only by the surface temperature on the CFC tiles. Most of the beryllium mock-ups were of the flat tile type. Joining techniques were brazing, hot isostatic pressing (HIP) and diffusion bonding. HIPed and diffusion bonded Be/Cu modules have not yet reached the standards for application in high heat flux components. The limit of this production method is reached for heat loads of approximately 5 MW m -2 . Brazing with and without silver seems to be a more robust solution. A flat tile mock-up with CuMnSnCe braze was loaded at 5.4 MW m -2 for 1000 cycles without damage The first test with a beryllium monoblock joined to a CuCrZr tube by means of Incusil brazing shows promising results; it survived 1000 cycles at 4.5 MW m -2 without failure. (orig.)

  16. Fusion blanket inherent safety assessment

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  17. The blanket interface to TSTA

    International Nuclear Information System (INIS)

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

    1988-01-01

    The requirements of tritium technology are centered in three main areas, (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The Tritium Systems Test Assembly (TSTA) now in operation at Los Alamos National Laboratory (LANL) is dedicated to developing and demonstrating the tritium technology for fuel processing and containment. TSTA is the only fusion fuel processing facility that can operate in a continuous closed-loop mode. The tritium throughput of TSTA is 1000 g/d. However, TSTA does not have a blanket interface system. The authors have initiated a study to define a Breeder Blanket Interface (BBIO) for TSTA. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. Various methods of tritium recovery from liquid lithium were assessed: yttrium gettering, permeation windows, and molten salt extraction. The authors' evaluation concluded that the best method was molten salt extraction

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

    International Nuclear Information System (INIS)

    Werner, R.W.

    1978-01-01

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

  19. 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)

  20. Blanket safety by GEMSAFE methodology

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Saito, Masaki

    2001-01-01

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

  1. Mockup of an automated material transport system for remote handling

    International Nuclear Information System (INIS)

    Porter, M.L.

    1992-01-01

    The automated material transport system (AMTS) was conceived for the transport of samples within the material and process control laboratory (MPCL), located in the plutonium processing building of the special isotope separation (SIS) facility. The MPCL was designed with a dry sample handling laboratory and a wet chemistry analysis laboratory. Each laboratory contained several processing glove boxes. The function of the AMTS was to automate the handling of materials, multiple process samples, and bulky items between process stations with a minimum of operator intervention and with a minimum of waiting periods and nonproductive activities. The AMTS design requirements, design verification mockup plan, and AMTS mockup procurement specification were established prior to cancellation of the SIS project. Due to the AMTS's flexibility, the need for technology development, and applicability to other US Department of Energy facilities, mockup of the AMTS continued. This paper discusses the system design features, capabilities, and results of initial testing

  2. Mockup Didatic Set for Students Development in Automotive Electronic

    Directory of Open Access Journals (Sweden)

    Fabio Delatore

    2013-05-01

    Full Text Available The automotive engineering education area, specifically on internal combustion engine, requires the use of suitable systems, capable to simulate, test and obtain specifics data from its operation. Automotive engines are so complex due to it is a mix of engineering subjects, so, a mockup was created to help its study. The mockup is an exactly the same engine that equips a vehicle, but assembled in a mechanical base, equipped with all the necessary components for running it up. The objective of this work is to develop a mockup with a suitable Electronic Control Unit (ECU board, in order to obtain the sensors/actuators signals from the engine and control some important engine functions by using an external ECU, so that the students may test their own strategies, compare with the original ECU.

  3. Development of blanket remote maintenance system

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  4. Development of blanket remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

  5. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1986-06-01

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

  6. ABOUT DIGITAL MOCK-UP FOR MECHANICAL PRODUCTS

    Directory of Open Access Journals (Sweden)

    GHERGHINA George

    2015-06-01

    The digital mock-up of the product is built at a design stage, and is applicable to the whole life-cycle of the product, including design, manufacture, marketing and aftermarket. The digital mock-up could achieve interference check, motion analysis, simulation of performance and manufacturing, technical training, advertising and maintenance, planning etc. The DMU of mechanical products, as important engineering data in a company, is supposed to be able to support all the activities in the whole life-cycle of the product including design, manufacture, marketing and aftermarket

  7. Fabrication of small mock-ups reflecting the design features of the ITER semi-prototype

    International Nuclear Information System (INIS)

    Jung, Yang-Il; Choi, Byoung-Kwon; Park, Jeong-Yong; Kim, Suk-Kwon; Lee, Dong Won; Kim, Byoung Yoon

    2012-01-01

    The ITER semi-prototype was designed to qualify the manufacturing technology for the ITER blanket first wall. However, its fabrication is expected to face great difficulty due to a design complexity. Even though joining technology for different materials such as beryllium, CuCrZr, and stainless steel (SS) was developed during the first stage of qualification, the joining is still a key issue for the fabrication of the semi-prototype. In this study, small mock-ups (SMU) were fabricated to realize and verify the manufacturing of the semi-prototype reflecting the described design features. The joining of multiple beryllium tiles on the angled CuCrZr surface was confirmed with SMU no. 1. Six beryllium tiles were joined using hot isostatic pressing (HIP), and slitting was then performed to form multiple tiles. In SMU no. 2, HIP was performed two times in order to facilitate the cooling channels at the CuCrZr/SS interface, and to join the beryllium tiles on CuCrZr/SS. The method used to form a pressure boundary for the complex cooling channels was also developed by fabricating the SMU no. 3. The SMUs confirmed the applicability of the HIP for the manufacturing of the semi-prototype.

  8. The supply of small scale mock-ups of the primary wall module concepts for ITER

    International Nuclear Information System (INIS)

    Walsh, G.; Cheyne, K.; Lorenzetto, P.

    1998-01-01

    The present design of Blanket Shield and Primary Wall for ITER envisages construction of the wall with a water cooled, stainless steel outer layer and a water cooled, copper liner on the inside plasma facing surface. Protection of the inner copper surface with an armour layer is necessary to cope with plasma to wall interaction. There are a number of armour materials under consideration, for this project beryllium was used. The scope of work was to produce a series of mock-ups, each consisting of a different combination of materials, which included Dispersion Strengthened Copper, Copper-Chrome-Zirconium alloy, Beryllium and Stainless Steel. Hot Isostatic Pressing (HIP) was the method used to ensure that a fully diffused bonded joint was achieved giving the necessary strength and thermal conductivity. The first five of the mock ups have been successfully completed and are being tested at the various laboratories in Europe. The remaining mock ups are awaiting the results of this test work prior to being completed. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-15

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

  10. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

    Rocaboy, Alain.

    1982-06-01

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

  11. Thermal-Fatigue Analysis of W-joined Ferritic-Martensitic Steel Mockup for Fusion Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Kim, Suk Kwon; Park, Seong Dae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shin, Kyu In [Gentec Co., Daejeon (Korea, Republic of); Moon, Se Yeon; Hong, Bong Guen [Chonbuk National University, Chonbuk (Korea, Republic of)

    2015-10-15

    Through the ITER blanket first wall (BFW) development project in Korea, the joining methods were developed with a beryllium (Be) layer as a plasma-facing material, a copper alloy (CuCrZr) layer as a heat sink, and type 316L austenitic stainless steel (SS316L) as a structural material. And joining methods were developed such as Be as an armor and FMS as a structural material, or W as an armor and FMS as a structural material were developed through the test blanket module (TBM) program. As a candidate of PFC for DEMO, W/FMS joining methods have been developed and a new Ti interlayer was applied differently from the previous work. In the present study, the W/FMS PFC development was introduced with the following procedure to apply to the PFCs for a fusion reactor: (1) Three W/FMS mockups were fabricated using the developed HIP followed by a post-HIP heat treatment (PHHT). (2) Because the High Heat Flux (HHF) test should be performed over the thermal lifetime of the mockup under the proper test conditions to confirm the joint's integrity, the test conditions were determined through a preliminary analysis. In this study, commercial ANSYS-CFX for thermalhydraulic analysis and ANSYS-mechanical for the thermo-mechanical analysis are used to evaluate the thermal-lifetime of the mockup to determine the test conditions. Also, the Korea Heat Load Test facility with an Electron Beam (KoHLT-EB) will be used and its water cooling system is considered to perform the thermal-hydraulic analysis especially for considering the two-phase analysis with a higher heat flux conditions. From the analysis, the heating and the cooling conditions were determined for 0.5- and 1.0-MW/m{sup 2} heat fluxes, respectively. Elastic-plastic analysis is performed to determine the lifetime and finally, the 1.0 MW/m{sup 2} heat flux conditions are determined up to 4,306 cycles. The test will be done in the near future and the measured temperatures will be compared with the present simulation results.

  12. Status of fusion reactor blanket evaluation studies in France

    International Nuclear Information System (INIS)

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

    1985-03-01

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

  13. Thermal-Fatigue Analysis of W-coated Ferritic-Martensitic Steel Mockup for Fusion Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Kim, Suk Kwon; Park, Seong Dae; Kim, Dong Jun [KAERI, Daejeon (Korea, Republic of); Moon, Se Yeon; Hong, Bong Guen [Chonbuk University, Jeonju (Korea, Republic of)

    2016-05-15

    In this study, commercial ANSYS-CFX for thermalhydraulic analysis and ANSYS-mechanical for the thermo-mechanical analysis are used to evaluate the thermal-lifetime of the mockup to determine the test conditions. Also, the Korea Heat Load Test facility with an Electron Beam (KoHLT-EB) will be used and its water cooling system is considered to perform the thermal-hydraulic analysis especially for considering the two-phase analysis with a higher heat flux conditions. Through the ITER blanket first wall (BFW) development project in Korea, the joining methods were developed with a beryllium (Be) layer as a plasma-facing material, a copper alloy (CuCrZr) layer as a heat sink, and type 316L austenitic stainless steel (SS316L) as a structural material. And joining methods were developed such as Be as an armor and FMS as a structural material, or W as an armor and FMS as a structural material were developed through the test blanket module (TBM) program. As a candidate of PFC for DEMO, a new W/FMS joining methods, W coating with plasma torch, have been developed. The HHF test conditions are found by performing a thermal-hydraulic and thermo-mechanical analysis with the conventional codes such as ANSYSCFX and .mechanical especially for considering the two-phase condition in cooling tube.

  14. Disruption problematics in segmented blanket concepts

    International Nuclear Information System (INIS)

    Crutzen, Y.; Fantechi, S.; Farfaletti-Casali, F.

    1994-01-01

    In Tokamaks, the hostile operating environment originated by plasma disruption events requires that the first wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence there is a need to improve the safety features of the blanket design concepts satisfying the disruption problematics and to formulate guidelines on the required internal reinforcements of the blanket components. The present paper describes the recent investigations on blanket reinforcement systems needed in order to optimize the first-wall/blanket/shield structural design for next step and commercial fusion reactors in the context of ITER, DEMO and SEAFP activities

  15. Tritium inventory in Li2ZrO3 blanket

    International Nuclear Information System (INIS)

    Nishikawa, M.; Baba, A.

    1998-01-01

    Recently, we have presented the way to estimate the tritium inventory in a solid breeder blanket considering effects of diffusion of tritium in the grain, absorption of water in the bulk of grain, and adsorption of water on the surface of grain, together with two types of isotope exchange reactions. It is reported in our previous paper that the estimated tritium inventory for a LiAlO 2 blanket agrees well with data observed in various in situ experiments when the effective diffusivity of tritium from the EXOTIC-6 experiment is used and that the better agreement is obtained when existence of some water vapor is assumed in the purge gas. The same way as used for a LiAlO 2 blanket is applied to a Li 2 ZrO 3 blanket in this study and the estimated tritium inventory shows a good agreement with data obtained in such in situ experiments as MOZART, EXOTIC-6 and TRINE experiments. (orig.)

  16. Siloette, Siloe mock-up; Siloette, modele nucleaire de siloe

    Energy Technology Data Exchange (ETDEWEB)

    Delcroix, V; Jeanne, G; Mitault, G; Schulhof, P [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-07-01

    Siloette is the Siloe mock-up. The main installations are described: various tanks, building, auxiliaries, control systems... Precis ions are given about precautions taken for using spent fuel elements. (authors) [French] Siloette est le modele nucleaire de SILOE. On decrit ses diverses installations: bassins, batiments, auxiliaires, controle... Des precisions sont donnees sur les precautions prises pour y utiliser des elements uses. (auteurs)

  17. 305 Building K basin mockup facility functions and requirements

    International Nuclear Information System (INIS)

    Steele, R.M.

    1994-01-01

    This document develops functions and requirements for installation and operation of a cold mockup test facility within the 305 Building. The test facility will emulate a portion of a typical spent nuclear fuel storage basin (e.g., 105-KE Basin) to support evaluation of equipment and processes for safe storage and disposition of the spent nuclear fuel currently within the K Basins

  18. Demonstration of the LHC Safety Training Tunnel Mock-Up

    CERN Multimedia

    Brice, Maximilien

    2014-01-01

    Members of CERN's management visit the LHC tunnel mock-up at the Safety Training Centre on the Prévessin site. The facility is used to train personnel in emergency responses including the use of masks and safe evacuation.

  19. FBR core mock-up RAPSODIE I - experimental analysis

    International Nuclear Information System (INIS)

    Brochard, D.; Buland, P.; Gantenbein, F.

    1990-01-01

    The main phenomena which influence the LMFBR core response to a seismic excitation are the fluid structure interaction and the impacts between subassemblies. To study the core behaviour, seismic tests have been performed on the core mock-up RAPSODIE with and without fluid and restraint ring and for different levels of excitation. This paper summarizes the results of these tests. (author)

  20. Final Report: Characterization of Canister Mockup Weld Residual Stresses

    International Nuclear Information System (INIS)

    Enos, David; Bryan, Charles R.

    2016-01-01

    Stress corrosion cracking (SCC) of interim storage containers has been indicated as a high priority data gap by the Department of Energy (DOE) (Hanson et al., 2012), the Electric Power Research Institute (EPRI, 2011), the Nuclear Waste Technical Review Board (NWTRB, 2010a), and the Nuclear Regulatory Commission (NRC, 2012a, 2012b). Uncertainties exist in terms of the environmental conditions that prevail on the surface of the storage containers, the stress state within the container walls associated both with weldments as well as within the base metal itself, and the electrochemical properties of the storage containers themselves. The goal of the work described in this document is to determine the stress states that exists at various locations within a typical storage canister by evaluating the properties of a full-diameter cylindrical mockup of an interim storage canister. This mockup has been produced using the same manufacturing procedures as the majority of the fielded spent nuclear fuel interim storage canisters. This document describes the design and procurement of the mockup and the characterization of the stress state associated with various portions of the container. It also describes the cutting of the mockup into sections for further analyses, and a discussion of the potential impact of the results from the stress characterization effort.

  1. Astronaut Curtis Brown on flight deck mockup during training

    Science.gov (United States)

    1994-01-01

    Astronaut Curtis L. Brown, STS-66 pilot, mans the pilot's station during a rehearsal of procedures to be followed during the launch and entry phases of their scheduled November 1994 flight. This rehearsal, held in the crew compartment trainer (CCT) of JSC's Shuttle mockup and integration laboratory, was followed by a training session on emergency egress procedures.

  2. Final Report: Characterization of Canister Mockup Weld Residual Stresses

    Energy Technology Data Exchange (ETDEWEB)

    Enos, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-12-01

    Stress corrosion cracking (SCC) of interim storage containers has been indicated as a high priority data gap by the Department of Energy (DOE) (Hanson et al., 2012), the Electric Power Research Institute (EPRI, 2011), the Nuclear Waste Technical Review Board (NWTRB, 2010a), and the Nuclear Regulatory Commission (NRC, 2012a, 2012b). Uncertainties exist in terms of the environmental conditions that prevail on the surface of the storage containers, the stress state within the container walls associated both with weldments as well as within the base metal itself, and the electrochemical properties of the storage containers themselves. The goal of the work described in this document is to determine the stress states that exists at various locations within a typical storage canister by evaluating the properties of a full-diameter cylindrical mockup of an interim storage canister. This mockup has been produced using the same manufacturing procedures as the majority of the fielded spent nuclear fuel interim storage canisters. This document describes the design and procurement of the mockup and the characterization of the stress state associated with various portions of the container. It also describes the cutting of the mockup into sections for further analyses, and a discussion of the potential impact of the results from the stress characterization effort.

  3. Fusion reactor blanket-main design aspects

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  4. Beryllium research on FFHR molten salt blanket

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  5. Status of the Digital Mock-up System for the dismantling of the nuclear facilities

    International Nuclear Information System (INIS)

    Park, Hee Seoung; Kim, S. K.; Lee, K. W.; Oh, W. J.

    2004-12-01

    The database system have already developed is impossible to solve a quantitative evaluation about a various situation from the dismantle activities of the reactor had contaminated with radioactivity. To satisfy the requirements for safety and economical efficiency among a major decommissioning technologies, it need a system that can evaluate and estimate dismantling scheduling, amount of radioactive waste being dismantled, and decommissioning cost. We have review and analyzed status of the digital mock-up system to get a technical guide because we have no experience establishment of one relation to dismantling of research reactor and nuclear power plant

  6. Evaluation of compost blankets for erosion control from disturbed lands.

    Science.gov (United States)

    Bhattarai, Rabin; Kalita, Prasanta K; Yatsu, Shotaro; Howard, Heidi R; Svendsen, Niels G

    2011-03-01

    Soil erosion due to water and wind results in the loss of valuable top soil and causes land degradation and environmental quality problems. Site specific best management practices (BMP) are needed to curb erosion and sediment control and in turn, increase productivity of lands and sustain environmental quality. The aim of this study was to investigate the effectiveness of three different types of biodegradable erosion control blankets- fine compost, mulch, and 50-50 mixture of compost and mulch, for soil erosion control under field and laboratory-scale experiments. Quantitative analysis was conducted by comparing the sediment load in the runoff collected from sloped and tilled plots in the field and in the laboratory with the erosion control blankets. The field plots had an average slope of 3.5% and experiments were conducted under natural rainfall conditions, while the laboratory experiments were conducted at 4, 8 and 16% slopes under simulated rainfall conditions. Results obtained from the field experiments indicated that the 50-50 mixture of compost and mulch provides the best erosion control measures as compared to using either the compost or the mulch blanket alone. Laboratory results under simulated rains indicated that both mulch cover and the 50-50 mixture of mulch and compost cover provided better erosion control measures compared to using the compost alone. Although these results indicate that the 50-50 mixtures and the mulch in laboratory experiments are the best measures among the three erosion control blankets, all three types of blankets provide very effective erosion control measures from bare-soil surface. Results of this study can be used in controlling erosion and sediment from disturbed lands with compost mulch application. Testing different mixture ratios and types of mulch and composts, and their efficiencies in retaining various soil nutrients may provide more quantitative data for developing erosion control plans. Copyright © 2010 Elsevier

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

    International Nuclear Information System (INIS)

    1979-11-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  9. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Directory of Open Access Journals (Sweden)

    Košťál Michal

    2016-01-01

    Full Text Available A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1–10 MeV and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1. Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  10. Fabrication of the full scale separable first wall of ITER shielding blanket

    International Nuclear Information System (INIS)

    Kosaku, Yasuo; Kuroda, Toshimasa; Hatano, Toshihisa; Enoeda, Mikio; Miki, Nobuharu; Akiba, Masato

    2002-10-01

    Shielding blanket for ITER-FEAT applies the unique first wall structure which is separable from the shield block for the purpose of radio-active waste reduction in the maintenance work and cost reduction in fabrication process. Also, it is required to have various types of slots in both of the first wall and the shield block, to reduce the eddy current for reduction of electro-magnetic force in disruption events. Such unique features of blanket structure required technological clarification from the technical base of the previous achievement of the blanket module fabrication development. Previously, within the EDA Task T216+, a prototype for the no.4 Primary Wall Module of the ITER Shield Blanket with integrated first wall has been manufactured by forging and drilling and the first wall has been manufactured and joined to the shield block by Hot Isostatic Pressing (HIP) in one step process. This work has been performed to clarify the remaining R and D issues which have not been covered in the previous R and D. This report summarizes the demonstrative fabrication of the real scale separable first wall for ITER shielding blanket designed for ITER-FEAT, together with the essential technology developments such as, the slit grooving of the first wall with beryllium armor and SS shield block and fabrication of a partial mockup of beryllium armored first wall panel with built-in cooling channels. This work has been performed under the task agreement of G 16 TT 95 FJ (T420-1) in ITER Engineering Design Activity Extension Period. By the demonstration of the Be armor joining to the first wall panel, the joining technique of Be and DSCu developed previously, was shown to be applicable to the realistic structure of first wall panel. Also, the slit grooving by an end-mill method and an electron discharge machining method have been applied to the first wall mockup with Be armor tiles and demonstrated the applicability within the design tolerance. As the slit grooving technique

  11. WASTE PACKAGE CORROSION STUDIES USING SMALL MOCKUP EXPERIMENTS

    International Nuclear Information System (INIS)

    B.E. Anderson; K.B. Helean; C.R. Bryan; P.V. Brady; R.C. Ewing

    2005-01-01

    The corrosion of spent nuclear fuel and subsequent mobilization of radionuclides is of great concern in a geologic repository, particularly if conditions are oxidizing. Corroding A516 steel may offset these transport processes within the proposed waste packages at the Yucca Mountain Repository (YMR) by retaining radionuclides, creating locally reducing conditions, and reducing porosity. Ferrous iron, Fe 2+ , has been shown to reduce UO 2 2+ to UO 2(s) [1], and some ferrous iron-bearing ion-exchange materials adsorb radionuclides and heavy metals [2]. Of particular interest is magnetite, a potential corrosion product that has been shown to remove TcO 4 - from solution [3]. Furthermore, if Fe 2+ minerals, rather than fully oxidized minerals such as goethite, are produced during corrosion, then locally reducing conditions may be present. High electron availability leads to the reduction and subsequent immobilization of problematic dissolved species such as TcO 4 - , NpO 2 + , and UO 2 2+ and can also inhibit corrosion of spent nuclear fuel. Finally, because the molar volume of iron material increases during corrosion due to oxygen and water incorporation, pore space may be significantly reduced over long time periods. The more water is occluded, the bulkier the corrosion products, and the less porosity is available for water and radionuclide transport. The focus of this paper is on the nature of Yucca Mountain waste package steel corrosion products and their effects on local redox state, radionuclide transport, and porosity

  12. Water Mock-up for the Sodium Waste Treatment Process

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Ho Yun; Kim, Jong Man; Kim, Byung Ho; Lee, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    It is important to safely treat the waste sodium which was produced from the sodium cooled fast reactors and the sodium facilities. About 1.3 tons of sodium waste has accumulated at KAERI from the sodium experiments which have been carried out since 1990. Also, large scaled sodium experiments are scheduled to verify the design of the sodium cooled fast reactor. As a treatment method for the waste sodium produced at the sodium facility, an investigation of the reaction procedure of the waste sodium with the sodium hydroxide aqueous has been developed. The NOAH process was developed in France for the treatment of waste sodium produced from sodium facilities and reactors. In the NOAH process, a small amount of sodium waste is continuously injected into the upper space which is formed on the free surface of the aqueous and slowly reacted with sodium hydroxide aqueous. Since the density of the sodium is lower than that of the aqueous, the injected sodium waste sometimes accumulates above the free surface of the sodium hydroxide aqueous, and its reaction rate becomes slow or suddenly increases. In the improved process, the sodium was injected into a reaction vessel filled with a sodium hydroxide aqueous through an atomizing nozzle installed on a lower level than that of the aqueous to maintain the reaction uniformly. Fig.1 shows the sodium waste process which was proposed in KAERI. The aqueous is composed of 60% sodium hydroxide, and its temperature is about 60 .deg. C. The process is an exothermic reaction. The hydrogen gas is generated, and the concentration of the sodium hydroxide increases in this process. It needs several systems for the process, i.e. a waste sodium injection, a cooling of the aqueous, hydrogen ventilation, and neutralization with nitric acid. The atomizing nozzle was designed to inject the sodium with the nitrogen gas which supplies a heat to the sodium to prevent its solidification and to uniformly mix the sodium with the aqueous. There are

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-01

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

  15. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    1983-10-01

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

  16. Design requirement on HYPER blanket fuel assembly

    International Nuclear Information System (INIS)

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

    2000-07-01

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

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

    International Nuclear Information System (INIS)

    Song, P.M.

    1990-01-01

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

  18. A new magnet for the LHC mock-up

    CERN Multimedia

    HSE Unit

    2013-01-01

    This year, the safety training centre on the Prévessin site acquired a mock-up of the LHC, which simulates the work and safety conditions in the tunnel.   Photo: Christoph Balle. A new dummy quadrupole has just been added to the magnet chain, making the mock-up even more realistic. The new facility, which was a joint endeavour by the TE, GS, BE and EN Departments, will significantly improve the quality of the various training courses held at the centre, particularly the course on the use of self-rescue masks. To consult the safety training catalogue and/or sign up for radiation protection training, please go to: https://cta.cern.ch. For further information, please contact the Safety Training and Awareness service by telephone on 73811 or 79935 or by e-mail to safety-training@cern.ch.  

  19. Liquid metal cooled blanket concept for NET

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  20. Fusion blankets for high efficiency power cycles

    International Nuclear Information System (INIS)

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

    1980-04-01

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

  1. Analysis of high heat flux testing of mock-ups

    International Nuclear Information System (INIS)

    Salavy, J.-F.; Giancarli, L.; Merola, M.; Picard, F.; Roedig, M.

    2003-01-01

    ITER EU Home Team is performing a large R and D effort in support of the development of high heat flux components for ITER. In this framework, this paper describes the thermal analyses, the fatigue lifetime evaluation and the transient VDE with material melting related to the high heat flux thermo-mechanical tests performed in the JUDITH facility. It reports on several mock-ups representative of different proposed component designs based on Be, W and CFC as armour materials

  2. HYBRID REPRESENTATION OF DIGITAL MOCKUP FOR HERITAGE BUILDINGS MANAGEMENT

    Directory of Open Access Journals (Sweden)

    G. Nicolas

    2013-07-01

    Full Text Available This article deals with the implementation of tool allowing the portability of the digital mock-up for architectural projects on the building renovation place and the use of representation layers giving functions adapted to the different workers open to work in this place. Our test case is applied to renovation works on old windows in an ancient abbey where it is necessary to improve the thermal efficiency.

  3. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

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

    1987-10-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  5. On the conditions of existence of cold-blanket systems

    International Nuclear Information System (INIS)

    Lehnert, B.

    1977-12-01

    An extende analysis of the partially ionized boundary layer of a magnetized plasma has been performed, leading to the following results: (i) In a first approximation the ion density at the inner ''edge'' of the layer becomes related to the wall-near neutral gas density, in a way being independent of the spatial distribution of the ionization rate. (ii) The particle and momentum balance equations, and the associated impermeability condition of the plasma with respect to neutral gas penetration, are not sufficient to specify a cold-blanket state, but have to be combined with considerations of the heat blance. This leads to lower and upper power input limits, thus defining conditions for the existence of a cold-blanket state. At decreasing beta values , or increasing radiation losses, there are situations where such a state cannot exist at all. (iii) It should become possible to fulfill the cold-blanket conditions in full-scale reactors as well as in certain model experiments. Probably these conditions can also be satisfied in large tokamaks like JET, and by fast gas injection in devices such as Alcator, but not in medium-size tokamaks being operated at moderately high ion densities. (iv) A strong ''boundary layer stabilization'' mechanism due to the joint viscosity-resistivity-pressure effects is available under cold-blanket conditions. (author)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  7. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    International Nuclear Information System (INIS)

    Nakahira, M.; Takeda, N.; Kakudate, S.; Onozuka, M.

    2008-01-01

    The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV

  8. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, M. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)], E-mail: nakahira.masataka@jaea.go.jp; Takeda, N.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Onozuka, M. [Mitsubishi Nuclear Energy Systems, Inc., 1700K Street NW, Suite 440, Washington, DC 20006 (United States)

    2008-12-15

    The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV.

  9. Thermal transient and the temperature profile in a HELICA mock-up simulated by a new finite element homogenous model

    International Nuclear Information System (INIS)

    Zaccari, Nicola; Aquaro, Donato

    2013-01-01

    Highlights: • We have developed a numerical model of the pebble beds is based on the results of a theoretical and experimental research activity performed. • The model has been used to simulate the experimental tests performed on HELICA mock-up (ENEA Italy). • Moreover the numerical results are compared with the experimental ones. Finally, a discussion on results obtained by other authors involved in the benchmark is reported. -- Abstract: This paper deals with a numerical approach for simulating the thermal and mechanical behaviour of pebble beds used as breeder and neutron multiplier in breeding blanket of nuclear fusion reactor. The model of the pebble beds is based on the results of a theoretical and experimental research activity performed by the Authors on ceramic pebble beds (lithium ortosilicate and lithium metatitanate). The results of this activity permitted to determine the effective thermal conductivity of the beds, versus the temperature and the axial pressure and to implement a homogenous model of pebble bed in a FEM code. This paper illustrates an application of the implemented model, considering pebble beds under several cycles of heating and cooling. The examined geometry corresponds to the HELICA mock-up tested by ENEA in the research centre Brasimone. The experimental tests performed on HELICA have been used as a benchmark problem in order to assess the different approaches for simulating pebble beds. In this paper, the simulations performed with two-dimensional models are illustrated. Moreover the numerical results are compared with the experimental ones. Finally, a discussion on results obtained by other authors involved in the benchmark is reported

  10. Preparation of W/CuCrZr monoblock test mock-up using vacuum brazing technique

    International Nuclear Information System (INIS)

    Singh, Kongkham Premjit; Khirwadkar, Samir S.; Bhope, Kedar; Patel, Nikunj; Mokaria, Prakash K.; Mehta, Mayur

    2015-01-01

    Development of the joining for W/CuCrZr monoblock PFC test mock-up is an interest area in Fusion R and D. W/Cu bimetallic material has prepared using OFHC copper casting approach on the radial surface of W monoblock tile surface. The W/Cu bimetallic material has been joined with CuCrZr tube (heat sink) material with the vacuum brazing route. Vacuum brazing of W/Cu-CuCrZr has been performed @ 970 °C for 10 mins using NiCuMn-37 filler material under deep vacuum environment (10 -6 mbar). Graphite fixtures were used for OFHC copper casting and vacuum brazing experiments. The joint integrity of W/Cu-CuCrZr monoblock mock-up on W/Cu and Cu-CuCrZr has been checked using ultrasonic immersion technique. Micro-structural examination and Spot-wise elemental analysis have been carried out using HR-SEM and EDAX. The results of the experimental work will be discussed in the paper. (author)

  11. Tests and measurements with a thermal VXD mock-up for BELLE II

    International Nuclear Information System (INIS)

    Huebner, Lars

    2015-03-01

    As part of the Belle detector upgrade, located at the KEK in Tsukuba, Japan, a CO 2 cooling system will be added. Using new detector components, which are easily damageable or influenced by heat, make this step necessary. Particularly the next to the beam pipe located PXD is strained by high thermal load and therefore requires cooling. The CDC needs a constant temperature for precise measurements, but it could be influenced by heat from the SVD. Knowledge about the heat generation and distribution is needed before assembling the full detector. A mock-up of the innermost parts of the detector and a CO 2 cooling system is under construction at DESY in Hamburg, Germany, to gather such knowledge. The mock-up should be able to emulate the thermal properties of the final detector. Within the scope of this bachelor's thesis, the outermost VXD Layer 6 was studied in a flat arrangement. Focus lay on the heat dissipation at the sensors and on pressure drop measurements of the cooling pipe. It was investigated whether the applied heat load can be sufficiently lead away and how large the pressure drop is along the experiment line. Despite cooling was applied, a remarkable rise in temperature was observed. However, the unfavorable position of the thermistors make reliable quantitative statements of the sensor dummies' temperatures impossible. The pressure drop was determined, but is of limited accuracy due to large uncertainties. Further investigations have to be made with a better set-up.

  12. (D,T) Driven thorium hybrid blankets

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  13. 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.)

  14. Methods to enhance blanket power density

    International Nuclear Information System (INIS)

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

    1982-06-01

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

  15. 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)

  16. Manufacturing pre-qualification of a Short Breeder Unit mockup (SHOBU) as part of the roadmap toward the out-of-pile validation of a full scale Helium Cooled Pebble Bed Breeder Unit

    International Nuclear Information System (INIS)

    Hernández, Francisco A.; Rey, Jorg; Neuberger, Heiko; Krasnorutskyi, Sergii; Niewöhner, Reinhard; Felde, Alexander

    2015-01-01

    Highlights: • A relevant mockup of a HCPB Breeder Unit for ITER (SHOBU) has been manufactured. • The manufacturing technologies used in SHOBU and its assembly sequence are reported. • Preliminary qualification of the welds has been successfully done after codes. • Future work foreseen to manufacture a feasibility mockup according to RCC-MRx code. - Abstract: The key components of the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM) in ITER are the Breeder Units (BU). These are the responsible for the tritium breeding and part of the heat extraction in the HCPB TBM. After a detailed design and engineering phase performed during the last years in the Karlsruhe Institute of Technology (KIT), a reference model for the manufacturing of a HCPB BU mock-up has been obtained. The mid-term is the out-of-pile qualification of the thermal and thermo-mechanical performance of a full-scale HCPB BU mock-up in a dedicated helium loop. Several key manufacturing technologies have been developed for the fabrication of the HCPB BU. In order to pre-qualify these techniques, a Short Breeder Unit mock-up (SHOBU) is under construction and to be tested. This paper aims at describing the relevance of SHOBU with a full-scale HCPB BU, the constitutive parts of SHOBU, the manufacturing and joining technologies involved, the assembly sequence (taking into consideration functional steps like its filling with Li_4SiO_4 pebbles or its assembly in the HCPB TBM) and the welding procedures studied. The paper concludes with a description of the required pre-qualification tests performed to SHOBU, i.e. pressure and leak tightness tests, according to the standards.

  17. NET test blanket design and remote maintenance

    International Nuclear Information System (INIS)

    Holloway, C.; Hubert, P.

    1991-01-01

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

  18. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

    2009-01-01

    Selecting relevant features is in demand when a large data set is of interest in a classification task. It produces a tractable number of features that are sufficient and possibly improve the classification performance. This paper studies a statistical method of Markov blanket induction algorithm...... for filtering features and then applies a classifier using the Markov blanket predictors. The Markov blanket contains a minimal subset of relevant features that yields optimal classification performance. We experimentally demonstrate the improved performance of several classifiers using a Markov blanket...... induction as a feature selection method. In addition, we point out an important assumption behind the Markov blanket induction algorithm and show its effect on the classification performance....

  19. Blanket materials for DT fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1981-01-01

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

  20. Interactive virtual mock-ups for Remote Handling compatibility assessment of heavy components

    Energy Technology Data Exchange (ETDEWEB)

    Oosterhout, J. van, E-mail: j.vanoosterhout@differ.nl [FOM Institute DIFFER (Dutch Institute for Fundamental Energy Research), Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Heemskerk, C.J.M.; Koning, J.F. [Heemskerk Innovative Technology B.V., Jonckerweg 12, 2201 DZ Noordwijk 6 (Netherlands); Ronden, D.M.S.; Baar, M. de [FOM Institute DIFFER (Dutch Institute for Fundamental Energy Research), Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands)

    2014-10-15

    Highlights: •Specific ITER components require RHCA on hardware mock-ups. •Hardware mock-ups are expensive and have a long lead time. •Interactive Virtual Reality mock-ups are readily available and easily adapted. •This paper analysis and proposes improvements to simulator capabilities. -- Abstract: ITER standards Tesini (2009) require hardware mock-ups to validate the Remote Handling (RH) compatibility of RH class 1- and critical class 2-components. Full-scale mock-ups of large ITER components are expensive, have a long lead time and lose their relevance in case of design changes. Interactive Virtual Reality simulations with real time rigid body dynamics and contact interaction allow for RH Compatibility Assessment during the design iterations. This paper explores the use of interactive virtual mock-ups to analyze the RH compatibility of heavy component handling and maintenance. It infers generic maintenance operations from the analysis and proposes improvements to the simulator capabilities.

  1. Interactive virtual mock-ups for Remote Handling compatibility assessment of heavy components

    International Nuclear Information System (INIS)

    Oosterhout, J. van; Heemskerk, C.J.M.; Koning, J.F.; Ronden, D.M.S.; Baar, M. de

    2014-01-01

    Highlights: •Specific ITER components require RHCA on hardware mock-ups. •Hardware mock-ups are expensive and have a long lead time. •Interactive Virtual Reality mock-ups are readily available and easily adapted. •This paper analysis and proposes improvements to simulator capabilities. -- Abstract: ITER standards Tesini (2009) require hardware mock-ups to validate the Remote Handling (RH) compatibility of RH class 1- and critical class 2-components. Full-scale mock-ups of large ITER components are expensive, have a long lead time and lose their relevance in case of design changes. Interactive Virtual Reality simulations with real time rigid body dynamics and contact interaction allow for RH Compatibility Assessment during the design iterations. This paper explores the use of interactive virtual mock-ups to analyze the RH compatibility of heavy component handling and maintenance. It infers generic maintenance operations from the analysis and proposes improvements to the simulator capabilities

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  3. Beryllium armored mockups for fusion high flux application

    International Nuclear Information System (INIS)

    Gervash, A.; Giniyatulin, R.; Mazul, I.; Watson, R.

    1998-01-01

    One of the main requirements to use Be as a candidate for plasma facing component in ITER is providing a reliable joint between Be and Cu-alloy heat sink structure. In this work authors present the results of recent activity on this way. To create Be/CuCrZr joints the unique fast e-beam brazing technology was developed in Russia. The numbers of Be/CuCrZr mock-ups were manufactured in Efremov Institute by fast e-beam brazing using Cu-Sn-In-Ni brazing alloy. These mock-ups were tested by Sandia Laboratory at the EBTS electron beam facility. The goals of the tests were to define the allowable dimensions of the armour tiles for the heat loads of more than 10 MW/m 2 , to find the limit of bond strength for the Be/CuCrZr joint and response to heat loads and to estimate the life time of the brazed tiles by thermo-cyclic testing. The screening and thermal fatigue results are presented. With the aim to check the applicability of developed fast brazing process to DS-Cu alloy (GlidCop) the actively cooled mock-up produced by fast brazing Be onto GlidCop structure was manufactured and tested in Efremov Institute, main results are discussed. Summarizing testing data authors show that the CuCrZr or DS-Cu heat sink armored with beryllium can survive high heat fluxes (≥10 MW/m 2 ) during thousand heating/cooling cycles without serious damaging in the armour material and its joint. (author)

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

    International Nuclear Information System (INIS)

    Zeile, Christian; Maione, Ivan A.

    2015-01-01

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

  5. Convertible shielding to ceramic breeding blanket

    International Nuclear Information System (INIS)

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

    1995-05-01

    Four concepts have been studied for the ITER convertible blanket: 1)Layered concept 2)BIT(Breeder-Inside-Tube)concept 3)BOT(Breeder-Out of-Tube)concept 4)BOT/mixed concept. All concepts use ceramic breeder and beryllium neutron multiplier, both in the shape of small spherical pebbles, 316SS structure, and H 2 O coolant (inlet/outlet temperatures : 100/150degC, pressure : 2 MPa). During the BPP, only beryllium pebbles (the primary pebble in case of BOT/mixed concept) are filled in the blanket for shielding purpose. Then, before the EPP operation, breeder pebbles will be additionally inserted into the blanket. Among possible conversion methods, wet method by liquid flow seems expecting for high and homogeneous pebble packing. Preliminary 1-D neutronics calculation shows that the BOT/mixed concept has the highest breeding and shielding performance. However, final selection should be done by R and D's and more detail investigation on blanket characteristics and fabricability. Required R and D's are also listed. With these efforts, the convertible blanket can be developed. However, the following should be noted. Though many of above R and D's are also necessary even for non-convertible blanket, R and D's on convertibility will be one of the most difficult parts and need significant efforts. Besides the installation of convertible blanket with required structures and lines for conversion will make the ITER basic machine more complicated. (author)

  6. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

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

    1993-11-01

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

  7. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

    Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

    1999-07-01

    The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m 2 ) the average neutron power load on the first wall is below 1 MWm .2 , which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

  8. Tritium transport analysis for CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-15

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

  9. ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Clikeman, F.M.

    1982-07-01

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

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

    International Nuclear Information System (INIS)

    Tanigawa, Hisashi; Enoeda, Mikio

    2010-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-15

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

  13. ITER breeding blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

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

  14. Digital mock-up for the spent fuel disassembly processes

    International Nuclear Information System (INIS)

    Lee, J. Y.; Kim, S. H.; Song, T. G.; Kim, Y. H.; Hong, D. H.; Yoon, J. S.

    2000-12-01

    In this study, the graphical design system is developed and the digital mock-up is implemented for designing the spent fuel handling and disassembly processes. The system consists of a 3D graphical modeling system, a devices assembling system, and a motion simulation system. This system is used throughout the design stages from the conceptual design to the motion analysis. By using this system, all the process involved in the spent fuel handling and disassembly processes are analyzed and optimized. Also, this system is used in developing the on-line graphic simulator which synchronously simulates the motion of the equipment in a real time basis by connecting the device controllers with the graphic server through the TCP/IP network. This simulator can be effectively used for detecting the malfunctions of the process equipment which is remotely operated. Thus, the simulator enhances the reliability and safety of the spent fuel handling process by providing the remote monitoring function of the process. The graphical design system and the digital mock-up system can be effectively used for designing the process equipment, as well as the optimized process and maintenance process. And the on-line graphic simulator can be an alternative of the conventional process monitoring system which is a hardware based system

  15. Beryllium R and D for blanket application

    Energy Technology Data Exchange (ETDEWEB)

    Dalle Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Longhurst, G.R. [Idaho National Engineering Lab., Idaho Falls (United States); Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-10-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.) 29 refs.

  16. Beryllium R and D for blanket application

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Scaffidi-Argentina, F.; Kawamura, H.

    1998-01-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.)

  17. Beryllium R&D for blanket application

    Science.gov (United States)

    Donne, M. Dalle; Longhurst, G. R.; Kawamura, H.; Scaffidi-Argentina, F.

    1998-10-01

    The paper describes the main problems and the R&D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point.

  18. Mechanical and thermal design of hybrid blankets

    International Nuclear Information System (INIS)

    Schultz, K.R.

    1978-01-01

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

  19. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  20. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  1. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  2. Space environment durability of beta cloth in LDEF thermal blankets

    Science.gov (United States)

    Linton, Roger C.; Whitaker, Ann F.; Finckenor, Miria M.

    1993-01-01

    Beta cloth performance for use on long-term space vehicles such as Space Station Freedom (S.S. Freedom) requires resistance to the degrading effects of the space environment. The major issues are retention of thermal insulating properties through maintaining optical properties, preserving mechanical integrity, and generating minimal particulates for contamination-sensitive spacecraft surfaces and payloads. The longest in-flight test of beta cloth's durability was on the Long Duration Exposure Facility (LDEF), where it was exposed to the space environment for 68 months. The LDEF contained 57 experiments which further defined the space environment and its effects on spacecraft materials. It was deployed into low-Earth orbit (LEO) in Apr. 1984 and retrieved Jan. 1990 by the space shuttle. Among the 10,000 plus material constituents and samples onboard were thermal control blankets of multilayer insulation with a beta cloth outer cover and Velcro attachments. These blankets were exposed to hard vacuum, thermal cycling, charged particles, meteoroid/debris impacts, ultraviolet (UV) radiation, and atomic oxygen (AO). Of these space environmental exposure elements, AO appears to have had the greatest effect on the beta cloth. The beta cloth analyzed in this report came from the MSFC Experiment S1005 (Transverse Flat-Plate Heat Pipe) tray oriented approximately 22 deg from the leading edge vector of the LDEF satellite. The location of the tray on LDEF and the placement of the beta cloth thermal blankets are shown. The specific space environment exposure conditions for this material are listed.

  3. Simulation in full-scale mock-ups: an ergonomics evaluation method?

    DEFF Research Database (Denmark)

    Andersen, Simone Nyholm; Broberg, Ole

    2014-01-01

    This paper presents and exploratory study of four simulation sessions in full-scale mock-ups of future hospital facilities.......This paper presents and exploratory study of four simulation sessions in full-scale mock-ups of future hospital facilities....

  4. Review: BNL Tokamak graphite blanket design concepts

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  5. Blanket design for imploding liner systems

    International Nuclear Information System (INIS)

    Schaffer, M. J.

    1980-01-01

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

  6. APT target-blanket fabrication development

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, D.L.

    1997-06-13

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

  7. Stress analysis of the tokamak engineering test breeder blanket

    International Nuclear Information System (INIS)

    Huang Zhongqi

    1992-01-01

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

  8. The fusion blanket program at Chalk River

    International Nuclear Information System (INIS)

    Hastings, I.J.

    1986-03-01

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

  9. Characterization of flaws in a tube bundle mock-up for reliability studies

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Bakhtiari, S.

    1997-01-01

    As part of an assessment of in-service inspection of steam generator tubes, the authors will assemble a steam generator mock-up for round robin studies and use as a test bed in evaluating emerging technologies. Progress is reported on the characterization of flaws that will be part of the mock-up. Eddy current and ultrasonic techniques are being evaluated as a means to characterize the flaws in the mock-up tubes before final assembly. Twenty Inconel 600 tubes with laboratory-grown cracks, typical of those to be used in the mock-up, were provided by Pacific Northwest National Laboratory for laboratory testing. After the tubes were inspected with eddy current and ultrasonic techniques, they were destructively analyzed to establish the actual depths, lengths, and profiles of the cracks. The analysis of the results will allow the best techniques to be used for characterizing the flaws in the mock-up tubes

  10. Characterization of flaws in a tube bundle mock-up for reliability studies

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Bakhtiari, S.

    1996-10-01

    As part of an assessment of in-service inspection of steam generator tubes, the authors will assemble a steam generator mock-up for round robin studies and use as a test bed in evaluating emerging technologies. Progress is reported on the characterization of flaws that will be part of the mock-up. Eddy current and ultrasonic techniques are being evaluated as a means to characterize the flaws in the mock-up tubes before final assembly. Twenty Inconel 600 tubes with laboratory-grown cracks, typical of those to be used in the mock-up, were provided by Pacific Northwest National Laboratory for laboratory testing. After the tubes were inspected with eddy current and ultrasonic techniques, they were destructively analyzed to establish the actual depths, lengths, and profiles of the cracks. The analysis of the results will allow the best techniques to be used for characterizing the flaws in the mock-up tubes

  11. Advanced high performance solid wall blanket concepts

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  12. Workshop on cold-blanket research

    International Nuclear Information System (INIS)

    1977-05-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  15. Investigation of aqueous slurries as fusion reactor blankets

    International Nuclear Information System (INIS)

    Schuller, M.J.

    1985-01-01

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

  16. Measurements relevant to simulating subcriticality in ADS facilities with blanket

    International Nuclear Information System (INIS)

    Titarenko, Yu. E.; Batyaev, V.F.; Borovlev, S.P.; Gladkikh, N.G.; Igumnov, M.M.; Legostaev, V.O.; Karpikhin, E.I.; Konev, V.N.; Kushnerev, Yu.T.; Popkov, V.N.; Ryazhsky, V.I.; Spiridonov, V.G.; Chernyavsky, E.V.; Shvedov, O.V.

    2009-10-01

    The work presents the results of determining the blanket subcriticality for a zero-power heavy water reactor MAKET at the Institute for Theoretical and Experimental Physics, Moscow. The blanket is hexagonal lattice made of 36 90%-enriched 235U fuel rods spaced 173mm apart. The subcriticality was varied from ∼0.3% to 5% by adjusting the heavy water level. The subcriticality values were calibrated using the dependence of reactivity on heavy water level. The pulsed neutron source technique was used to measure the temporal dependence of neutron field at different blanket points for the calibrated subcriticality values. The subciticality values obtained in terms of the 'inverse clock' formulae using the decay constants of the measured dependences proved to differ from the calibrated subcriticalities by not more than 7% at the average. The MCNP code-aided simulations of the experiment made has given the calibrated keff values at prescribed heavy water levels and led to the neutron field decay constants at given points, which differ on the average from their experimental values by not more than 7% too. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  19. ITER-FEAT vacuum vessel and blanket design features and implications for the R and D programme

    International Nuclear Information System (INIS)

    Ioki, K.; Cardella, A.; Elio, F.; Onozuka, M.; Daenner, W.; Koizumi, K.; Krylov, V.

    2001-01-01

    A tight fitting configuration of the VV to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the TF ripple. The blanket modules are supported directly by the VV. A full-scale VV sector model has provided critical information related to fabrication technology, and the magnitude of welding distortions and achievable tolerances. This R and D validated the fundamental feasibility of the double-wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and the robustness of solid HIP joining was demonstrated in R and D, by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal. (author)

  20. ITER-FEAT vacuum vessel and blanket design features and implications for the R and D programme

    International Nuclear Information System (INIS)

    Ioki, K.; Cardella, A.; Elio, F.; Onozuka, M.; Daenner, W.; Koizumi, K.; Krylov, V.A.

    2001-01-01

    A configuration in which the vacuum vessel (VV) fits tightly to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the toroidal field ripple. The blanket modules are supported directly by the VV. A full scale VV sector model has provided critical information related to fabrication technology and for testing the magnitude of welding distortions and achievable tolerances. This R and D validated the fundamental feasibility of the double wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and robustness of solid hot isostatic pressing joining were demonstrated in the R and D by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal. (author)

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

    International Nuclear Information System (INIS)

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

    1997-07-01

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

  2. ITER-FEAT vacuum vessel and blanket design features and implications for the R&D programme

    Science.gov (United States)

    Ioki, K.; Dänner, W.; Koizumi, K.; Krylov, V. A.; Cardella, A.; Elio, F.; Onozuka, M.; ITER Joint Central Team; ITER Home Teams

    2001-03-01

    A configuration in which the vacuum vessel (VV) fits tightly to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the toroidal field ripple. The blanket modules are supported directly by the VV. A full scale VV sector model has provided critical information related to fabrication technology and for testing the magnitude of welding distortions and achievable tolerances. This R&D validated the fundamental feasibility of the double wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and robustness of solid hot isostatic pressing joining were demonstrated in the R&D by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal.

  3. Characterization of plasma sprayed beryllium ITER first wall mockups

    Energy Technology Data Exchange (ETDEWEB)

    Castro, R.G.; Vaidya, R.U.; Hollis, K.J. [Los Alamos National Lab., NM (United States). Material Science and Technology Div.

    1998-01-01

    ITER first wall beryllium mockups, which were fabricated by vacuum plasma spraying the beryllium armor, have survived 3000 thermal fatigue cycles at 1 MW/m{sup 2} without damage during high heat flux testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. The thermal and mechanical properties of the plasma sprayed beryllium armor have been characterized. Results are reported on the chemical composition of the beryllium armor in the as-deposited condition, the through thickness and normal to the through thickness thermal conductivity and thermal expansion, the four-point bend flexure strength and edge-notch fracture toughness of the beryllium armor, the bond strength between the beryllium armor and the underlying heat sink material, and ultrasonic C-scans of the Be/heat sink interface. (author)

  4. Characterization of Plasma Sprayed Beryllium ITER First Wall Mockups

    International Nuclear Information System (INIS)

    Castro, Richard G.; Vaidya, Rajendra U.; Hollis, Kendall J.

    1997-10-01

    ITER first wall beryllium mockups, which were fabricated by vacuum plasma spraying the beryllium armor, have survived 3000 thermal fatigue cycles at 1 MW/sq m without damage during high heat flux testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. The thermal and mechanical properties of the plasma sprayed beryllium armor have been characterized. Results are reported on the chemical composition of the beryllium armor in the as-deposited condition, the through thickness and normal to the through thickness thermal conductivity and thermal expansion, the four-point bend flexure strength and edge-notch fracture toughness of the beryllium armor, the bond strength between the beryllium armor and the underlying heat sink material, and ultrasonic C-scans of the Be/heat sink interface

  5. Project W-314 performance mock-up test procedure

    International Nuclear Information System (INIS)

    CARRATT, R.T.

    1999-01-01

    The purpose of this Procedure is to assist construction in the pre-operational fabrication and testing of the pit leak detection system and the low point drain assembly by: (1) Control system testing of the pit leak detection system will be accomplished by actuating control switches and verifying that the control signal is initiated, liquid testing and overall operational requirements stated in HNF-SD-W314-PDS-003, ''Project Development Specification for Pit Leak Detection''. (2) Testing of the low point floor drain assembly by opening and closing the drain to and from the ''retracted'' and ''sealed'' positions. Successful operation of this drain will be to verify that the seal does not leak on the ''sealed'' position, the assembly holds liquid until the leak detector actuates and the assembly will operate from on top of the mock-up cover block

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  7. A constitutive model for the thermo-mechanical behaviour of fusion-relevant pebble beds and its application to the simulation of HELICA mock-up experimental results

    International Nuclear Information System (INIS)

    Vella, G.; Maio, P.A. Di; Giammusso, R.; Tincani, A.; Orco, G. Dell

    2006-01-01

    Within the framework of the activities promoted by European Fusion Development Agreement on the technology of the Helium Cooled Pebble Bed Test Blanket Module to be irradiated in one of the ITER equatorial ports, attention has been focused on the theoretical modelling of the thermo-mechanical constitutive behaviour of both beryllium and lithiated ceramics pebble beds, that are envisaged to act respectively as neutron multiplier and tritium breeder. The thermo-mechanical behaviour of the pebble beds and their nuclear performances in terms of tritium production depend on the reactor relevant conditions (heat flux and neutron wall load), the pebble sizes and the breeder cell geometries (bed thickness, pebble packing factor, bed overall thermal conductivity). ENEA-Brasimone and the Department of Nuclear Engineering (DIN) of the Palermo University have performed intense research activities intended to investigate fusion-relevant pebble bed thermo-mechanical behaviour by adopting both experimental and theoretical approaches. In particular, ENEA has carried out several experimental campaigns on small scale mock-ups tested in out-of-pile conditions, while DIN has developed a proper constitutive model that has been implemented on commercial FEM code, for the prediction of the thermal and mechanical performances of fusion-relevant pebble beds and for the comparison with the experimental results of the ENEA tests. In that framework, HELICA mock-up has been set-up and tested to investigate the behaviour of pebble bed in reactor-relevant geometries, providing useful data sets to be numerically reproduced by means of the DIN constitutive model, contributing to its assessment. The paper presents the constitutive model developed and the main experimental results of two test campaigns on HELICA mock-up carried out at HE-FUS 3 facility of ENEA Brasimone, the geometry of the mock-up, the adopted thermal and mechanical boundary conditions and the test operating conditions. The most

  8. Fabrication data package for HEDL dosimetry in the ORNL Poolside Facility: LWR Pressure Vessel Mock-up irradiation

    International Nuclear Information System (INIS)

    Lippincott, E.P.; McElroy, W.N.; Kellogg, L.S.; Gold, R.; Guthrie, G.L.; Ruddy, F.H.; Ulseth, J.A.

    1981-09-01

    This document provides a complete description of the HEDL dosimetry inserted in the metallurgical specimen irradiation in the LWR Pressure Vessel Mock-up at the Oak Ridge Reactor Poolside Facility (PSF). This experiment is being conducted under the Nuclear Regulatory Commission sponsored program on Surveillance Dosimetry Improvement. The irradiation started April 1980 with recovery of the 2 x 10 19 (nominal fluence with E > 1 MeV) capsule in September 1980, the 4 x 10 19 surveillance capsule in November 1981 and the pressure vessel and void box capaules about August 1982

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  11. Recent High Heat Flux Tests on W-Rod-Armored Mockups

    International Nuclear Information System (INIS)

    Nygren, Richard E.; Youchison, Dennis L.; McDonald, Jimmie M.; Lutz, Thomas J.; Miszkiel, Mark E.

    2000-01-01

    In the authors initial high heat flux tests on small mockups armored with W rods, done in the small electron beam facility (EBTS) at Sandia National Laboratories, the mockups exhibited excellent thermal performance. However, to reach high heat fluxes, they reduced the heated area to only a portion (approximately25%) of the sample. They have now begun tests in their larger electron beam facility, EB 1200, where the available power (1.2 MW) is more than enough to heat the entire surface area of the small mockups. The initial results indicate that, at a given power, the surface temperatures of rods in the EB 1200 tests is somewhat higher than was observed in the EBTS tests. Also, it appears that one mockup (PW-10) has higher surface temperatures than other mockups with similar height (10mm) W rods, and that the previously reported values of absorbed heat flux on this mockup were too high. In the tests in EB 1200 of a second mockup, PW-4, absorbed heat fluxes of approximately22MW/m 2 were reached but the corresponding surface temperatures were somewhat higher than in EBTS. A further conclusion is that the simple 1-D model initially used in evaluating some of the results from the EBTS testing was not adequate, and 3-D thermal modeling will be needed to interpret the results

  12. Fast critical experiments in FCA and their analysis

    International Nuclear Information System (INIS)

    Hirota, Jitsuya

    1984-02-01

    JAERI Fast Critical Facility FCA went critical for the first time in April, 1967. Since then, critical experiments and their analysis were carried out on thirty-five assemblies until march, 1982. This report summarizes many achievements obtained in these fifteen years and points out disagreements observed between the calculation and experiment for further studies. A series of mock-up experiments for Experimental Fast Reactor JOYO, a theoretical and numerical study of adjustment of group constants by using integral data and a development of proton-recoil counter system for fast neutron spectrum measurement won high praise. Studies of Doppler effect of structural materials, effect of fission product accumulation on sodium-void worth, axially heterogeneous core and actinide cross sections attracted world-side attention. Significant contributions were also made to Prototype Fast Breeder Reactor MONJU through the partial mock-up experiments. Disagreements between the calculation and experiment were observed in the following items; reaction rate distribution and reactivity worth of B 4 C absorber in radial blanket, central reactivity worth in core with reflector, plate/pin fuel heterogeneity effect on criticality, sodium-void effect in central core region, Doppler effect of structural materials, core neutron spectrum near large resonances of iron and oxygen, effect of fission product accumulation on sodium-void worth, physics property of heterogeneous core, reactivity change resulted from fuel slumping and so on. Further efforts should be made to solve these disagreements through recalculating the experimental results with newly developed data and methods and carrying out the experiments intended to identify the cause of disagreement. (author)

  13. Test Blanket Working Group's recent activities

    International Nuclear Information System (INIS)

    Vetter, J.E.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Ofek, R.

    1986-05-01

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

  16. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1994-11-01

    The BOT (Breeder Outside Tube) Solid Breeder Blanket for a fusion DEMO reactor is presented. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. In the paper the reference blanket design and external loops are described as well as the results of the theoretical and experimental work in the fields of neutronics, thermohydraulics, mechanical stresses, tritium control and extraction, development and irradiation of the ceramic breeder material, beryllium development, ferromagnetic forces caused by disruptions, safety and reliability. An outlook is given on the remaining open questions and on the required R and D program. (orig.) [de

  17. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

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

    1987-11-01

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

  18. Thermal and mechanical design of WITAMIR-I blanket

    International Nuclear Information System (INIS)

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

    1980-10-01

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

  19. A study on the enhancement of the reliability in gravure offset roll printing with blanket swelling control

    International Nuclear Information System (INIS)

    Kim, Ga Eul; Woo, Kyoohee; Kang, Dongwoo; Jang, Yunseok; Lee, Taik-Min; Kwon, Sin; Choi, Young-Man; Lee, Moon G

    2016-01-01

    In roll-offset printing (patterning) technology with a PDMS blanket as a transfer medium, one of the major reliability issues is the occurrence of swelling, which involves absorption of the ink solvent in the printing blanket with repeated printing. This study developed a method to resolve blanket swelling in gravure offset roll printing and performed experiments for performance verification. The physical phenomena of mass and heat transfer were applied to fabricate a device based on convection drying. The proposed device managed to effectively control blanket swelling through drying by blowing air and additional temperature control. The experiments verified that printing quality (in particular the variation of the width of printed patterns) was maintained over 500 continuous printing. (paper)

  20. Epoxy blanket protects milled part during explosive forming

    Science.gov (United States)

    1966-01-01

    Epoxy blanket protects chemically milled or machined sections of large, complex structural parts during explosive forming. The blanket uniformly covers all exposed surfaces and fills any voids to support and protect the entire part.

  1. Some new ideas for Tandem Mirror blankets

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.

    1981-01-01

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

  2. Fusion blanket high-temperature heat transfer

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1983-01-01

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

  3. Tritium behaviour in ceramic breeder blankets

    International Nuclear Information System (INIS)

    Miller, J.M.

    1989-01-01

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

  4. Breeding blanket development. Tritium release from breeder

    International Nuclear Information System (INIS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nagao, Yoshiharu

    2006-01-01

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

  5. Mockup testing of remote systems for zirconium fuel dissolution process at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Paige, D.M.

    1979-01-01

    A facility is being constructed at the Idaho National Engineering Laboratory for storage and dissolution of spent zirconium reactor fuels. The dissolution is carried out in chemical type equipment contained in a large shielded cell. The design provides for remote operations and maintenance as required. Equipment predicted to fail within 5 years is designed for remote maintenance. Each system was fabricated for mockup testing using readily available materials. The mockups were tested, redesigned, and retested until satisfactory remote designs were achieved. Records were made of all the work. All design changes were then incorporated into the ongoing detailed design for the actual equipment. Several of these systems are discussed and they include valve replacement, pump replacement, waste solids handling, mechanism operations and others. The mockup program has saved time and money by eliminating many future problems. In addition, the mockup program will continue through construction, cold startup, and hot operations

  6. High heat flux test of tungsten brazed mock-ups developed for KSTAR divertor

    Energy Technology Data Exchange (ETDEWEB)

    Song, J.H. [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, K.M., E-mail: kyungmin@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Hong, S.H.; Kim, H.T.; Park, S.H.; Park, H.K.; Ahn, H.J. [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, S.K.; Lee, D.W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    The tungsten (W) brazed flat type mock-up which consists of W, OFHC-Cu (oxygen-free high conductive copper) and CuCrZr alloy has been designed for KSTAR divertor in preparation for KSTAR upgrade with 17 MW heating power. For verification of the W brazed mock-up, the high heat flux test is performed at KoHLT-EB (Korea High Heat Load Test Facility-Electron Beam) in KAERI (Korea Atomic Energy Research Institute). Three mock-ups are tested for several thousand thermal cycles with absorbed heat flux up to 5 MW/m{sup 2} for 20 s duration. There is no evidence of the failure at the bonding joints of all mock-ups after HHF test. Finite element analysis (FEA) is performed to interpret the result of the test. As a result, it is considered that the local area in the water is in the subcooled boiling regime.

  7. Experimental test campaign on an ITER divertor mock-up

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D.

    2002-01-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests

  8. Experimental test campaign on an ITER divertor mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Dell' Orco, G. E-mail: giovanni.dellorco@brasimone.enea.it; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D

    2002-11-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests.

  9. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1988-01-01

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

  11. Tritium inventory and permeation in liquid breeder blankets

    International Nuclear Information System (INIS)

    Reiter, F.

    1990-01-01

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

  12. Neutronics Evaluation of Lithium-Based Ternary Alloys in IFE Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Jolodosky, A. [Univ. of California, Berkeley, CA (United States); Fratoni, M. [Univ. of California, Berkeley, CA (United States)

    2015-09-22

    , low electrical conductivity and therefore low MHD pressure drop, low chemical reactivity, and extremely low tritium inventory; the addition of sodium (FLiNaBe) has been considered because it retains the properties of FliBe but also lowers the melting point. Although many of these blanket concepts are promising, challenges still remain. The limited amount of beryllium available poses a problem for ceramic breeders such as the HCPB. FLiBe and FLiNaBe are highly viscous and have a low thermal conductivity. Lithium lead possesses a poor thermal conductivity which can cause problems in both DCLL and LiPb blankets. Additionally, the tritium permeation from these two blankets into plant components can be a problem and must be reduced. Consequently, Lawrence Livermore National Laboratory (LLNL) is attempting to develop a lithium-based alloy—most likely a ternary alloy—which maintains the beneficial properties of lithium (e.g. high tritium breeding and solubility) while reducing overall flammability concerns for use in the blanket of an inertial fusion energy (IFE) power plant. The LLNL concept employs inertial confinement fusion (ICF) through the use of lasers aimed at an indirect-driven target composed of deuterium-tritium fuel. The fusion driver/target design implements the same physics currently experimented at the National Ignition Facility (NIF). The plant uses lithium in both the primary coolant and blanket; therefore, lithium-related hazards are of primary concern. Although reducing chemical reactivity is the primary motivation for the development of new lithium alloys, the successful candidates will have to guarantee acceptable performance in all their functions. The scope of this study is to evaluate the neutronics performance of a large number of lithium-based alloys in the blanket of the IFE engine and assess their properties upon activation. This manuscript is organized as follows: Section 12 presents the models and methodologies used for the analysis; Section

  13. The Artificial Masterpiece - An analysis of mockup and live performance perception

    OpenAIRE

    Fagre, Jennifer Karen

    2017-01-01

    This dissertation will explore how mockups (sampled/synthesized music ) have evolved and incorporated themselves into today's music industry. Mockups are created using sample libraries, which are collections of digital or acoustic sound recordings, known as samples, for use by composers, arrangers, performers, and producers of music. Especially pertaining to film/TV and pop music, the amount of control available to producers in the studio with both sampled sounds and live recorded sounds ha...

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

    International Nuclear Information System (INIS)

    Lux, M.

    2007-01-01

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

  15. Investigation of Be/Cu joints via HHF tests of small-scale mockups

    Energy Technology Data Exchange (ETDEWEB)

    Giniatulin, R.; Gervash, A.; Komarov, V.L.; Litunovsky, N.; Mazul, I.; Yablokov, N. [Efremov Inst., St. Petersburg (Russian Federation)

    1998-01-01

    Beryllium-copper (Be/Cu) joints in divertor components work under cyclic heat loads. To develop reliable joints small-scale mockups are fabricated by divertor technologies and tested under the divertor conditions. One of the critical damaging factors that exist in the divertor and have to be simulated is thermocyclic heat loads in the range of 1-15 MW/m{sup 2}. This work presents the divertor mockups that have beryllium tiles with different dimensions (5 x 5 - 44 x 44) mm{sup 2} brazed with copper alloy heat sink. The electron beam was used to braze these mockups so as to decrease the formation of brittle intermetallic layers. The description of mockups design, geometry of armour tiles and fabrication techniques are presented in the paper. The results of screening and thermocyclic tests of these mockups in the heat flux range of 2-12 MW/m{sup 2} with a number of cycles {approx}10{sup 3} are presented. The results of metallographic analysis are also presented. The results of fabrication and testing with small-scale mockups for first wall application are also described. (author)

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

    International Nuclear Information System (INIS)

    Beller, D.E.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1980-09-01

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

  18. INTOR first wall/blanket/shield activity

    International Nuclear Information System (INIS)

    Gohar, Y.; Billone, M.C.; Cha, Y.S.; Finn, P.A.; Hassanein, A.M.; Liu, Y.Y.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.

    1986-01-01

    The main emphasis of the INTOR first wall/blanket/shield (FWBS) during this period has been upon the tritium breeding issues. The objective is to develop a FWBS concept which produces the tritium requirement for INTOR operation and uses a small fraction of the first wall surface area. The FWBS is constrained by the dimensions of the reference design and the protection criteria required for different reactor components. The blanket extrapolation to commercial power reactor conditions and the proper temperature for power extraction have been sacrificed to achieve the highest possible local tritium breeding ratio (TBR). In addition, several other factors that have been considered in the blanket survey study include safety, reliability, lifetime fluence, number of burn cycles, simplicity, cost, and development issues. The implications of different tritium supply scenarios were discussed from the cost and availability for INTOR conditions. A wide variety of blanket options was explored in a preliminary way to determine feasibility and to see if they can satisfy the INTOR conditions. This survey and related issues are summarized in this report. Also discussed are material design requirements, thermal hydraulic considerations, structure analyses, tritium permeation through the first wall into the coolant, and tritium inventory

  19. Optimization of beryllium for fusion blanket applications

    International Nuclear Information System (INIS)

    Billone, M.C.

    1993-01-01

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

  20. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  1. European blanket development for a demo reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-09-01

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

  3. MHD considerations for poloidal-toroidal coolant ducts of self-cooled blankets

    International Nuclear Information System (INIS)

    Hua, T.Q.; Walker, J.S.

    1990-01-01

    Magnetohydrodynamic flows of liquid metals through sharp elbow ducts with rectangular cross sections and with thin conducting walls in the presence of strong uniform magnetic fields are examined. The geometries simulate the poloidaltoroidal coolant channels in fusion tokamak blankets. Analysis for obtaining the three-dimensional numerical solutions are described. Results for pressure drop, velocity profiles and flow distribution are predicted for the upcoming joint ANL/KfK sharp elbow experiment. Results from a parametric study using fusion relevant parameters to investigate the three-dimensional pressure drop are presented for possible applications to blanket designs. 10 refs., 9 refs

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

    Science.gov (United States)

    Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

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

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

    International Nuclear Information System (INIS)

    Liu, K.C.

    1977-01-01

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

  6. The power-sharing formula for a seed/blanket core-resolution of a paradox

    International Nuclear Information System (INIS)

    Radkowsky, A.; Segev, M.; Galperin, A.

    1986-01-01

    The ''classical'' formula for the sharing of power between a seed and blanket was based on the two-group diffusion theory model and gave good agreement with experiments conducted in the original Shippingport program and with transport theory. Recently an extensive series of calculations on seed/blanket assemblies showed that the power sharing deviates widely from the classical formula but paradoxically is in good agreement with the one-group formula, which neglects the back leakage of thermal neutrons from the blanket to the seed. The power-sharing formula has now been rederived, and the paradox is resolved by taking into account epithermal absorptions in the seeds. The diffusion theory model is important as a guide to formulating innovative concepts for improved core designs

  7. Engineering studies of tritium recovery from CTR blankets and plasma exhaust

    International Nuclear Information System (INIS)

    Watson, J.S.

    1975-01-01

    Engineering studies on tritium handling problems in fusion reactors have included conceptual and experimental studies of techniques for recovery of tritium bred in the reactor blanket and conceptual designs for recovery and processing of tritium from plasma exhausts. The process requirements and promising techniques for the blanket system depend upon the materials used for the blanket, coolant, and structure and on the operating temperatures. Process requirements are likely to be set in some systems by allowable loss rates to the steam system or by inventory considerations. Conceptual studies have also been made for tritium handling equipment for fueling, recovery, and processing in plasma recycle systems of fusion reactors, and a specific design has been prepared for ''near-term'' Tokamak experiments. (auth)

  8. Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER

    International Nuclear Information System (INIS)

    Yum, Soo Been; Lee, Eo Hwak; Lee, Dong Won; Park, Goon Cherl

    2013-01-01

    Highlights: ► In this study, helium supplying system was constructed. ► Preparation for heat transfer experiment in KO TBM condition using helium supplying system was progressed. ► To get more applicable results, test matrix was made to cover the condition for KO TBM. ► Using CFD code; CFX 11, validation and modification for system code GAMMA was performed. -- Abstract: By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m 2 using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX

  9. A SCALE-UP Mock-Up: Comparison of Student Learning Gains in High- and Low-Tech Active-Learning Environments

    Science.gov (United States)

    Soneral, Paula A. G.; Wyse, Sara A.

    2017-01-01

    Student-centered learning environments with upside-down pedagogies (SCALE-UP) are widely implemented at institutions across the country, and learning gains from these classrooms have been well documented. This study investigates the specific design feature(s) of the SCALE-UP classroom most conducive to teaching and learning. Using pilot survey data from instructors and students to prioritize the most salient SCALE-UP classroom features, we created a low-tech “Mock-up” version of this classroom and tested the impact of these features on student learning, attitudes, and satisfaction using a quasi-­experimental setup. The same instructor taught two sections of an introductory biology course in the SCALE-UP and Mock-up rooms. Although students in both sections were equivalent in terms of gender, grade point average, incoming ACT, and drop/fail/withdraw rate, the Mock-up classroom enrolled significantly more freshmen. Controlling for class standing, multiple regression modeling revealed no significant differences in exam, in-class, preclass, and Introduction to Molecular and Cellular Biology Concept Inventory scores between the SCALE-UP and Mock-up classrooms. Thematic analysis of student comments highlighted that collaboration and whiteboards enhanced the learning experience, but technology was not important. Student satisfaction and attitudes were comparable. These results suggest that the benefits of a SCALE-UP experience can be achieved at lower cost without technology features. PMID:28213582

  10. Development of Joining Technologies for the ITER Blanket First Wall

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Byoung Kwon; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan; Lee, Dong Won; Kim, Suk Kwon [KAERI, Daejeon (Korea, Republic of)

    2011-01-15

    The design of the ITER blanket first wall includes the Beryllium amour tiles joined to CuCrZr heat sink with stainless steel cooling tubes. For the ITER application, the Be/CuCrZr/SS joint was proposed as a first wall material. The joining of Be/CuCrZr as well as CuCrZr/SS was generally carried out by using a hot isostatic pressing (CuC) in many countries. The joining strength for Be/CuCrZr is relatively lower than that for CuCrZr/SS, since we usually forms surface oxides (BeO) and brittle a metallics with Cu. Therefore, the joining technology for the Be/CuCrZr joint has been investigated. Be is apt to adsorb oxygen in an air atmosphere, so we should be etched to eliminate the surface pre-oxide using a chemical solution and Ar ions in a vacuum chamber. Then we is coated with a first was to prevent further oxidation. The kinds of a first we are chosen to be able to enhance the joining strength as inhibiting excessive be diffusion. The performance of the Be/CuCrZr/SS joint used for the ITER first wall is primarily dependent on the joining strength of the Be/CuCrZr interface. The Cr/Cu and Ti/Cr/Cu interlayers enabled the successful joining of be tile to CuCrZr plate. Moreover, ion-beam assisted deposition (IBAD) increased joining strength of the Be/CuCrZr joint mock-ups. IBAD induced the increased packing of depositing atoms, which resulted in denser and more adhesive interlayers. The interlayers formed by IBAD process revealed about 40% improved resistance to the scratch test. It is suggested that the improved adhesion of coating interlayers enabled tight joining of Be and CuCrZr blocks. As compared to without IBAD coating, the shear strength as well as the 4-point bend strength were increased more than 20% depending on interlayer types and coating conditions

  11. Development of Joining Technologies for the ITER Blanket First Wall

    International Nuclear Information System (INIS)

    Choi, Byoung Kwon; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan; Lee, Dong Won; Kim, Suk Kwon

    2011-01-01

    The design of the ITER blanket first wall includes the Beryllium amour tiles joined to CuCrZr heat sink with stainless steel cooling tubes. For the ITER application, the Be/CuCrZr/SS joint was proposed as a first wall material. The joining of Be/CuCrZr as well as CuCrZr/SS was generally carried out by using a hot isostatic pressing (CuC) in many countries. The joining strength for Be/CuCrZr is relatively lower than that for CuCrZr/SS, since we usually forms surface oxides (BeO) and brittle a metallics with Cu. Therefore, the joining technology for the Be/CuCrZr joint has been investigated. Be is apt to adsorb oxygen in an air atmosphere, so we should be etched to eliminate the surface pre-oxide using a chemical solution and Ar ions in a vacuum chamber. Then we is coated with a first was to prevent further oxidation. The kinds of a first we are chosen to be able to enhance the joining strength as inhibiting excessive be diffusion. The performance of the Be/CuCrZr/SS joint used for the ITER first wall is primarily dependent on the joining strength of the Be/CuCrZr interface. The Cr/Cu and Ti/Cr/Cu interlayers enabled the successful joining of be tile to CuCrZr plate. Moreover, ion-beam assisted deposition (IBAD) increased joining strength of the Be/CuCrZr joint mock-ups. IBAD induced the increased packing of depositing atoms, which resulted in denser and more adhesive interlayers. The interlayers formed by IBAD process revealed about 40% improved resistance to the scratch test. It is suggested that the improved adhesion of coating interlayers enabled tight joining of Be and CuCrZr blocks. As compared to without IBAD coating, the shear strength as well as the 4-point bend strength were increased more than 20% depending on interlayer types and coating conditions

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  13. Performance of silvered Teflon (trademark) thermal control blankets on spacecraft

    Science.gov (United States)

    Pippin, Gary; Stuckey, Wayne; Hemminger, Carol

    1993-01-01

    Silverized Teflon (Ag/FEP) is a widely used passive thermal control material for space applications. The material has a very low alpha/e ratio (less than 0.1) for low operating temperatures and is fabricated with various FEP thicknesses (as the Teflon thickness increases, the emittance increases). It is low outgassing and, because of its flexibility, can be applied around complex, curved shapes. Ag/FEP has achieved multiyear lifetimes under a variety of exposure conditions. This has been demonstrated by the Long Duration Exposure Facility (LDEF), Solar Max, Spacecraft Charging at High Altitudes (SCATHA), and other flight experiments. Ag/FEP material has been held in place on spacecraft by a variety of methods: mechanical clamping, direct adhesive bonding of tapes and sheets, and by Velcro(TM) tape adhesively bonded to back surfaces. On LDEF, for example, 5-mil blankets held by Velcro(TM) and clamping were used for thermal control over 3- by 4-ft areas on each of 17 trays. Adhesively bonded 2- and 5-mil sheets were used on other LDEF experiments, both for thermal control and as tape to hold other thermal control blankets in place. Performance data over extended time periods are available from a number of flights. The observed effects on optical properties, mechanical properties, and surface chemistry will be summarized in this paper. This leads to a discussion of performance life estimates and other design lessons for Ag/FEP thermal control material.

  14. A SCALE-UP Mock-Up: Comparison of Student Learning Gains in High- and Low-Tech Active-Learning Environments.

    Science.gov (United States)

    Soneral, Paula A G; Wyse, Sara A

    2017-01-01

    Student-centered learning environments with upside-down pedagogies (SCALE-UP) are widely implemented at institutions across the country, and learning gains from these classrooms have been well documented. This study investigates the specific design feature(s) of the SCALE-UP classroom most conducive to teaching and learning. Using pilot survey data from instructors and students to prioritize the most salient SCALE-UP classroom features, we created a low-tech "Mock-up" version of this classroom and tested the impact of these features on student learning, attitudes, and satisfaction using a quasi--experimental setup. The same instructor taught two sections of an introductory biology course in the SCALE-UP and Mock-up rooms. Although students in both sections were equivalent in terms of gender, grade point average, incoming ACT, and drop/fail/withdraw rate, the Mock-up classroom enrolled significantly more freshmen. Controlling for class standing, multiple regression modeling revealed no significant differences in exam, in-class, preclass, and Introduction to Molecular and Cellular Biology Concept Inventory scores between the SCALE-UP and Mock-up classrooms. Thematic analysis of student comments highlighted that collaboration and whiteboards enhanced the learning experience, but technology was not important. Student satisfaction and attitudes were comparable. These results suggest that the benefits of a SCALE-UP experience can be achieved at lower cost without technology features. © 2017 P. A. G. Soneral and S. A. Wyse. CBE—Life Sciences Education © 2017 The American Society for Cell Biology. This article is distributed by The American Society for Cell Biology under license from the author(s). It is available to the public under an Attribution–Noncommercial–Share Alike 3.0 Unported Creative Commons License (http://creativecommons.org/licenses/by-nc-sa/3.0).

  15. Packed-fluidized-bed blanket concept for a thorium-fueled commercial tokamak hybrid reactor

    International Nuclear Information System (INIS)

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

    1980-09-01

    A preliminary design of a thorium blanket was carried out as a part of the Commercial Tokamak Hybrid Reactor (CTHR) study. A fixed fuel blanket concept was developed as the reference CTHR blanket with uranium carbide fuel and helium coolant. A fixed fuel blanket was initially evaluated for the thorium blanket study. Subsequently, a new type of hybrid blanket, a packed-fluidized bed (PFB), was conceived. The PFB blanket concept has a number of unique features that may solve some of the problems encountered in the design of tokamak hybrid reactor blankets. This report documents the thorium blanket study and describes the feasibility assessment of the PFB blanket concept

  16. The lithium blanket program at the LOTUS facility

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-09-01

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

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

    International Nuclear Information System (INIS)

    Chen Hongli; Wu Yican; Bai Yunqing

    2006-01-01

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

  19. High-240Pu fuel worth in the Fast Test Reactor Engineering Mockup

    International Nuclear Information System (INIS)

    Daughtry, J.W.; Dobbin, K.D.

    1975-01-01

    Reactivity effects associated with the replacement of low- 240 Pu fuel with high- 240 Pu fuel were calculated and compared to measurements made in the FTR Engineering Mockup Critical (EMC). When the Pu and U isotopic compositions were changed in a way that increased the amounts of 240 Pu and 241 Pu and reduced the amounts of 239 Pu and 238 U while conserving total fissile mass and total fertile mass, the reactivity effect was positive. Calculation-to-experiment bias factors were obtained for this type of change and for the replacement of Fe 2 O 3 with U 3 O 8 in subassembly-size zones of the EMC. The k/sub e/--k/sub c/ bias decreased when high- 240 Pu fuel was introduced and increased when Fe 2 O 3 was replaced with U 3 O 8 . When the two changes were combined, their effects on the k/sub e/ --k/sub c/ bias tended to cancel out. The work described is related to plans for the utilization of light water reactor discharge Pu in the FTR

  20. Thermo-siphon Mock-up Test for the HANARO-CNS

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jungwoon; Lee, Kye Hong; Kim, Hark Rho; Kim, Youngki; Kim, Myong Seop; Wu, Sang Ik; Kim, Bong Su

    2006-04-15

    In order to moderate thermal neutrons into cold neutrons, the liquid hydrogen is selected as a moderator for the HANARO CNS. By the non-nuclear heat load and nuclear heat load induced from collision of gamma-ray, beta-ray, and thermal neutrons, the liquid hydrogen in the moderator cell evaporates and flows into the heat exchanger. This evaporated hydrogen gas is liquefied by the cryogenic helium supplied from the helium refrigeration system,, then flows back to the moderator cell. This is so-called two-phase thermo-siphon. The most important point in the stable thermo-siphon is to have the good balance between the cooling capacity of the HRS and the heat load on the moderator cell so as to maintain the stable two-phase liquid level in the moderator cell. Accordingly, for not only the experience of the cryogenic two-phase thermo-siphon but also setup of the operation procedure, the full-scaled mock-up test has been performed using the liquid hydrogen. Through the test, the stable thermo-siphon establishment is confirmed at the cold normal operation; furthermore, the detail design parameter is validated. On top of the normal operation procedure setup, the abnormal operation procedure is settled based on the understanding the abnormal pressure and temperature transient dynamics in the hydrogen system.

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

    International Nuclear Information System (INIS)

    Daigo Tsuru; Mikio Enoeda; Masato Akiba

    2006-01-01

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

  2. In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions in the LVR-15 nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, Jan [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Entler, Slavomir, E-mail: slavomir.entler@cvrez.cz [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Vsolak, Rudolf; Klabik, Tomas [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Zlamal, Ondrej [CEZ, Duhova 2/1444, 140 53 Praha 4 (Czech Republic); Bellin, Boris; Zacchia, Francesco [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • Irradiated thermal fatigue testing of the ITER primary first wall mock-ups. • Cyclic heat flux of 0.5 MW/m{sup 2} in the neutron field of the nuclear reactor core. • 17,040 thermal cycles. • Radiation damage in the range of 0.41–1.17 dpa depending on the material. - Abstract: The TW3 in-pile rig enabled the thermal fatigue testing of ITER primary first wall mock-ups in the core of the nuclear reactor. This experiment investigated the neutron irradiation influence on the design performance under high heat flux testing. A thermal flux of 0.5 MW/m{sup 2} in the neutron field of the core of the LVR-15 nuclear reactor was applied. Within the scope of the tests with simultaneous neutron irradiation, the TW3 rig reached a record of 17,040 thermal cycles with the radiation damage in the range of 0.41–1.17 dpa depending on the material. Even after a high number of thermal cycles, while being irradiated by neutrons, no damage of the tested mock-ups was visually observed. Further testing and analysis will follow in the Forschungszentrum Juelich.

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

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  5. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    Carvalho, S.H. de.

    1980-03-01

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

  6. 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)

  7. Flow balancing in liquid metal blankets

    International Nuclear Information System (INIS)

    Tillack, M.S.; Morley, N.B.

    1995-01-01

    Non-uniform flow distribution between parallel channels is one of the most serious concerns for self-cooled liquid metal blankets with electrically insulated walls. We show that uncertainties in flow distribution can be dramatically reduced by relatively simple design modifications. Several design features which impose flow uniformity by electrically coupling parallel channels are surveyed. Basic mechanisms for ''flow balancing'' are described, and a particular self-regulating concept using discrete passive electrodes is proposed for the US ITER advanced blanket concept. Scoping calculations suggest that this simple technique can be very powerful in equalizing the flow, even with massive insulator failures in individual channels. More detailed analyses and experimental verification will be required to demonstrate this concept for ITER. (orig.)

  8. Present status of irradiation tests on tritium breeding blanket for fusion reactor

    International Nuclear Information System (INIS)

    Futamura, Yoshiaki; Sagawa, Hisashi; Shimakawa, Satoshi; Tsuchiya, Kunihiko; Kuroda, Toshimasa; Kawamura, Hiroshi.

    1994-01-01

    To develop a tritium breeding blanket for a fusion reactor, irradiation tests in fission reactors are indispensable for obtaining data on irradiation effects on materials, and neutronics/thermal characteristics and tritium production/recovery performance of the blanket. Various irradiation tests have been conducted in the world, especially to investigate tritium release characteristics from tritium breeding and neutron multiplier materials, and materials integrity under irradiation. In Japan, VOM experiments at JRR-2 for ceramic breeders and experiments at JMTR for ceramic breeders and beryllium as a neutron multiplier have been performed. Several universities have also investigated ceramic breeders. In the EC, the EXOTIC experiments at HFR in the Netherlands and the SIBELIUS, the LILA, the LISA and the MOZART experiments for ceramic breeders have carried out. In Canada, NRU has been used for the CRITIC experiments. The TRIO experiments at ORR(ORNL), experiments at RTNS-II, FUBR and ATR have been conducted in the USA. The last two are experiments with high neutron fluence aiming at investigating materials integrity under irradiation. The BEATRIX-I and -II experiments have proceeded under international collaboration of Japan, Canada, the EC and the USA. This report shows the present status of these irradiation tests following a review of the blanket design in the ITER CDA(Conceptual Design Activity). (author)

  9. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  11. Conceptual design of Blanket Remote Handling System for CFETR

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  12. Conceptual design of Blanket Remote Handling System for CFETR

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  13. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Waganer, L.M.

    1985-01-01

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

  14. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    1980-08-01

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

  15. Nuclear Analysis of an ITER Blanket Module

    Science.gov (United States)

    Chiovaro, P.; Di Maio, P. A.; Parrinello, V.

    2013-08-01

    ITER blanket system is the reactor's plasma-facing component, it is mainly devoted to provide the thermal and nuclear shielding of the Vacuum Vessel and external ITER components, being intended also to act as plasma limiter. It consists of 440 individual modules which are located in the inboard, upper and outboard regions of the reactor. In this paper attention has been focused on to a single outboard blanket module located in the equatorial zone, whose nuclear response under irradiation has been investigated following a numerical approach based on the Monte Carlo method and adopting the MCNP5 code. The main features of this blanket module nuclear behaviour have been determined, paying particular attention to energy and spatial distribution of the neutron flux and deposited nuclear power together with the spatial distribution of its volumetric density. Moreover, the neutronic damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rates. Finally, an activation analysis has been performed with FISPACT inventory code using, as input, the evaluated neutron spectrum to assess the module specific activity and contact dose rate after irradiation under a specific operating scenario.

  16. Tokamak blanket design study, final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

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

  17. Overview of the Last Progresses for the European Test Blanket Modules Projects

    International Nuclear Information System (INIS)

    Salavy, J.-F.; Rampal, G.; Boccaccini, L.V.; Meyder, R.; Neuberger, H.; Laesser, R.; Poitevin, Y.; Zmitko, M.; Rigal, E.

    2006-01-01

    The long-term objective of the EU Breeding Blankets programme is the development of DEMO breeding blankets, which shall assure tritium self-sufficiency, an economically attractive use of the heat produced inside the blankets for electricity generation and a sufficiently high shielding of the superconducting magnets for long time operation. In the short-term so-called DEMO relevant Test Blanket Modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. The Helium Cooled Lithium-Lead (HCLL) breeder blanket and the Helium Cooled Pebble Bed (HCPB) concepts are the two breeder blanket lines presently developed by the EU. The main objective of the EU test strategy related to TBMs in ITER is to provide the necessary information for the design and fabrication of breeding blankets for a future DEMO reactor. EU TBMs shall therefore use the same structural and functional materials, apply similar fabrication technologies, and test adequate processes and components. This paper gives an overview of the last progresses in terms of system design, calculations, test program, safety and R-and-D done these last two years in order to cope with the ambitious objective to introduce two EU TBM systems for day-1 of ITER operation. The engineering design of the two systems is mostly concluded and the priority is now on the development and qualification of the fabrication technologies. From calculations point of view, the last modelling efforts related to the thermal-hydraulic of the first wall, the tritium behaviour, and the box thermal and mechanical resistance in accidental conditions are presented. Last features of the TBM and cooling system designs and integration in ITER reactor are highlighted. In particular, this paper also describes the safety and licensing analyses performed for each concept to be able to include the TBM systems in the ITER preliminary safety report

  18. Mock-up test of remote controlled dismantling apparatus for large-sized vessels (contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Myodo, Masato; Miyajima, Kazutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Okane, Shogo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-03-01

    The Remote dismantling apparatus, which is equipped with multi-units for functioning of washing, cutting, collection of cut pieces and so on, has been constructed to dismantle the large-sized vessels in the JAERI's Reprocessing Test Facility (JRTF). The apparatus has five-axis movement capability and its operation is performed remotely. The mock-up tests were performed to evaluate the applicability of the apparatus to actual dismantling activities by using the mock-ups of LV-3 and LV-5 in the facility. It was confirmed that each unit was satisfactory functioned by remote operation. Efficient procedures for dismantling the large-sized vessel was studied and various date was obtained in the mock-up tests. This apparatus was found to be applicable for the actual dismantling activity in JRTF. (author)

  19. Mock-up qualification and prototype manufacture for ITER current leads

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Tingzhi, E-mail: tingszhou@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Lu, Kun; Ran, Qingxiang; Ding, Kaizhong; Feng, Hansheng; Wu, Huan; Liu, Chenglian; Song, Yuntao [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Niu, Erwu [CNDA, Ministry of Science and Technology, Beijing (China); Bauer, Pierre; Devred, Arnaud [Magnet Division, ITER Organization, Cadarache (France)

    2015-10-15

    Highlights: • Vacuum brazing and electron beam welding qualification. • Machine and assembly strategy of fin type heat exchanger. • Soldering and joint resistance test of superconducting joint. • Pre-preg technology with vacuum bag on insulation. - Abstract: Three types of high temperature superconducting current leads (HTSCL) are designed to carry 68 kA, 55 kA or 10 kA to the ITER magnets. Before the supply of the HTS current lead series, the design and manufacturing process is qualified through mock-ups and prototypes. Seven mock-ups, representing the critical technologies of the current leads, were built and tested successfully in the Institute of Plasma Physics of the Chinese Academy of Sciences (ASIPP) in 2013. After the qualification some design features of the HTS leads were updated. This paper summarizes the qualification through mock-ups. In 2014 ASIPP started the manufacture of the prototypes. The preparation and manufacturing process are also described.

  20. Thermo-mechanical tests of a CFC divertor mock-up

    International Nuclear Information System (INIS)

    Cardella, A.; Akiba, M.; Duwe, R.; Di Pietro, E.; Suzuki, S.; Satoh, K.; Reheis, N.

    1994-01-01

    Thermo-mechanical tests have been performed on a divertor mock-up consisting of a metallic tube armoured with five carbon fibre composite tiles. The tube is inserted the tiles and brazed with TiCuSil braze (monoblock concept). The tube material is TZM, a molybdenum alloy, and the armour material is SEP CARB N112, a high conductivity carbon-carbon composite. Using special surface preparation consisting of laser drilling, small (≅ 500 μm) holes in the composite have been made to increase the surface wetted by the braze and the resistance. The mock-up has been tested at the JAERI 400 kW electron beam test facility JEBIS. The aim of the test was to assess the performance of the mock-up in screening and thermal fatigue tests with particular attention to the behaviour of the armour to heat sink joint. (orig.)

  1. Mock-up test of remote controlled dismantling apparatus for large-sized vessels (contract research)

    International Nuclear Information System (INIS)

    Myodo, Masato; Miyajima, Kazutoshi; Okane, Shogo

    2001-03-01

    The Remote dismantling apparatus, which is equipped with multi-units for functioning of washing, cutting, collection of cut pieces and so on, has been constructed to dismantle the large-sized vessels in the JAERI's Reprocessing Test Facility (JRTF). The apparatus has five-axis movement capability and its operation is performed remotely. The mock-up tests were performed to evaluate the applicability of the apparatus to actual dismantling activities by using the mock-ups of LV-3 and LV-5 in the facility. It was confirmed that each unit was satisfactory functioned by remote operation. Efficient procedures for dismantling the large-sized vessel was studied and various date was obtained in the mock-up tests. This apparatus was found to be applicable for the actual dismantling activity in JRTF. (author)

  2. Nuclear characteristics of D-D fusion reactor blankets

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao

    1978-01-01

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

  3. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.A.

    1980-01-01

    Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture

  4. Seismic analysis methods for LMFBR core and verification with mock-up vibration tests

    International Nuclear Information System (INIS)

    Sasaki, Y.; Kobayashi, T.; Fujimoto, S.

    1988-01-01

    This paper deals with the vibration behaviors of a cluster of core elements with the hexagonal cross section in a barrel under the dynamic excitation due to seismic events. When a strong earthquake excitation is applied to the core support, the cluster of core elements displace to a geometrical limit determined by restraint rings in the barrel, and collisions could occur between adjacent elements as a result of their relative motion. For these reasons, seismic analysis on LMFBR core elements is a complicated non-linear vibration problem, which includes collisions and fluid interactions. In an actual core design, it is hard to include hundreds of elements in the numerical calculations. In order to study the seismic behaviors of core elements, experiments with single row 29 elements (17 core fuel assemblies, 4 radial blanket assemblies, and 8 neutron shield assemblies) simulated all elements in MONJU core central row, and experiments with 7 cluster rows of 37 core fuel assemblies in the core center were performed in a fluid filled tank, using a large-sized shaking table. Moreover, the numerical analyses of these experiments were performed for the validation of simplified and detailed analytical methods. 4 refs, 18 figs

  5. ITER baffle module small-scale mock-ups: first wall thermo-mechanical testing results

    International Nuclear Information System (INIS)

    Severi, Y.; Giancarli, L.; Poitevin, Y.; Salavy, J.F.; Le Marois, G.; Roedig, M.; Vieider, G.

    1998-01-01

    The EU-home team is in charge of the R and D related to the ITER baffle first wall. Five small-scale mock-ups, using Be, CFC and W tiles and different armour/heat-sink material joints under development, have been fabricated and thermomechanically tested in FE200 (Le Creusot) and JUDITH (Juelich) electron beam facilities. The small-scale mock-ups have been submitted to thermo-mechanical fatigue tests (up to failure using accelerating techniques). The objective was to determine the performances of the armour material joints under high heat flux cycles. (orig.)

  6. Fabrication of ITER first wall mock-ups with beryllium armour

    International Nuclear Information System (INIS)

    Mohri, K.; Nomoto, Y.; Uda, M.; Enoeda, M.; Akiba, M.

    2004-01-01

    This paper presents the fabric ability development for the ITER first wall through the fabrication of a real size first wall panel mock-up without beryllium armor and a partial mock-up of the first wall panel with beryllium armor. Microscopic observation and mechanical test of the hot isostatic pressed Be/Cu-alloy joints were also performed of which results showed good bond ability of the joints. Finally the fabrication procedure of the ITER first wall panel has been established. (author)

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

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1987-01-01

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

  8. Behavior of underclad cracks in reactor pressure vessels - evaluation of mechanical analyses with tests on cladded mock-ups

    International Nuclear Information System (INIS)

    Moinereau, D.; Rousselier, G.; Bethmont, M.

    1993-01-01

    Innocuity of underclad flaws in the reactor pressure vessels must be demonstrated in the French safety analyses, particularly in the case of a severe transient at the end of the pressure vessel lifetime, because of the radiation embrittlement of the vessel material. Safety analyses are usually performed with elastic and elasto-plastic analyses taking into account the effect of the stainless steel cladding. EDF has started a program including experiments on large size cladded specimens and their interpretations. The purpose of this program is to evaluate the different methods of fracture analysis used in safety studies. Several specimens made of ferritic steel A508 C1 3 with stainless steel cladding, containing small artificial defects, are loaded in four-point bending. Experiments are performed at very low temperature to simulate radiation embrittlement and to obtain crack instability by cleavage fracture. Three tests have been performed on mock-ups containing a small underclad crack (with depth about 5 mn) and a fourth test has been performed on one mock-up with a larger crack (depth about 13 mn). In each case, crack instability occurred by cleavage fracture in the base metal, without crack arrest, at a temperature of about - 170 deg C. Each test is interpreted using linear elastic analysis and elastic-plastic analysis by two-dimensional finite element computations. The fracture are conservatively predicted: the stress intensity factors deduced from the computations (K cp or K j ) are always greater than the base metal toughness. The comparison between the elastic analyses (including two plasticity corrections) and the elastic-plastic analyses shows that the elastic analyses are often conservative. The beneficial effect of the cladding in the analyses is also shown : the analyses are too conservative if the cladding effects is not taken into account. (authors). 9 figs., 6 tabs., 10 refs

  9. Electromagnetic analysis of ITER shield blanket under VDE

    International Nuclear Information System (INIS)

    Kang Weishan; Chen Jiming; Wu Jihong; Wang Mingxu

    2010-01-01

    Electromagnetic force and torque of ITER shield blanket system and their surrounding major component under vertical displacement event (VDE) were calculated with finite element method. ANSYS APDL was used to simulate the shape and magnitude of plasmas current dynamically in the VDE course, and external magnetic field was imposed, then the induced current distribution inside the all conductor including the blanket was obtained from the calculation. The force and torque for every blanket module was obtained to assess the safety of blanket system under VDE. (authors)

  10. Minimum thickness blanket-shield for fusion reactors

    International Nuclear Information System (INIS)

    Karni, Y.; Greenspan, E.

    1989-01-01

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

  11. Helium Loop for the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  12. Status report. Characterization of Weld Residual Stresses on a Full-Diameter SNF Interim Storage Canister Mockup.

    Energy Technology Data Exchange (ETDEWEB)

    Enos, David [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-08-01

    This report documents the mockup specifications and manufacturing processes; the initial cutting of the mockup into three cylindrical pieces for testing and the measured strain changes that occurred during the cutting process; and the planned weld residual stress characterization activities and the status of those activities.

  13. Comparison between FEM and high heat flux thermal fatigue testing results of ITER divertor plasma facing mock-ups

    Energy Technology Data Exchange (ETDEWEB)

    Crescenzi, F., E-mail: fabio.crescenzi@enea.it; Roccella, S.; Visca, E.; Moriani, A.

    2014-10-15

    Highlights: • Divertor is an important part of the ITER machine. • Finite element analysis allows designers to explore multiple design options, reducing physical prototypes and optimizing design performance. • The hydraulic thermal-mechanical analysis performed by ANSYS and the test results on small-scale mock-ups manufactured by HRP were compared. • FEA results confirmed many experimental data, then it could be very useful for next design optimization. - Abstract: The divertor is one of the most challenging components of “DEMO” the next step ITER machine, so many tasks regarding modeling and experiments have been made in the past years to assess manufacturing processes, materials and thus the life-time of the components. In this context the finite element analysis (FEA) allows designers to explore multiple design options, to reduce physical prototypes and to optimize design performance. The comparison between the hydraulic thermal-mechanical analysis performed by ANSYS WORKBENCH 14.5 and the test results [1] on small-scale mock-ups manufactured with the Hot Radial Pressing (HRP) [2] technology is presented in this paper. During the thermal fatigue testing in the Efremov TSEFEY facility to assess the heat flux load-carrying capability of the mock-ups, only the surface temperature was measured, so the FEA was important because it allowed to know any other information (temperature inside the materials, local water temperature, local stress, etc.). FEA was performed coupling the thermal-hydraulic analysis, that calculated the temperature distributions on the components and the heat transfer coefficient (HTC) between water and heat sink tube, with the mechanical analysis. The comparison between analysis and testing results was based on the temperature maps of the loaded surface and on number of the cycles supported during the testing and those predicted by the mechanical analysis using the experimental fatigue curves for CuCrZr-IG, that is the structural

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

    International Nuclear Information System (INIS)

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

    1989-10-01

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

  15. Systematic methodology for estimating direct capital costs for blanket tritium processing systems

    International Nuclear Information System (INIS)

    Finn, P.A.

    1985-01-01

    This paper describes the methodology developed for estimating the relative capital costs of blanket processing systems. The capital costs of the nine blanket concepts selected in the Blanket Comparison and Selection Study are presented and compared

  16. Structural materials for fusion reactor blanket systems

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  17. About possible technologies of creation nanostructures blankets

    International Nuclear Information System (INIS)

    Blednova, Zh.M.; Chaevskij, M.I.; Rusinov, P.O.

    2008-01-01

    Possible technologies of formation nanostructures blankets are considered: a method of thermal carrying over of weights in the conditions of a high gradient of temperatures; the combined method including cathode-plasma nitriding in the conditions of low pressure and drawing of nitride of the titan in a uniform work cycle; the combined method including high-frequency ionic nitriding and drawing of carbide of chrome by pyrolysis chrome and organic of connections in plasma of the decaying category. Possibility of formation layered nanostructures layers is shown.

  18. ITER solid breeder blanket materials database

    International Nuclear Information System (INIS)

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

    1993-11-01

    The databases for solid breeder ceramics (Li 2 ,O, Li 4 SiO 4 , Li 2 ZrO 3 and LiAlO 2 ) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized

  19. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-06-01

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

  20. Development of joining processes and fabrication of US first wall qualification mockups for ITER

    International Nuclear Information System (INIS)

    Watson, Roger M.; Puskar, Joseph David; Ulrickson, Michael Andrew; Goods, Steven Howard

    2009-01-01

    We report here the fabrication processes used to manufacture US Party Team First Wall Qualification Mockups along with the detailed microstructural characterization and mechanical properties of the Be/CuCrZr/316L HIP bonds. A companion submission to this conference describes details of the PMTF heat flux testing and the performance of the first US FWQM.

  1. The Use of Facade Mockups in in LCA Based Architectural Design

    DEFF Research Database (Denmark)

    Naboni, Emanuele

    2017-01-01

    Life Cycle Assessment is increasingly becoming important in façade architectural design. The presented research aims to describe an LCA architectural design approach based on the use of a Façade Mockup. The approach is applied and tested for the design Telecom Sustainable Campus’ façade, in Rome...

  2. F.B.R. Core mock-up RAPSODIE- I: Experimental analysis

    International Nuclear Information System (INIS)

    Brochard, D.; Buland, P.; Gantenbein, F.

    1990-01-01

    The main phenomena which influence the LMFBR core response to a seismic excitation are the fluid structure interaction and the impacts between subassemblies. To study the core behaviour, seismic tests have been performed on the core mock-up RAPSODIE with or without fluid and restraint ring and for different levels of excitation. This paper summarizes the results of these tests

  3. Design and fabrication of the NASA HL-20 support cradle and interior mockup

    Science.gov (United States)

    Exum, Thurman

    1991-01-01

    An extensive test program involving analysis in both the horizontal and vertical attitudes of the HL-20 will be conducted by NASA-Langley. This necessitated the fabrication of a steel support cradle for the composite Personnel Launch System (PLS) model and an internal mockup to simulate the pilot and passenger compartments.

  4. Team training using full-scale reactor coolant pump seal mock-ups

    International Nuclear Information System (INIS)

    McDonald, T.J.; Hamill, R.W.

    1987-01-01

    The use of full-scale reactor coolant pump (RCP) seal mock-ups has greatly enhanced Northeast Utilities' ability to effectively utilize the team training approach to technical training. With the advent of the Institute of Nuclear Power Operations accreditation come a new emphasis and standards for the integrated training of plant engineering personnel, maintenance mechanics, quality control personnel, and health physics personnel. The results of purchasing full-scale RCP mock-ups to pilot the concept of team training have far exceeded expectations and cost-limiting factors. The initial training program analysis identified RCP seal maintenance as a task that required training for maintenance department personnel. Due to radiation exposure considerations and the unavailability of actual plant equipment for training purposes, the decision was made to procure a mock-up of an RCP seal assembly and housing. This mock-up was designed to facilitate seal cartridge removal, disassembly, assembly, and installation, duplicating all internal components of the seal cartridge and housing area in exact detail

  5. F.B.R. Core mock-up RAPSODIE - II - numerical models

    International Nuclear Information System (INIS)

    Brochard, D.; Hammami, L.; Gantenbein, F.

    1990-01-01

    To study the behaviour of LMFBR cores excited by a seism, tests have been performed on the RAPSODIE core mock-up. The aim of this paper is to present the numerical models used to interprete these tests and the comparisons between calculations and experimental results

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  8. Overview of the TFTB lithium blanket module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

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

  9. Remote handling demonstration of ITER blanket module replacement

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  10. Accelerator driven heavy water blanket on circulating fuel

    International Nuclear Information System (INIS)

    Kazaritsky, V.D.; Blagovolin, P.P.; Mladov, V.R.; Okhlopkov, M.L.; Batyaev, V.F.; Stepanov, N.V.; Seliverstov, V.V.

    1997-01-01

    A conceptual design of a heavy water blanket with circulating fuel for an accelerator driven transmutation system is described. The hybrid system consists of a high-current linear accelerator of protons and 4 targets, each placed inside a subcritical blanket

  11. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2005-03-01

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

  12. An assessment of the base blanket for ITER

    International Nuclear Information System (INIS)

    Raffray, A.R.; Abdou, M.A.; Ying, A.

    1991-01-01

    Ideally, the ITER base blanket would provide the necessary tritium for the reactor to be self-sufficient during operation, while having minimal impact on the overall reactor cost, reliability and safety. A solid breeder blanket has been developed in CDA phase in an attempt to achieve such objectives. The reference solid breeder base blanket configurations at the end of the CDA phase has many attractive features such as a tritium breeding ratio (TBR) of 0.8--0.9 and a reasonably low tritium inventory. However, some concerns regarding the risk, cost and benefit of the base blanket have been raised. These include uncertainties associated with the solid breeder thermal control and the potentially high cost of the amount of Be used to achieve high TBR and to provide the necessary thermal barrier between the high temperature solid breeder and low temperature coolant. This work addresses these concerns. The basis for the selection of a breeding blanket is first discussed in light of the incremental risk, cost and benefits relative to a non-breeding blanket. Key issues associated with the CDA breeding blanket configurations are then analyzed. Finally, alternative schemes that could enhance the attractiveness and flexibility of a breeding blanket are explored

  13. Objectives and status of EUROfusion DEMO blanket studies

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  14. 18 CFR 284.303 - OCS blanket certificates.

    Science.gov (United States)

    2010-04-01

    ... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false OCS blanket certificates. 284.303 Section 284.303 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... Pipelines on Behalf of Others § 284.303 OCS blanket certificates. Every OCS pipeline [as that term is...

  15. The TFTR lithium blanket module program

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  16. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    International Nuclear Information System (INIS)

    Gavila, P.; Riccardi, B.; Pintsuk, G.; Ritz, G.; Kuznetsov, V.; Durocher, A.

    2015-01-01

    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m"2, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m"2. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing program

  17. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Gavila, P., E-mail: pierre.gavila@f4e.europa.eu [Fusion for Energy, 08019 Barcelona (Spain); Riccardi, B. [Fusion for Energy, 08019 Barcelona (Spain); Pintsuk, G. [Forschungszentrum Juelich, 52425 Juelich (Germany); Ritz, G. [AREVA NP, Centre Technique France, 71205 Le Creusot (France); Kuznetsov, V. [JCS “Efremov Institute”, Doroga na Metallostroy 3, Metallostroy, Saint-Petersburg 196641 (Russian Federation); Durocher, A. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 Saint Paul-lez-Durance (France)

    2015-10-15

    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m{sup 2}, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m{sup 2}. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  19. Vibration damage testing of thermal barrier fibrous blanket insulation

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  20. Axial blanket enrichment optimization of the NPP Krsko fuel

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2001-01-01

    In this paper optimal axial blanket enrichment of the NPP Krsko fuel is investigated. Since the optimization is dictated by economic categories that can significantly vary in time, two step approach is applied. In the first step simple relationship between the equivalent change in enrichment of axial blankets and central fuel region is established. The relationship is afterwards processed with economic criteria and constraints to obtain optimal axial blanket enrichment. In the analysis realistic NPP Krsko conditions are considered. Except for the fuel enrichment all other fuel characteristics are the same as in the fuel used in the few most recent cycles. A typical reload cycle after the plant power uprate is examined. Analysis has shown that the current blanket enrichment is close to the optimal. Blanket enrichment reduction results in an approximately 100 000 US$ savings per fuel cycle.(author)

  1. Self-cooled liquid-metal blanket concept

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  2. An electro-hydraulic servo control system research for CFETR blanket RH

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Changqi [Hefei University of Technology, Hefei 230009, Anhui (China); Tang, Hongjun, E-mail: taurustang@126.com [Hefei University of Technology, Hefei 230009, Anhui (China); Qi, Songsong [Hefei University of Technology, Hefei 230009, Anhui (China); Cheng, Yong; Feng, Hansheng; Peng, Xuebing; Song, Yuntao [Institute of Plasma Physics Chinese Academy of Sciences, Hefei 230031, Anhui (China)

    2014-11-15

    Highlights: • We discussed the conceptual design of CFETR blanket RH maintenance system. • The mathematical model of electro-hydraulic servo system was calculated. • A fuzzy adaptive PD controller was designed based on control theory and experience. • The co-simulation models of the system were established with AMESim/Simulink. • The fuzzy adaptive PD algorithm was designed as the core strategy of the system. - Abstract: Based on the technical design requirements of China Fusion Engineering Test Reactor (CFETR) blanket remote handling (RH) maintenance, this paper focus on the control method of achieving high synchronization accuracy of electro-hydraulic servo system. Based on fuzzy control theory and practical experience, a fuzzy adaptive proportional-derivative (PD) controller was designed. Then a more precise co-simulation model was established with AMESim/Simulink. Through the analysis of simulation results, a fuzzy adaptive PD control algorithm was designed as the core strategy of electro-hydraulic servo control system.

  3. An electro-hydraulic servo control system research for CFETR blanket RH

    International Nuclear Information System (INIS)

    Chen, Changqi; Tang, Hongjun; Qi, Songsong; Cheng, Yong; Feng, Hansheng; Peng, Xuebing; Song, Yuntao

    2014-01-01

    Highlights: • We discussed the conceptual design of CFETR blanket RH maintenance system. • The mathematical model of electro-hydraulic servo system was calculated. • A fuzzy adaptive PD controller was designed based on control theory and experience. • The co-simulation models of the system were established with AMESim/Simulink. • The fuzzy adaptive PD algorithm was designed as the core strategy of the system. - Abstract: Based on the technical design requirements of China Fusion Engineering Test Reactor (CFETR) blanket remote handling (RH) maintenance, this paper focus on the control method of achieving high synchronization accuracy of electro-hydraulic servo system. Based on fuzzy control theory and practical experience, a fuzzy adaptive proportional-derivative (PD) controller was designed. Then a more precise co-simulation model was established with AMESim/Simulink. Through the analysis of simulation results, a fuzzy adaptive PD control algorithm was designed as the core strategy of electro-hydraulic servo control system

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

    International Nuclear Information System (INIS)

    Dutta Gupta, P.B.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    1995-01-01

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

  8. Proposal to negotiate the renewal of two blanket purchase contracts for the supply of automation systems and fieldbus equipment

    CERN Document Server

    2005-01-01

    This document concerns the renewal of two blanket purchase contracts for the supply of automation systems and fieldbus equipment for the CERN accelerators, technical services and experiments. The Finance Committee is invited to agree to the negotiation of the renewal of two blanket purchase contracts with SCHNEIDER ELECTRIC (FR) and SIEMENS SUISSE (CH) for the supply of automation systems and fieldbus equipment for a period of three years. The total annual amount covering both contracts will not exceed 3 000 000 Swiss francs, subject to revision for inflation.

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  10. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1983-06-01

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

  11. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1984-02-01

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

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

    International Nuclear Information System (INIS)

    Ishitsuka, E.

    2002-01-01

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

  13. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-12-01

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

  14. Heat transfer problems in gas-cooled solid blankets

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    International Nuclear Information System (INIS)

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

    1986-06-01

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

  16. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

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

    1992-06-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  18. Availability analysis of the ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

    2015-01-01

    The ITER blanket remote handling system (BRHS) is required to replace 440 blanket first wall panels in a two-year maintenance period. To investigate this capability, an availability analysis of the system was carried out. Following the analysis procedure defined by the ITER organization, the availability analysis consists of a functional analysis and a reliability block diagram analysis. In addition, three measures to improve availability were implemented: procurement of spare parts, in-vessel replacement of cameras, and simultaneous replacement of umbilical cables. The availability analysis confirmed those measures improve the availability and capability of the BRHS to replace 440 blanket first wall panels in two years. (author)

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

    International Nuclear Information System (INIS)

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

    1985-02-01

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

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

    International Nuclear Information System (INIS)

    1984-03-01

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

  1. A Li-particulate blanket concept for ITER

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  2. Status report on active stabilisation of a linear collider final focus quadrupole mock-up

    International Nuclear Information System (INIS)

    Lottin, J.; Brunetti, L.; Formosa, F.; Adloff, C.; Bastian, Y.; Bolzon, B.; Cadoux, F.; Geffroy, N.; Girard, C.; Jeremie, A.; Karyotakis, Y.; Peltier, F.

    2006-01-01

    The measurements done with the sensors available in our laboratories used for ground motion analysis are presented. The first sensors studied are seismic sensors measuring ground velocity, other sensors are accelerometers available for measuring ground acceleration. The first step has been to characterize the sensors, the second step has been to model and simulate the acceleration in order to identify Eigen frequencies and to display mode shapes. The third step has been to assess the performances of a new algorithm for disturbance rejection. In order to facilitate the analysis, a reduced-size mock-up has been used. The goal was to eliminate or at least to reduce as much as possible the main frequencies of the disturbance. A new mock-up is currently being developed that will have a geometry closer to a final focus quadrupole. Measurements will be done to validate the whole system in view of active stabilization for a future linear collider

  3. Fabrication of small mock-ups for the KO HCCR TBM

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Jae Sung; Kim, Suk Kwon; Lee, Eo Hwak; Jin, Hyung Gon; Lee, Dong Won [KAERI, Daejeon (Korea, Republic of); Cho, Seung Yon [NFRI, Daejeon (Korea, Republic of)

    2016-05-15

    A fabrication procedure for the manufacturing of the HCCR TBM sub-module was performed and small mock-ups were fabricated using an E-beam and laser beam weld to verify the manufacturing procedure and method of the HCCR TBM sub-module. To establish and optimize the welding procedure in an E-beam weld from ARAA material, the distortion and radiographic tests were carried out from the E-beam weld results. It could be noted that a small amount of distortion occurred, but the values are small enough to neglect for the fabrication. In addition, a helium leak test and water pressure test will be performed for verification of the fabricated small mock-ups.

  4. Status report on active stabilisation of a linear collider final focus quadrupole mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Lottin, J.; Brunetti, L.; Formosa, F. [Universite de Savoie, ESIA, 74 - Annecy (France); Adloff, C.; Bastian, Y.; Bolzon, B.; Cadoux, F.; Geffroy, N.; Girard, C.; Jeremie, A.; Karyotakis, Y.; Peltier, F. [LAPP-IN2P3-CNRS, 74 - Annecy-le-Vieux (France)

    2006-07-01

    The measurements done with the sensors available in our laboratories used for ground motion analysis are presented. The first sensors studied are seismic sensors measuring ground velocity, other sensors are accelerometers available for measuring ground acceleration. The first step has been to characterize the sensors, the second step has been to model and simulate the acceleration in order to identify Eigen frequencies and to display mode shapes. The third step has been to assess the performances of a new algorithm for disturbance rejection. In order to facilitate the analysis, a reduced-size mock-up has been used. The goal was to eliminate or at least to reduce as much as possible the main frequencies of the disturbance. A new mock-up is currently being developed that will have a geometry closer to a final focus quadrupole. Measurements will be done to validate the whole system in view of active stabilization for a future linear collider.

  5. Overview of the EU small scale mock-up tests for ITER high heat flux components

    International Nuclear Information System (INIS)

    Vieider, G.; Barabash, V.; Cardella, A.

    1998-01-01

    This task within the EU R and D for ITER was aimed at the development of basic manufacturing solutions for the high heat flux plasma facing components such as the divertor targets, the baffles and limiters. More than 50 representative small-scale mock-ups have been manufactured with beryllium, carbon and tungsten armour using various joining technologies. High heat flux testing of 20 of these mock-ups showed the carbon mono-blocks to be the most robust solution, surviving 2000 cycles at absorbed heat fluxes of up to 24 MW m -2 . With flat armour tiles rapid joint failures occurred at 5-16 MW m -2 depending on joining technology and armour material. These test results serve as a basis for the selection of manufacturing options and materials for the prototypes now being ordered. (orig.)

  6. Chemical-Cleaning Demonstration Test No. 2 in a mock-up steam generator

    International Nuclear Information System (INIS)

    Jevec, J.M.; Leedy, W.S.

    1983-04-01

    This report describes the results of the mockup demonstration test of the first modified baseline process under Contract S-127, Chemical Cleaning of Nuclear Steam Generators. The objective of this program is to determine the feasibility of cleaning the secondary side of nuclear steam generators with state-of-the-art chemical cleaning technology. The first step was to benchmark a baseline process. This process was then modified to attempt to eliminate the causes of unacceptable cleaning performance. The modified baseline process consists of an EDTA/H 2 O 2 -based copper solvent and a near-neutral, EDTA/N 2 H 4 -based magnetite and crevice solvent. This report also presents the results of three inhibitor evaluation mockup runs used in the evaluation of the modified baseline process

  7. Anaerobic digestion of cheese whey using up-flow anaerobic sludge blanket reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yan, J.Q.; Lo, K.V.; Liao, P.H.

    1989-01-01

    Anaerobic treatment of cheese whey using a 17.5-litre up-flow anaerobic sludge blanket reactor was investigated in the laboratory. The reactor was studied over a range of influent concentration from 4.5 to 38.1 g chemical oxygen demand per litre at a constant hydraulic retention time of 5 days. The reactor start-up and the sludge acclimatization were discussed. The reactor performance in terms of methane production, volatile fatty acids conversion, sludge net growth and chemical oxygen demand reduction were also presented in this paper. Over 97% chemical oxygen demand reduction was achieved in this experiment. At the influent concentration of 38.1 g chemical oxygen demand per litre, an instability of the reactor was observed. The results indicated that the up-flow anaerobic sludge blanket reactor process could treat cheese whey effectively.

  8. Preliminary lifetime predictions for 304 stainless steel as the LANL ABC blanket material

    International Nuclear Information System (INIS)

    Park, J.J.; Buksa, J.J.; Houts, M.G.; Arthur, E.D.

    1997-11-01

    The prediction of materials lifetime in the preconceptual Los Alamos National Laboratory (LANL) Accelerator-Based Conversion of Plutonium (ABC) is of utmost interest. Because Hastelloy N showed good corrosion resistance to the Oak Ridge National Laboratory Molten Salt Reactor Experiment fuel salt that is similar to the LANL ABC fuel salt, Hastelloy N was originally proposed for the LANL ABC blanket material. In this paper, the possibility of using 304 stainless steel as a replacement for the Hastelloy N is investigated in terms of corrosion issues and fluence-limit considerations. An attempt is made, based on the previous Fast Flux Test Facility design data, to predict the preliminary lifetime estimate of the 304 stainless steel used in the blanket region of the LANL ABC

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Werner, R.W.

    1977-01-01

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

  11. Intercomparison of liquid metal fast reactor seismic analysis codes. V. 2: Verification and improvement of reactor core seismic analysis codes using core mock-up experiments. Proceedings of a research co-ordination meeting held in Vienna, 26-28 September 1994

    International Nuclear Information System (INIS)

    1995-10-01

    This report (Volume II) contains the papers summarizing the verification of and improvement to the codes on the basis of the French and Japanese data. Volume I: ''Validation of the Seismic Analysis Codes Using the Reactor Code Experiments'' (IAEA-TECDOC-798) included the Italian PEC reactor data. Refs, figs and tabs

  12. Intercomparison of liquid metal fast reactor seismic analysis codes. V. 2: Verification and improvement of reactor core seismic analysis codes using core mock-up experiments. Proceedings of a research co-ordination meeting held in Vienna, 26-28 September 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-01

    This report (Volume II) contains the papers summarizing the verification of and improvement to the codes on the basis of the French and Japanese data. Volume I: ``Validation of the Seismic Analysis Codes Using the Reactor Code Experiments`` (IAEA-TECDOC-798) included the Italian PEC reactor data. Refs, figs and tabs.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  15. Full scale mock-up tests for rod bundle thermal-hydraulics in Japan

    International Nuclear Information System (INIS)

    Sugawara, S.

    1995-01-01

    This poster describes tests aimed at development and validation of principal design methodology of rod bundle thermal-hydraulics correlations. The works are based on domestic data base using the full-scale mock-up test facilities. The scope of the tests comprises DNB heat flux, transient DNB heat flux, post DNB heat transfer, pressure drop and void distribution. The works have been performed under collaboration among electric facilities, NPP vendors, universities, governmental corporations. 1 tab., 14 figs

  16. Ultrasonic testing results of fatigue cracks in PWR mock-up

    International Nuclear Information System (INIS)

    Gondard, C.

    1990-01-01

    The Ispra Joint Research Center has entered, since many years a study on fatigue crack propagation in PWR reactor vessels. The objective of this study is to establish a relation between the size and the location of defects and the lifetime of the vessel. For verifying the theoretical models validity a mockup has been built. This document gives the results of CEA for 6 in service inspection during 5 years [fr

  17. Design, fabrication, and mockup testing in the Remote Maintenance Development Facility

    International Nuclear Information System (INIS)

    Carter, J.A.; Jacobs, R.T.; Bingham, G.E.

    1978-01-01

    The Remote Maintenance Development Facility (RMDF) at the Idaho National Engineering Laboratory (INEL) was installed and used extensively for full-scale development, mockup and testing of remote maintenance requirements for the New Waste Calcining Facility (NWCF). By performing remote handling tests, the NWCF handling concepts, techniques and remote capabilities were proven workable prior to construction. Presented in this paper is a description of the RMDF and its purpose, functions, and handling capabilities as they were used in support of the NWCF

  18. Design, fabrication, and mockup testing in the remote maintenance development facility

    International Nuclear Information System (INIS)

    Carter, J.A.; Jacobs, R.T.; Bingham, G.E.

    1978-01-01

    The Remote Maintenance Development Facility at the Idaho National Engineering Laboratory was installed and used extensively for full-scale development, mockup, and testing of remote maintenance requirements for the New Waste Calcining Facility (NWCF). By performing remote handling tests, the NWCF handling concepts, techniques, and remote capabilities were proven workable prior to construction. A description of the RMDF and its purpose, functions, and handling capabilities as they were used in support of the NWCF is presented

  19. ASTP crewmen in Soyuz orbital module mock-up during training session at JSC

    Science.gov (United States)

    1975-01-01

    An interior view of the Soyuz orbital module mock-up in bldg 35 during Apollo Soyuz Test Project (ASTP) joint crew training at JSC. The ASTP crewmen are Astronaut Vance D. Brand (on left), command module pilot of the American ASTP prime crew; and Cosmonaut Valeriy N. Kubasov, engineer on the Soviet ASTP first (prime) crew. The training session simulated activities on the second day in Earth orbit.

  20. Mock-up test results of monoblock-type CFC divertor armor for JT-60SA

    Energy Technology Data Exchange (ETDEWEB)

    Higashijima, S. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)], E-mail: higashijima.satoru@jaea.go.jp; Sakurai, S.; Suzuki, S.; Yokoyama, K.; Kashiwa, Y.; Masaki, K.; Shibama, Y.K.; Takechi, M.; Shibanuma, K.; Sakasai, A.; Matsukawa, M.; Kikuchi, M. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

    2009-06-15

    The JT-60 Super Advanced (JT-60SA) tokamak project starts under both the Japanese domestic program and the international program 'Broader Approach'. The maximum heat flux to JT-60SA divertor is estimated to {approx}15 MW/m{sup 2} for 100 s. Japan Atomic Energy Agency (JAEA) has developed a divertor armor facing high heat flux in the engineering R and D for ITER, and it is concluded that monoblock-type CFC divertor armor is promising for JT-60SA. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin oxygen-free high conductivity copper (OFHC-Cu) buffer layer between the CFC monoblock and the screw-tube. CFC/OFHC-Cu and OFHC-Cu/CuCrZr joints are essential for the armor, and these interfaces are brazed. Needed improvements from ITER engineering R and D are good CFC/OFHC-Cu and OFHC-Cu/CuCrZr interfaces and suppression of CFC cracking. For these purposes, metalization inside CFC monoblock is applied, and we confirmed again that the mock-up has heat removal capability in excess of ITER requirement. For optimization of the fabrication method and understanding of the production yield, the mock-ups corresponding to quantity produced in one furnace at the same time is also produced, and the half of the mock-ups could remove 15 MW/m{sup 2} as required. This paper summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor.

  1. Astronauts McMonagle and Brown on flight deck mockup during training

    Science.gov (United States)

    1994-01-01

    Astronauts Donald R. McMonagle, STS-66 mission commander, left, and Curtis L. Brown, STS-66 pilot, man the commander's and pilot's stations, respectively, during a rehearsal of procedures to be followed during the launch and entry phases of their scheduled November 1994 flight. This rehearsal, held in the crew compartment trainer (CCT) of JSC's Shuttle mockup and integration laboratory, was followed by a training session on emergency egress procedures.

  2. Manufacturing and testing of relevant scale mockup based on monoblock concept

    International Nuclear Information System (INIS)

    Di Pietro, E.; Orsini, A.; Sacchetti, M.; Libera, S.; Cardella, A.; Vieider, G.

    1993-01-01

    The results obtained from small-scale mockups manufactured on the monoblock design concept have proven that the solution appears promising for a conventional divertor operating with heat fluxes in the range 10 to 15 MW/m 2 with a thermal fatigue cycle exceeding 1000 cycles at full power. The divertor mock-up consists of six half meter-long armored tubes obtained by brazing CFC to TZM molybdenum alloy. Two types of CFC were used to investigate the advantages of 3-d CFCs with respect to more conventional and cheaper 2-d CFC. The brazing process utilizes three variants of a process developed in laboratory trials and based on selected combinations of active braze filler/CFC surface conditioning procedures. The supporting structure is based on the sliding support concept intended to assure a compromise between the requested thermal stability of the component and the buildup of secondary stresses deriving from mechanical constraints. The FE thermal and thermal mechanical analysis of the divertor mockup structure is reported and the critical areas of sliding support are highlighted for comparison with experimental results. The main results of NDE and experimental high heat flux tests are reported and discussed

  3. Solid State Track Recorder fission rate measurements in low power light water reactor pressure vessel mockups

    International Nuclear Information System (INIS)

    Ruddy, F.H.; Roberts, J.H.; Kellogg, L.S.

    1985-01-01

    The results of extensive SSTR measurements made at the Pool Critical Assembly (PCA) facility at Oak Ridge National Laboratory have been reported previously. Measurements were made at key locations in PCA which is an idealized mockup of the water gap, thermal shield, pressure vessel geometry of a light water reactor. Recently, additional SSTR fission rate measurements have been carried out for 237-Np, 238-U, and 235-U in key locations in the NESTOR Shielding and Dosimetry Improvement Program (NESDIP) mockup facility located at Winfrith, England. NESDIP is a replica of the PCA facility, and comparisons will be made between PCA and NESDIP measurements. The results of measurements made at the engineering mockup at the VENUS critical assembly at CEN/SCK, Mol, Belgium will also be reported. Measurements were made at selected radial and azimuthal locations in VENUS, which models the in-core and near-core regions of a pressurized water reactor. Comparisons of absolute SSTR fission rates with absolute fission rates made with the Mol miniature fission chamber will be reported. Absolute fission rate comparisons have also been made between the NBS fission chamber, radiometric fission foils, and SSTRs, and these results will be summarized

  4. Performance test results of mock-up test facility of HTTR hydrogen production system

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo

    2004-01-01

    For the purpose to demonstrate effectiveness of high-temperature nuclear heat utilization, Japan Atomic Energy Research Institute has been developing a hydrogen production system and has planned to connect the hydrogen production system to High Temperature Engineering Test Reactor (HTTR). Prior to construction of a HTTR hydrogen production system, a mock-up test facility was constructed to investigate transient behavior of the hydrogen production system and to establish system controllability. The Mock-up test facility with a full-scale reaction tube is an approximately 1/30-scale model of the HTTR hydrogen production system and an electric heater is used as a heat source instead of a reactor. After its construction, a performance test of the test facility was carried out in the same pressure and temperature conditions as those of the HTTR hydrogen production system to investigate its performance such as hydrogen production ability, controllability and so on. It was confirmed that hydrogen was stably produced with a hot helium gas about 120m 3 /h, which satisfy the design value, and thermal disturbance of helium gas during the start-up could be mitigated within the design value by using a steam generator. The mock-up test of the HTTR hydrogen production system using this facility will continue until 2004. (author)

  5. Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

  6. In-Pile thermal fatigue of First Wall mock-ups under ITER relevant conditions

    International Nuclear Information System (INIS)

    Blom, F.; Schmalz, F.; Kamer, S.; Ketema, D.J.

    2006-01-01

    The objective of this study is to perform in-pile thermal fatigue testing of three actively cooled First Wall (FW) mock-ups to check the effect of neutron irradiation on the Be/CuCrZr joints under representative FW operation conditions. Three FW mock-ups with Beryllium armor tiles will be neutron irradiated at 1 dpa (in Be) with parallel thermal fatigue testing for 30,000 cycles. The temperatures, stress distributions and stress amplitudes at the Be/CuCrZr interface of the mock-ups will be as close as possible to the values calculated for ITER FW panels. For this objective the PWM mocks-up subjected to thermal fatigue will be integrated with high density (W) plates on the Be-side to provide heat flux by nuclear heating. The assembly will be placed in the pool-side facility of the HFR and thermal cycling is then arranged by mechanical movement towards and from the core box. As the thermal design of the irradiation rig is very critical a pilot-irradiation will be performed to cross check the models used in the thermal design of the rig. The project is currently in the design phase of both the pilot and actual irradiation rig. The irradiation of the actual rig is planned to start at mid 2007 and last for two years. (author)

  7. Solid-state track recorder neutron dosimetry in light water reactor pressure vessel surveillance mockups

    International Nuclear Information System (INIS)

    Ruddy, F.H.; Roberts, J.H.; Gold, R.; Preston, C.C.

    1984-09-01

    Solid-State Track Recorder (SSTR) measurements of neutron-induced fission rates have been made in several pressure vessel mockup facilities as part of the US Nuclear Regulatory Commission's (NRC) Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP). The results of extensive physics-dosimetry measurements made at the Pool Critical Assembly (PCA) at Oak Ridge National Laboratory (ORNL) in Oak Ridge, TN are summarized. Included are 235 U, 238 U, 237 Np and 232 Th fission rates in the PCA 12/13, 8/7, and 4/12 SSC configurations. Additional low power measurements have been made in an engineering mockup at the VENUS critical assembly at CEN-SCK, Mol, Belgium. 237 Np and 238 U fission rates were made at selected locations in the VENUS mockup, which models the in-core and near-core regions of a pressurized water reactor (PWR). Absolute core power measurements were made at VENUS by exposing solid-state track recorders (SSTRs) to polished fuel pellets within in-core fuel pins. 8 references, 4 figures, 10 tables

  8. Basic data for making a mockup of the bronchus(I)

    International Nuclear Information System (INIS)

    Kim, Heung Tai; Kwon, Dal Gwan

    1984-01-01

    This is presented as basic data for making a mockup of the bronchus. Tubercluosis is an infectious disease designated by law and efforts to conquer it are exerted nationally. Whether the patient is infected or not can be diagnosed by photo fluorography intended for groups and also by Radio graphing which enables people to observe the film which is as long as the object. It is important to take photographs for the purpose of compacting as much information as possible in a film. But it is more important to gather information about the living bodies in the photographed film. Therefore, Knowledge of anatomy is most important for the accurate diagnosis of tuberculosis. Blood vessels of the lungs and the bronchus have a very complicated structure, so that they require efforts to understand them. To make a mockup of the bronchus is needed for better understanding of the complicated structure of the blood vessels in the lungs and the bronchus. I wrote this because I thought making a mockup is a foundation to understand the anatomy of the lungs

  9. Beryllium mock-ups development and ultrasonic testing for ITER divertor conditions

    International Nuclear Information System (INIS)

    Barabash, V.R.; Bykov, V.A.; Giniyatulin, R.N.; Gervash, A.A.; Gurieva, T.M.; Egorov, K.E.; Komarov, V.L.; Korolkov, M.D.; Mazul, I.V.; Gitarsky, L.S.; Strulia, I.L.; Sizenev, V.S.; Pronyakin, V.T.

    1995-01-01

    At the present time beryllium is considered as the most suitable armour material for the ITER divertor application. Different types of Be-divertor mock-up construction are compared in the report. Two different technologies of beryllium tiles joining to a heat sink body are analysed: high temperature brazing and thermodiffusion bonding. The comparative analysis of different constructions has been performed on the basis of 2-D finite element calculation for temperatures and stresses. The main parameters and diagnostic capabilities of electron beam facility for HHF testing of beryllium mock-ups are described. The first results of HHF tests of ''beryllium-copper saddle-MAGT tube'' and ''beryllium-copper plate-SS body'' mock-ups are presented. The reasons of the damages during the HHF are analysed. The technique of ultrasonic testing of the thermodifussion bonding and brazing quality for beryllium-copper joints is presented. The recorded results are prepared in the form of ultrasound grams. The testing results are compared with the metallographic analysis. (orig.)

  10. Exploring climatic controls on blanket bog litter decomposition across an altitudinal gradient

    Science.gov (United States)

    Bell, Michael; Ritson, Jonathan P.; Clark, Joanna M.; Verhoef, Anne; Brazier, Richard E.

    2016-04-01

    The hydrological and ecological functioning of blanket bogs is strongly coupled, involving multiple ecohydrological feedbacks which can affect carbon cycling. Cool and wet conditions inhibit decomposition, and favour the growth of Sphagnum mosses which produce highly recalcitrant litter. A small but persistent imbalance between production and decomposition has led to blanket bogs in the UK accumulating large amounts of carbon. Additionally, healthy bogs provide a suite of other ecosystems services including water regulation and drinking water provision. However, there is concern that climate change could increase rates of litter decomposition and disrupt this carbon sink. Furthermore, it has been argued that the response of these ecosystems in the warmer south west and west of the UK may provide an early analogue for later changes in the more extensive northern peatlands. In order to investigate the effects of climate change on blanket bog litter decomposition, we set-up a litter bag experiment across an altitudinal gradient spanning 200 m of elevation (including a transition from moorland to healthy blanket bog) on Dartmoor, an area of hitherto unstudied, climatically marginal blanket bog in the south west of the UK. At seven sites, water table depth and soil and surface temperature were recorded continuously. Litter bags filled with the litter of three vegetation species dominant on Dartmoor were incubated just below the bog surface and retrieved over a period of 12 months. We found significant differences in the rate of decomposition between species. At all sites, decomposition progressed in the order Calluna vulgaris (dwarf shrub) > Molinia caerulea (graminoid) > Sphagnum (bryophyte). However, while soil temperature did decrease along the altitudinal gradient, being warmer in the lower altitudes, a hypothesised accompanying decrease in decomposition rates did not occur. This could be explained by greater N deposition at the higher elevation sites (estimated

  11. Development work for the manufacture of a blanket shield prototype for ITER

    International Nuclear Information System (INIS)

    Boireau, B.; Boudot, C.; Cottin, A.; Lorenzetto, P.; Jacquinot, F.; Bucci, P.; Gillia, O.; Vidotto, F.

    2006-01-01

    In the frame of the blanket development for ITER, an R-and-D programme was implemented for the manufacture of a shield prototype by powder Hot Isostatic Pressing (HIPping). The shield consists of a Stainless steel forged block drilled and machined, at the back of which 3D bent tubes are HIPped inside a powder layer. This paper describes the development work through the manufacturing of several mock ups that leads us to be confident for the shield prototype manufacturing. The paper is divided into 2 parts, the first one related to the machining development and validation, the second one relating to the HIP development and validation. A partial full scale mock up for the machining development (machined PFSMU) was manufactured with no particular problems, all the main identified difficulties in machining like deep drilling and castellation machining where overcome and the mock up was conform to the specification. The manufacturing of a HIPped PFSMU for the HIP development was done after the manufacturing of smaller mock-ups each representing a particular detailed design point. A computer simulation work gave us some design recommendation, and the compared analysis of the numerical simulation and experimental results lead us to predict the distortions on the PFSMU HIPped mock up. The HIPped distortions that were the main uncertainty were assessed through small mock ups and bigger one. The mechanical characteristics of the joints are conform to the specification. Associated to the mock up manufacturing is the ultrasonic test development which consists in designing and manufacturing a miniaturized probe travelling inside the bent tube after the HIP cycle to examine the joint tube / powder among others. This ultrasonic development allowed the examination of the HIPped PFSMU mock up that concluded this development work. (author)

  12. Optimization of up-flow anaerobic sludge blanket reactor for ...

    African Journals Online (AJOL)

    Optimization of up-flow anaerobic sludge blanket reactor for treatment of composite ... AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... Granules grown in the bottom part of UASB reactor were more compact and tense ...

  13. Electromagnetic effects involving a tokamak reactor first wall and blanket

    International Nuclear Information System (INIS)

    Turner, L.R.; Evans, K. Jr.; Gelbard, E.; Prater, R.

    1980-01-01

    Four electromagnetic effects experienced by the first wall and blanket of a tokamak reactor are considered. First, the first wall provides reduction of the growth rate of vertical axisymmetric instability and stabilization of low mode number interval kink modes. Second, if a rapid plasma disruption occurs, a current will be induced on the first wall, tending to maintain the field formerly produced by the plasma. Third, correction of plasma movement can begin on a time scale much faster than the L/R time of the first wall and blanket. Fourth, field changes, especially those from plasma disruption or from rapid discharge of a toroidal field coil, can cause substantial eddy current forces on elements of the first wall and blanket. These effects are considered specifically for the first wall and blanket of the STARFIRE commercial reactor design study

  14. Blanket options for high-efficiency fusion power

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  15. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  16. Fusion blanket for high-efficiency power cycles

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  17. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-11-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  20. Aqueous self-cooled blanket concepts for fusion reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Dudziak, D.J.

    1977-09-01

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

  3. Application of vanadium alloys to a fusion reactor blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-05-01

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

  4. Failure analysis of PB-1 (EBTS Be/Cu mockup)

    International Nuclear Information System (INIS)

    Odegard, B.C. Jr.; Cadden, C.H.

    1996-11-01

    Failure analysis was done on PB-1 (series of Be tiles joined to Cu alloy) following a tile failure during a high heat flux experiment in EBTS (electron beam test system). This heat flux load simulated ambient conditions inside ITER; the Be tiles were bonded to the Cu alloy using low-temperature diffusion bonding, which is being considered for fabricating plasma facing components in ITER. Results showed differences between the EBTS failure and a failure during a room temperature tensile test. The latter occurred at the Cu-Be interface in an intermetallic phase formed by reaction of the two metals at the bonding temperature. Fracture strengths measured by these tests were over 300 MPa. The high heat flux specimens failed at the Cu-Cu diffusion bond. Fracture morphology in both cases was a mixed mode of dimple rupture and transgranular cleavage. Several explanations for this difference in failure mechanism are suggested

  5. Calculation of ex-core detector weighting functions for a sodium-cooled tru burner mockup using MCNP5

    International Nuclear Information System (INIS)

    Pham Nhu Viet Ha; Min Jae Lee; Sunghwan Yun; Sang Ji Kim

    2015-01-01

    Power regulation systems of fast reactors are based on the signals of excore detectors. The excore detector weighting functions, which establish correspondence between the core power distribution and detector signal, are very useful for detector response analyses, e.g., in rod drop experiments. This paper presents the calculation of the weighting functions for a TRU burner mockup of the Korean Prototype Generation-IV Sodium-cooled Fast Reactor (named BFS-76-1A) using the MCNP5 multi-group adjoint capability. For generation of the weighting functions, all fuel assemblies were considered and each of them was divided into ten horizontal layers. Then the weighting functions for individual fuel assembly horizontal layers, the assembly weighting functions, and the shape annealing functions at RCP (Reactor Critical Point) and at conditions under which a control rod group was fully inserted into the core while other control rods at RCP were determined and evaluated. The results indicate that the weighting functions can be considered relatively insensitive to the control rods position during the rod drop experiments and therefore those weighting values at RCP can be applied to the dynamic rod worth simulation for the BFS-76-1A. (author)

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    Romero, J.B.

    1980-03-01

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

  9. Main features and potentialities of gas-blanket systems

    International Nuclear Information System (INIS)

    Lehnert, B.

    1977-02-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  12. Overview of first wall/blanket/shield technology

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1983-04-01

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

  13. Applications of the Aqueous Self-Cooled Blanket concept

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  14. 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.

  15. Electrical connectors for blanket modules in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Poddubnyi, I., E-mail: poddubnyyii@nikiet.ru [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Khomiakov, S.; Kolganov, V. [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Sadakov, S.; Calcagno, B.; Chappuis, Ph.; Roccella, R.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Danilov, I.; Leshukov, A.; Strebkov, Y. [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Ulrickson, M. [Sandia National Laboratories MS-1129, PO Box 5800, Albuquerque, NM 87185 (United States)

    2014-10-15

    Highlights: • Analysis of static and cyclic strength for L-shaped and Z-shaped ES has been performed. • Analysis results do show that for L-shaped ES static and cyclic strength criteria are not satisfied. • Static and cyclic strength criteria are met well by ES with Z-shaped elastic elements. • ES with Z-shaped elastic elements has been adopted as a new baseline design for ITER. - Abstract: Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2 mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137 kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9 T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak. The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was

  16. Design and development of ceramic breeder demo blanket

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Science.gov (United States)

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

    2017-12-01

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

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

    International Nuclear Information System (INIS)

    Sako, K.; Minato, A.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Tanaka, Satoru

    1987-01-01

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

  20. Calculations of fission rate distribution in the core of WWER-1000 mock-up on the LR-0 reactor using alternative methods and comparison with results of measurements

    International Nuclear Information System (INIS)

    Zaritskiy, S.; Kovalishin, A.; Tsvetkov, T.; Rypar, V.; Svadlenkova, M.

    2011-01-01

    General review of experimental and calculation researches on WWER-440 and WWER-1000 mock-ups on the reactor LR-0 was introduced on the twentieth Symposium AER. The experimental core fission rate distribution was obtained by means of gamma-scanning of the fuel pins - 140La single peak (1596 keV) measurements and wide energy range (approximately 600-900 keV) measurements. Altogether from 260 to 500 fuel pins were scanned in different experiments. The measurements were arranged in the middle of the fuel (the active part of pin). Pin-to-pin calculations of the WWER-1000 mock-up core fission rate distribution were performed with several codes: Monte Carlo codes MCU-REA/2 and MCNPX with different nuclear data libraries, diffusion code RADAR (63 energy groups library) and code SVL based on Surface Harmonics Method (69 energy groups). Calculated data are compared with experimental ones. The obtained results allow developing the benchmark for core calculations methodologies, evaluating and validating source reliability for the out-of-core (inside and outside pressure vessel) neutron transport calculations. (Authors)

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Sahin, Sumer

    1982-01-01

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

  3. Robot vision system R and D for ITER blanket remote-handling system

    International Nuclear Information System (INIS)

    Maruyama, Takahito; Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Tesini, Alessandro

    2014-01-01

    For regular maintenance of the International Thermonuclear Experimental Reactor (ITER), a system called the ITER blanket remote-handling system is necessary to remotely handle the blanket modules because of the high levels of gamma radiation. Modules will be handled by robotic power manipulators and they must have a non-contact-sensing system for installing and grasping to avoid contact with other modules. A robot vision system that uses cameras was adopted for this non-contact-sensing system. Experiments for grasping modules were carried out in a dark room to simulate the environment inside the vacuum vessel and the robot vision system's measurement errors were studied. As a result, the accuracy of the manipulator's movements was within 2.01 mm and 0.31°, which satisfies the system requirements. Therefore, it was concluded that this robot vision system is suitable for the non-contact-sensing system of the ITER blanket remote-handling system

  4. Prospects of the aqueous self-cooled blanket concept for NET

    International Nuclear Information System (INIS)

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

    1989-01-01

    A low-technology Aqueous Self-Cooled Blanket (ASCB) concept has been proposed for the Next European Torus (NET). This concept relies on structural material and cooling water, with small amounts of lithium compounds for tritium production. Following preliminary investigations, LiOH, LiNO 3 , LiNO 2 and Li 2 SO 4 are currently under consideration as tritium breeding materials in solution. The concept may benefit from the proven technologies from the PWRs and from the CANDU tritium extraction systems. It combines good shielding and breeding capabilities. It would serve as a reliable environment for experimenting with several DEMOnstration reactor-relevant blanket modules in NET. Since net tritium breeding is not a design requirement for NET, sufficient tritium breeding can be obtained without the application of external neutron multipliers if enrichment in 6 Li is utilized. For a DEMOnstration reactor ASCB-based blanket, neutron multipliers have to be incorporated and temperature and pressure have to be increased. Radiolysis and corrosion aspects are of particular concern and need further investigation. (orig.)

  5. Robot vision system R and D for ITER blanket remote-handling system

    Energy Technology Data Exchange (ETDEWEB)

    Maruyama, Takahito, E-mail: maruyama.takahito@jaea.go.jp [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Tesini, Alessandro [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France)

    2014-10-15

    For regular maintenance of the International Thermonuclear Experimental Reactor (ITER), a system called the ITER blanket remote-handling system is necessary to remotely handle the blanket modules because of the high levels of gamma radiation. Modules will be handled by robotic power manipulators and they must have a non-contact-sensing system for installing and grasping to avoid contact with other modules. A robot vision system that uses cameras was adopted for this non-contact-sensing system. Experiments for grasping modules were carried out in a dark room to simulate the environment inside the vacuum vessel and the robot vision system's measurement errors were studied. As a result, the accuracy of the manipulator's movements was within 2.01 mm and 0.31°, which satisfies the system requirements. Therefore, it was concluded that this robot vision system is suitable for the non-contact-sensing system of the ITER blanket remote-handling system.

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

    International Nuclear Information System (INIS)

    Malang, S.; Smith, D.L.

    1984-04-01

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

  7. Engineering challenges and development of the ITER Blanket System and Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Merola, Mario, E-mail: mario.merola@iter.org; Escourbiac, Frederic; Raffray, Alphonse Rene; Chappuis, Philippe; Hirai, Takeshi; Gicquel, Stefan

    2015-10-15

    The ITER Blanket System and the Divertor are the main components which directly face the plasma. Being the first physical barrier to the plasma, they have very demanding design requirements, which include accommodating: (1) surface heat flux and neutronic volumetric heating, (2) electromagnetic loads, (3) nuclear shielding function, (4) capability of being assembled and remote-handled, (5) interfaces with other in-vessel components, and (6) high heat flux technologies and complex welded structures in the design. The main functions of the Blanket System have been substantially expanded and it has now also to provide limiting surfaces that define the plasma boundary during startup and shutdown. As regards the Divertor, the ITER Council decided in November 2013 to start the ITER operation with a full-tungsten armour in order to minimize costs and already gain operational experience with tungsten during the non-active phase of the machine. This paper gives an overview of the design and technology qualification of the Blanket System and the Divertor.

  8. Transmutation blanket design for a Tokamak system

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  9. Numerical analysis of heat transfer in the first wall of CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Deng, Weiping; Ge, Zhihao; Li, Yuanjie

    2016-04-15

    Highlights: • Detailed numerical analysis of heat transfer in a water-cooling first wall was carried out based on the conceptual design of CFETR WCSB blanket. • Investigation of the influences of buoyancy effect and surface roughness on heat transfer in the water-cooling first wall was presented. • Analysis of the effect of the front wall thickness on temperature was carried out for the water-cooling first wall design. • Simulation results of two 1D CFD methods were evaluated by the 3D CFD data. - Abstract: China Fusion Engineering Test Reactor (CFETR), the first fusion reactor experiment project planned in China, is now being investigated in detail. Recently, a conceptual structural design of the Water-Cooled-Solid-Breeder (WCSB) blanket was proposed as one of the breeding blanket candidates for CFETR. In this research, based on the present design of the CFETR WCSB blanket, the heat transfer performance in the first wall (FW) under the pressurized water cooling condition was analyzed. The 3D computational fluid dynamics (CFD) results show that the maximal temperature of the FW will not exceed the limited temperature under normal or even higher heat flux condition. In addition, the effect of buoyancy on heat transfer is negligible under both conditions. The influence of roughness becomes increasingly important when the roughness height lies in the fully turbulent regime. The maximal temperature increases approximately linearly as the thickness of the front wall increases. It is also found that the heat flux and the local heat transfer coefficient are extremely non-uniform in the circumferential direction. Two 1D CFD methods are also evaluated by 3D CFD data, with the conclusion that both 1D results have some differences with the 3D data. The improved 1D method is more accurate than the former one. However, we ascertain that 1D methods should be used with caution for the water-cooling FW design.

  10. Hydrogen permeation through Flinabe fluoride molten salts for blanket candidates

    Energy Technology Data Exchange (ETDEWEB)

    Nishiumi, Ryosuke, E-mail: r.nishiumi@aees.kyushu-u.ac.jp; Fukada, Satoshi; Nakamura, Akira; Katayama, Kazunari

    2016-11-01

    Highlights: • H{sub 2} diffusivity, solubility and permeability in Flinabe as T breeder are determined. • Effects in composition differences among Flibe, Fnabe and Flinabe are compared. • Changes of pressure dependence of Flinabe permeation rate are clarified. - Abstract: Fluoride molten salt Flibe (2LiF + BeF{sub 2}) is a promising candidate for the liquid blanket of a nuclear fusion reactor, because of its large advantages of tritium breeding ratio and heat-transfer fluid. Since its melting point is higher than other liquid candidates, another new fluoride molten salt Flinabe (LiF + NaF + BeF{sub 2}) is recently focused on because of its lower melting point while holding proper breeding properties. In this experiment, hydrogen permeation behavior through the three molten salts of Flibe (2LiF + BeF{sub 2}), Fnabe (NaF + BeF{sub 2}) and Flinabe are investigated in order to clarify the effects of their compositions on hydrogen transfer properties. After making up any of the three molten salts and purifying it using HF, hydrogen permeability, diffusivity and solubility of the molten salts are determined experimentally by using a system composed of tertiary cylindrical tubes. Close agreement is obtained between experimental data and analytical solutions. H{sub 2} permeability, diffusivity and solubility are correlated as a function of temperature and are compared among the three molten salts.

  11. Tokamak blanket design study: FY 78 summary report

    International Nuclear Information System (INIS)

    1979-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-12-01

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

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

    International Nuclear Information System (INIS)

    Enoeda, Mikio; Akiba, Masato; Ohara, Yoshihiro

    2001-12-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-01-01

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

  16. Thermal fatigue testing of a diffusion-bonded beryllium divertor mock-up under ITER relevant conditions

    International Nuclear Information System (INIS)

    Youchison, D.L.; Guiniiatouline, R.; Watson, R.D.

    1994-01-01

    Thermal response and thermal fatigue tests of four 5 mm thick beryllium tiles on a Russian divertor mock-up were completed on the Electron Beam Test System at Sandia National Laboratories. The beryllium tiles were diffusion bonded onto an OFHC copper saddleblock and a DSCu (MAGT) tube containing a porous coating. Thermal response tests were performed on the tiles to an absorbed heat flux of 5 MW/m 2 and surface temperatures near 300 degrees C using 1.4 MPa water at 5.0 m/s flow velocity and an inlet temperature of 8-15 degrees C. One tile was exposed to incrementally increasing heat fluxes up to 9.5 MW/m 2 and surface temperatures up to 690 degrees C before debonding at 10 MW/m 2 . A third tile debonded after 9200 thermal fatigue cycles at 5 MW/m 2 , while another debonded after 6800 cycles. In all cases, fatigue failure occurred in the intermetallic layers between the beryllium and copper. No fatigue cracking of the bulk beryllium was observed. During thermal cycling, a gradual loss of porous coating produced increasing sample temperatures. These experiments indicate that diffusion-bonded beryllium tiles can survive several thousand thermal cycles under ITER relevant conditions without failure. However, the reliability of the diffusion bonded Joint remains a serious issue

  17. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

    International Nuclear Information System (INIS)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-01-01

    Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning

  18. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-02-28

    Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning.

  19. NDE of explosion welded copper stainless steel first wall mock-up

    International Nuclear Information System (INIS)

    Taehtinen, S.; Kauppinen, P.; Jeskanen, H.; Lahdenperae, K.; Ehrnsten, U.

    1997-04-01

    The study showed that reflection type C-mode scanning acoustic microscope (C-SAM) and internal ultrasonic inspection (IRIS) equipment can be applied for ultrasonic examination of copper stainless steel compound structures of ITER first wall mock-ups. Explosive welding can be applied to manufacture fully bonded copper stainless steel compound plates. However, explosives can be applied only for mechanical tightening of stainless steel cooling tubes within copper plate. If metallurgical bonding between stainless steel tubes and copper plate is required Hot Isostatic Pressing (HIP) method can be applied. (orig.)

  20. Application of virtual reality tools for assembly of WEST components: Comparison between simulations and physical mockups

    International Nuclear Information System (INIS)

    Pilia, Arnaud; Brun, Cyril; Doceul, Louis; Gargiulo, Laurent; Hatchressian, Jean-Claude; Keller, Delphine; Le, Roland; Poli, Serge; Zago, Bertrand

    2015-01-01

    Highlights: • VR technologies applied to Fusion enable better and faster understanding of integration issues. • Assembly problems are solved and validated on a numerical mockup. • Integration and accessibility issues can be identified in the earliest design on numerical mockup. • Problems are solved and validated on a physical mockup. • VR technologies are very helpful for assembly and maintenance operation simulations. - Abstract: The WEST project (Tungsten (W) Environment in Steady state Tokamak) is an upgrade of the existing fusion machine, Tore Supra. The goal is to equip the tokamak with a fully cooled tungsten divertor and to transform the machine in a test platform open to all ITER partners. The main assembly challenge of this project consists of an implementation of two magnet systems, called divertors, with an accuracy of 1 mm. Indeed, each divertor has about 4 m as diameter and has a heavy weight of 10 tons; also it introduces piece by piece in the original vessel through tight ports then assembled inside. To ensure a perfect fitting between these new components and a very constrained environment, it is necessary to use the latest CAD technologies available. Beyond conventional CAD tools, the virtual reality (VR) room of the institute provides several useful tools. Thanks to the 185″ stereoscopic 3D screen and a force feedback arm linked to clash detection software developed by the CEA LIST, a new way to carry out design and assembly studies was performed. In order to improve VR results, metrology data (3D scan) enhance simulations. Therefore, it becomes possible to be aware of the real size of a component and future difficulties in assembling it. At last, performance of such simulations is evaluated and compared to physical mockup in order to bring enhancement to the VR tools, before to be compared to the real operations on Tore Supra. The aim is to build a design tool that helps the designer since early stage of the design of complex systems