WorldWideScience

Sample records for blankets gas

  1. Divertor and gas blanket impurity control study

    International Nuclear Information System (INIS)

    A simple calculational model for the transport of particles across the scrap off region between the plasma and the wall in the presence of a divertor or a gas blanket has been developed. The model departs from previous work in including: (a) the entire impurity transport as well as its effect on the energy balance equations; (b) the recycling neutrals from the divertor, and (c) the reflected neutrals from the wall. Results obtained with this model show how the steady state impurity level in the plasma depends on the divertor parameters such as the neutral backflow from the divertor, the particle residence time and the scrape off thickness; and on the gas blanket parameters such as the neutral source strength and the gas blanket thickness. The variation of the divertor or gas blanket performance as a function of the heat and particle fluxes escaping from the plasma, the wall material and the cross field diffusion is examined and numerical examples are given

  2. 76 FR 2093 - Eni USA Gas Marketing LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2011-01-12

    ... Gas Marketing LLC; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY... November 30, 2010, by Eni USA Gas Marketing LLC (Eni USA), requesting blanket authorization to export... purchasing and marketing supplies of natural gas and LNG. Eni USA is a customer of the Cameron Terminal...

  3. Neutral gas blanket effects in a gaseous divertor

    International Nuclear Information System (INIS)

    The gaseous divertor employs a neutral gas blanket to absorb the plasma heat flux in the divertor chamber. This novel method for resolving the heat loading problem in a conventional divertor system is simulated experimentally. In our operational range (nsub(e) 13 cm-3, Tsub(e) <= 5 eV) it is demonstrated that the localized plasma heat flux is scattered relatively uniformly with neutral pressures of a few microns. At large neutral pressures the plasma stream is neutralized without touching a material wall. Plasma pumping inhibits neutral backflow and can sustain a neutral pressure difference comparable to the plasma pressure. Effective divertor channel conductance is measured to be reduced by a factor of six. (orig.)

  4. Main features and potentialities of gas-blanket systems

    International Nuclear Information System (INIS)

    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)

  5. Growth rates of interchange modes in the dense z-pinch and gas blanket effects

    International Nuclear Information System (INIS)

    A class of models of the dense z-pinch in which the bulk of the plasma is neutrally stable against interchange is studied. At a radius a the plasma is abruptly terminated and surrounded by a gas blanket. Instability results from the surface current necessary to maintain pressure balance at the plasma-gas interface. A single dispersion relation applies to all models in this class and shows two effects of the gas blanket in reducing growth rate. (1) Partial support of plasma pressure by external gas pressure reduces the driving force for the instability, and (2) mass loading of the relatively dense neutral gas reduces growth rate

  6. 76 FR 18216 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2011-04-01

    ... Federal Energy Regulatory Commission Southern Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on March 16, 2011, Southern Natural Gas Company (Southern), Post Office Box 2563... and 157.216 of the Commission's Regulations under the Natural Gas Act (NGA) as amended, to abandon...

  7. 75 FR 3232 - Northern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-01-20

    ... Energy Regulatory Commission Northern Natural Gas Company; Notice of Request Under Blanket Authorization January 8, 2010. Take notice that on December 30, 2009, Northern Natural Gas Company (Northern), 1111... sections 157.205 and 157.214 of the Commission's regulations under the Natural Gas Act for authorization...

  8. 75 FR 13535 - Northern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-03-22

    ... Energy Regulatory Commission Northern Natural Gas Company; Notice of Request Under Blanket Authorization March 16, 2010. Take notice that on March 12, 2010, Northern Natural Gas Company (Northern), 1111 South... External Affairs, Northern Natural Gas Company, 1111 South 103rd Street, Omaha, Nebraska 68124, at...

  9. 77 FR 31004 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2012-05-24

    ... Energy Regulatory Commission Southern Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on May 9, 2012, Southern Natural Gas Company (Southern), 569 Brookwood Village, Suite....210 of the Federal Energy Regulatory Commission's regulations under the Natural Gas Act (NGA),...

  10. 78 FR 4400 - Eni USA Gas Marketing LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2013-01-22

    ... include LNG in 10 CFR 590.102(i). \\3\\ Eni USA Gas Marketing referenced 49 FR 6684, February 22, 1984. Eni... blanket authorization recently granted to The Dow Chemical Company (DOE/FE Order No. 3162) on October 11... domestic demand for natural gas.\\4\\ \\4\\ The Dow Chemical Company, DOE/FE Order No. 3162 (October 11,...

  11. 78 FR 76827 - Midwestern Gas Transmission Company; Prior Notice of Activity Under Blanket Certificate

    Science.gov (United States)

    2013-12-19

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Midwestern Gas Transmission Company; Prior Notice of Activity Under Blanket Certificate On December 4, 2013, Midwestern Gas Transmission Company (Midwestern) filed with the Federal Energy Regulatory Commission...

  12. Blanketing effect of expansion foam on liquefied natural gas (LNG) spillage pool

    International Nuclear Information System (INIS)

    Highlights: • Reveal the existence of blocking effect of high expansion foam on an LNG pool. • Study the blanketing effect of high expansion foam quantitatively. • Correlate heat flux for vaporization with foam breaking rate. • Propose the physical mechanism of blanketing effect. - Abstract: With increasing consumption of natural gas, the safety of liquefied natural gas (LNG) utilization has become an issue that requires a comprehensive study on the risk of LNG spillage in facilities with mitigation measures. The immediate hazard associated with an LNG spill is the vapor hazard, i.e., a flammable vapor cloud at the ground level, due to rapid vaporization and dense gas behavior. It was believed that high expansion foam mitigated LNG vapor hazard through warming effect (raising vapor buoyancy), but the boil-off effect increased vaporization rate due to the heat from water drainage of foam. This work reveals the existence of blocking effect (blocking convection and radiation to the pool) to reduce vaporization rate. The blanketing effect on source term (vaporization rate) is a combination of boil-off and blocking effect, which was quantitatively studied through seven tests conducted in a wind tunnel with liquid nitrogen. Since the blocking effect reduces more heat to the pool than the boil-off effect adds, the blanketing effect contributes to the net reduction of heat convection and radiation to the pool by 70%. Water drainage rate of high expansion foam is essential to determine the effectiveness of blanketing effect, since water provides the boil-off effect

  13. 75 FR 38092 - The Dow Chemical Company; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-07-01

    ... The Dow Chemical Company (Dow), requesting blanket authorization to export liquefied natural gas (LNG... with its principal place of business in Midland, Michigan. Dow is an international chemical and...\\ The Dow Chemical Company, DOE/FE Order No. 2754 issued February 25, 2010. Current Application In...

  14. 75 FR 2140 - Florida Gas Transmission Company, LLC; Notice of Staff Protest to Proposed Blanket Certificate...

    Science.gov (United States)

    2010-01-14

    ... Energy Regulatory Commission Florida Gas Transmission Company, LLC; Notice of Staff Protest to Proposed Blanket Certificate Activity January 7, 2010. Commission staff (Protestor) hereby protests the prior..., subpart A, of the Commission's regulations.\\1\\ \\1\\ Section 157.205(f) provides that a protested...

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

    International Nuclear Information System (INIS)

    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)

  16. 75 FR 38093 - ConocoPhillips Alaska Natural Gas Corporation and Marathon Oil Company; Application for Blanket...

    Science.gov (United States)

    2010-07-01

    ... will not jeopardize service to the local markets into which this natural gas might otherwise be sold... critical back-up natural gas supply service for the local market in times of peak needs on the coldest days...Phillips Alaska Natural Gas Corporation and Marathon Oil Company; Application for Blanket Authorization...

  17. Molecule-surface interaction processes of relevance to gas blanket type fusion device divertor design

    Energy Technology Data Exchange (ETDEWEB)

    Snowdon, K.J. [Newcastle Univ. (United Kingdom). Dept. of Physics; Tawara, H.

    1997-01-01

    The mechanisms which may lead to the departure of molecular species from surfaces exposed to low energy (0.1-100 eV) particle or photon and electron irradiation are reviewed. Where possible, the charge and electronic state, angular, translational and internal energy distributions of the departing molecules are described and the physical origin of the nature of those distributions identified. The consequences, for the departing molecules, of certain material choices become apparent from such an analysis. Such information may help guide the choice of appropriate materials for plasma facing components of gas-blanket type divertors such as that recently proposed for the International Thermonuclear Experimental Reactor (ITER). (author). 71 refs.

  18. Detail Design of the hydrogen system and the gas blanketing system for the HANARO-CNS

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jung Woon; Kim, Hark Rho; Kim, Young Ki; Wu, Sang Ik; Kim, Bong Su; Lee, Yong Seop

    2007-04-15

    The cold neutron source (CNS), which will be installed in the vertical CN hole of the reflector tank at HANARO, makes thermal neutrons to moderate into the cold neutrons with the ranges of 0.1 {approx} 10 meV passing through a moderator at about 22K. A moderator to produce cold neutrons is liquid hydrogen, which liquefies by the heat transfer with cryogenic helium flowing from the helium refrigeration system (HRS). Because of its installed location, the hydrogen system is designed to be surrounded by the gas blanketing system to notify the leakage on the system and to prevent hydrogen leakage out of the CNS. The hydrogen system, consisted of hydrogen charging unit, hydrogen storage unit, hydrogen buffer tank, and hydrogen piping, is designed to smoothly and safely supply hydrogen to and to draw back hydrogen from the IPA of the CNS under the HRS operation mode. Described is that calculation for total required hydrogen amount in the CNS as well as operation schemes of the hydrogen system. The gas blanketing system (GBS) is designed for the supply of the compressed nitrogen gas into the air pressurized valves for the CNS, to isolate the hydrogen system from the air and the water, and to prevent air or water intrusion into the vacuum system as well as the hydrogen system. All detail descriptions are shown inhere as well as the operation scheme for the GBS.

  19. Tritium concentration monitoring of the purge gas stream of HCPB breeder blankets in future fusion reactors

    International Nuclear Information System (INIS)

    In fusion technology it is necessary to monitor tritiated gases for process monitoring. Such a system should be able to monitor the gas without taking samples. It should also be compact, cheap, the system stability should be excellent and it should recognize changes in the activity fast. Standard tools for activity measurements are ionization chambers and calorimeters. Ionization chambers work without sample taking but they are gas species dependent. Also pressures in the 100 mbar range are needed. Calorimeters are not suitable to be used as process monitors and it takes several hours to get a result. For activity measurements with a calorimeter it is necessary to extract gas samples. The Tritium Activity Chamber Experiment (TRACE) is a specially designed prototype to monitor traces of tritium in a gas sample utilizing Beta Induced X-Ray Spectroscopy (BIXS). Future fusion plants like ITER or DEMO could use such a system to monitor the purge gas streams in HCPB breeder blankets. TRACE will explore the possibility to monitor the expected 10 ppm tritium in the helium purge gas stream. We will evaluate if a BIXS system can be used as a standard monitoring system for tritiated gases in the range of (10-5-100) mbar tritium partial pressure.

  20. 75 FR 6021 - Florida Gas Transmission Company, LLC; Notice of Withdrawal of Staff Protest to Proposed Blanket...

    Science.gov (United States)

    2010-02-05

    ... of the request was published in the Federal Register on November 17, 2009 (74 FR 59,161). Kimberly D... Energy Regulatory Commission Florida Gas Transmission Company, LLC; Notice of Withdrawal of Staff Protest... withdraws its Protest to the Proposed Blanket Certificate Activity filed in the above-referenced...

  1. Gas permeability of SiC/SiC composite as blanket material of fusion reactor

    International Nuclear Information System (INIS)

    Gas permeability of SiC/SiC composite materials, which is one of the most important properties in application of SiC/SiC composite for first wall and blanket of fusion reactors, was measured by using a vacuum apparatus. The cylindrical SiC/SiC composite specimens were prepared by three different processes. The measurement on permeability for three materials was carried out with helium gas pressure ranging from 102 to 105 Pa at room temperature. The pressure in bottom chamber down stream of specimens increased with the helium gas pressure within the applied pressure range. The helium gas flow through the material is regarded as molecular flow. The material made by PIP method showed the highest permeability. The lowest permeability was observed in the one made by PIP followed by RS method. The material, SA-TyrannoHexTM made by hot pressing was in the second position. The difference of the permeability can be related with the macroscopic structure represented by pores and cracks. (author)

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

    International Nuclear Information System (INIS)

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

  3. Helium Gas Permeability of SiC/SiC Composite Developed for Blanket Component

    International Nuclear Information System (INIS)

    To employ SiC/SiC composite as blanket components of a fusion reactor, permeation behavior of helium gas has to be investigated since the helium is used as the coolant. For this purpose, a vacuum system consisting of upstream and downstream chambers was fabricated for the measurement of permeability, and the permeability was measured for several SiC/SiC composite materials recently developed. For the pressure range from 102 to 105 Pa in the upstream chamber, the pressure rise due to the permeation of helium in the downstream chamber linearly increased with the pressure of the upstream chamber. Then, the permeability was roughly constant for the pressure range of the upstream chamber.The permeabilities of SiC/SiC composites produced by polymer impregnation and pyrolysis (PIP), hot pressing (HP) and melt-infiltration-finished PIP were 5 x 10-5, 4 x 10-6, and 9 x 10-7 m2/s, respectively. In the matrix structure of the SiC/SiC composite made by the PIP method with a high permeability, cracking in the matrix and pores of micron size were observed. Compared to these materials, SiC/SiC composites produced by liquid phase sintering using submicron or nanopowder of β-SiC and the HP method had extremely low permeabilities. The permeability of the SiC/SiC composite made by using submicron or nanopowder of β-SiC became 1.5 x 10-9 or 4 x 10-11 m2/s.Based upon the present data, the helium gas flow was analyzed for a blanket module consisting of only SiC/SiC composite. If a vacuum pump is attached to the module, the helium leak into a plasma can be ignored, compared to the heliums produced by fusion reactions. Hence, the entire module can be made by only SiC/SiC composite, from a viewpoint of helium permeation

  4. The gas-cooled Li2O moderator/breeder canister blanket for fusion-synfuels

    International Nuclear Information System (INIS)

    A new integrated power and breeding blanket is described. The blanket incorporates features that make it suitable for synthetic fuel production. It is matched to the thermal and electrical requirements of the General Atomic water-splitting process for producing hydrogen. The fusion reaction is the Tandem Mirror Reactor (TMR) using Mirror Advanced Reactor Study (MARS) physics. The canister blanket is a high temperature, pressure balanced, crossflow heat exchanger contained within a low activity, independently cooled, moderate temperature, first wall structural envelope. The canister uses Li2O as the moderator/breeder and helium as the coolant. ''In situ'' tritium control, combined with slip stream processing and self-healing permeation barriers, assures a hydrogen product essentially free of tritium. The blanket is particularly adapted to synfuels production but is equally useful for electricity production or co-generation

  5. Design study on water- and gas-cooled outboard shield blankets for NET

    International Nuclear Information System (INIS)

    The Karlsruhe Nuclear Research Center entered into an agreement with the Commission of the European Communities on execution of development work geared to shielding blankets for NET. The concept to be investigated concerned water-cooled shielding blankets and, as a backup solution, a variation with helium cooling. The NET standard version as of late 1985 was considered as the basis of the investigations. The results of the study prepared in cooperation with the Sulzer company, Winterthur, and relating to the outboard blanket are contained in this report, which shows that it is relatively easy to fabricate water-cooled blankets. The stresses acting on the material during operation as a result of temperature gradients and coolant pressure are low. By addition of lithium salts to the coolant a great potential of tritium generation is offered. On the other hand, helium cooling is associated with some difficulties in design and with higher expenditure in fabrication. However, these difficulties can probably be overcome. (orig.)

  6. 75 FR 53966 - Kinder Morgan Interstate Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-09-02

    ... Interstate Gas Transmission, LLC (Kinder Morgan), 370 Van Gordon Street, Lakewood, Colorado 80228-8304 filed... Gordon Street, Lakewood, Colorado 80228-8304, or by calling (303) 763-3258 (telephone) or (303) 984-3272... Interstate Gas Transmission, LLC, 370 Van Gordon Street, Lakewood, Colorado 80228-8304, or by calling...

  7. Experimental study and analysis of the purge gas pressure drop across the pebble beds for the fusion HCPB blanket

    International Nuclear Information System (INIS)

    Highlights: ► The pressure drop significantly increases with decreasing the pebbles diameter. ► The pressure drop slightly increases with increasing the packing factor. ► The pressure drop is directly proportional to pebble bed length and inlet pressure. ► Predictions of Ergun equation agree well with the measured values of pressure drop. ► The filters resistance has a small contribution to the total pressure drop. -- Abstract: The lithium ceramic and beryllium pebble beds of the breeder units (BU), in the fusion breeding blanket, are purged by helium to extract the bred tritium. Therefore, the objective of this study is to support the design of the BU purge gas system by studying the effect of pebbles diameter, packing factor, pebble bed length, and flow inlet pressure on the purge gas pressure drop. The pebble bed was formed by packing glass pebbles in a rectangular container (56 mm × 206 mm × 396 mm) and was integrated into a gas loop to be purged by helium at BU-relevant pressures (1.1–3.8 bar). To determine the pressure drop across the pebble bed, the static pressure was measured at four locations along the pebble bed as well as at the inlet and outlet locations. The results show: (i) the pressure drop significantly increases with decreasing the pebbles diameter and slightly increases with increasing the packing factor, (ii) for a constant inlet flow velocity, the pressure drop is directly proportional to the pebble bed length and inlet pressure, and (iii) predictions of Ergun's equation agree well with the experimental values of the pressure drop

  8. Breeding blanket for DEMO

    International Nuclear Information System (INIS)

    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 being investigated within the framework of the European Test-Blanket Development Programme. (orig.)

  9. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

    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

  10. Materials for breeding blankets

    International Nuclear Information System (INIS)

    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

  11. Fissile fuel breeding in DT fusion reactor blankets

    International Nuclear Information System (INIS)

    Results of neutronic evaluations of fissile fuel breeding in a variety of DT fusion hybrid-reactor blankets are presented. The blankets are of the fast-fission or fission-suppressed rather than fission-enhanced designs, i.e. in the blankets considered emphasis is on fissile fuel rather than power production. For 233U breeding, when Li metal is the coolant for the first wall and the graphite moderator and the tritium breeding constituent of the blanket, the number of atoms of 233U produced per fusion in blankets that could be of practical interest is in the range 0.5 - 0.68, with the lower value applying to water-cooled ThO2 fertile fuel, the upper to gas-cooled Th-metal fuel located next to the reactor first wall. Neutron multipliers like Pb or Be can increase the production to about 0.74. For 239Pu breeding, the production ratio in practical blankets is 0.6 - 1.64, with the best results being for gas, Na- or Li-metal-cooled U-metal fuels located adjacent to the first wall (the U is depleted uranium). Gas-cooled U-Th-metal blankets, optimized for 233U breeding, yield 0.76 atoms of 233U and 0.38 atoms of 239Pu. The blanket energy multiplication factors are in the range 1.6 - 2.5 for Th blankets, 2.5 - 9.0 for U blankets and approximately 5.5 for the U-Th-metal blanket. The tritium breeding ratio in all blankets is 1.075. Blankets with other first wall, coolant and tritium breeding constituents are also considered. The fusion power requirements of hybrids that could supply the fuel needs of thorium-burning CANDU power reactors, and the allowed costs for building the hybrids are indicated

  12. Neutronics Assessment of Molten Salt Breeding Blanket Design Options

    International Nuclear Information System (INIS)

    Neutronics assessment has been performed for molten salt breeding blanket design options that can be utilized in fusion power plants. The concepts evaluated are a self-cooled Flinabe blanket with Be multiplier and dual-coolant blankets with He-cooled FW and structure. Three different molten salts were considered including the high melting point Flibe, a low melting point Flibe, and Flinabe. The same TBR can be achieved with a thinner self-cooled blanket compared to the dual-coolant blanket. A thicker Be zone is required in designs with Flinabe. The overall TBR will be ∼1.07 based on 3-D calculations without breeding in the divertor region. Using Be yields higher blanket energy multiplication than obtainable with Pb. A modest amount of tritium is produced in the Be (∼3 kg) over the blanket lifetime of ∼3 FPY. Using He gas in the dual-coolant blanket results in about a factor of 2 lower blanket shielding effectiveness. We show that it is possible to ensure that the shield is a lifetime component, the vacuum vessel is reweldable, and the magnets are adequately shielded. We conclude that molten salt blankets can be designed for fusion power plants with neutronics requirements such as adequate tritium breeding and shielding being satisfied

  13. Limitations on blanket performance

    International Nuclear Information System (INIS)

    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.)

  14. Solid breeder blanket concepts

    International Nuclear Information System (INIS)

    An investigation is made of a mechanical concept for the blanket with solid breeders in view of the possible adaptation to power reactor. A special arrangement of the multiplier and breeder materials is developed to permit a further neutronic optimisation

  15. Fast-core thermal-blanket breeder reactor

    International Nuclear Information System (INIS)

    A preliminary assessment of the performance expected from a specific type of FCTB reactor, consisting of a gas-cooled fast system for the core and natural-uranium light-water thermal system for the blanket is reported. Both the core and the blanket use the 238U-Pu fuel cycle. When all the neutrons leaking out of the core reach the blanket, the blanket-to-core power ratio is estimated to be about 1.3. By reducing its water-to-fuel volume ratio below 1.5, the light water blanket can be designed to have a higher ksub(eff), while maintaining an equilibrium fissile fuel content. Compared with conventional FBRs, having the same power output, the FCTB reactor considered offers the following advantages: a lower fissile fuel content, easier and safer control, no need for Pu separation. (B.G.)

  16. Tritium recovery from ceramic breeder blanket

    International Nuclear Information System (INIS)

    It is known that chemical forms of tritium released from ceramic breeders are T2O and T2. Among issues relevant to the tritium chemical form, tritium inventory is one of the major criteria in the selection of breeder material. The primary purpose of this report is to study the dependence of tritium inventory in a blanket with ceramic solid breeder on the tritium chemical form. In this light, tritium inventory in a Li2O blanket has been evaluated as a function of tritium chemical form under the conditions of the Japanese Fusion Experimental Reactor (FER). It was shown that in a blanket with Li2O as a breeder, which has a strong affinity to water vapor, the inventory due to T2O adsorption becomes quite large. In order to reduce the T2O adsorption inventory, conversion of the tritium chemical form through an isotope exchange reaction with hydrogen added to the sweep gas (T2O + 2 H2 → H2O + 2 HT) has been proposed, and its advantages and problems have been examined. Lithium hydroxide formation and mass transfer, which are considered to be inherent in the Li2O-T2O system and to be critical issues for the feasibility of a Li2O blanket, have been also discussed. (author)

  17. Review: BNL graphite blanket design concepts

    International Nuclear Information System (INIS)

    A review of the Brookhaven National Laboratory (BNL) minimum activity graphite blanket designs is made. Three designs are identified and discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a thick graphite screen (typically 30 cm or greater, depending on type as well as application-experimental power reactor or commercial reactor). 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. This energy is then either radiated to a secondary blanket with coolant tubes, as in types A and B, or is removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the structural material of the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude by the graphite screen, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma, whatever the degree of radiation damage

  18. Tailorable Advanced Blanket Insulation (TABI)

    Science.gov (United States)

    Sawko, Paul M.; Goldstein, Howard E.

    1987-01-01

    Single layer and multilayer insulating blankets for high-temperature service fabricated without sewing. TABI woven fabric made of aluminoborosilicate. Triangular-cross-section flutes of core filled with silica batting. Flexible blanket formed into curved shapes, providing high-temperature and high-heat-flux insulation.

  19. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

    The blanket of the present invention can keep the temperature of breeding materials within a predetermined range even if the breeding materials are consumed and the amount of heat generated from the breeding materials is reduced, thereby enabling to release tritium stably. That is, a neutron incident amount control means is disposed to the blanket for controlling the amount of neutrons incident to the breeding materials. Alternatively, a material to form hollow layers are disposed to the periphery of the breeding materials. With such constitution, the neutron incident amount control means enables to control the incident amount of neutrons from plasmas to the breeding materials, thereby enabling to suppress the change of the amount of heat generated in the breeding materials. In addition, the hollow layers formed at the periphery of the breeding materials enables selective filling of fluids having different heat transfer characteristics thereby enabling to control heat resistance between the breeding materials and cooling tubes. Accordingly, temperature of the breeding materials can be kept constant even in any of the cases. (I.S.)

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

    International Nuclear Information System (INIS)

    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.)

  1. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    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

  2. Blanket comparison and selection study. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    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. (MOW)

  3. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  5. Development of blanket remote maintenance system

    International Nuclear Information System (INIS)

    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)

  6. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

    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

  7. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

    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 usable 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

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

    International Nuclear Information System (INIS)

    Two different breeding blanket concepts are being elaborated at Forschungszentrum Karlsruhe within the framework of the DEMO breeding blanket development, the concept of a helium cooled solid breeding blanket and the concept of a self-cooled liquid metal blanket. The breeder material used in the first concept is Li4SiO4 as a pebble bed arranged separate from the beryllium pebble bed, which serves as multiplier. The breeder material zone is cooled by several toroidally-radially configurated helium cooling plates which, at the same time, act as reinforcements of the blanket structures. In the liquid metal blanket concept lead-lithium is used both as the breeder material and the coolant. It flows at low velocity in poloidal direction downwards and back in the blanket front zone. In both concepts the First Wall is cooled by helium gas. This report deals with the thermohydraulics design and thermomechanics analysis of the two blanket concepts. The performance data derived from the Monte-Carlo computations serve as a basis for the design calculations. The coolant inlet and outlet temperatures are chosen with the design criteria and the economics aspects taken into account. Uniform temperature distribution in the blanket structures can be achieved by suitable branching and routing of the coolant flows which contributes to reducing decisively the thermal stress. The computations were made using the ABAQUS computer code. The results obtained of the stresses have been evaluated using the ASME code. It can be demonstrated that all maximum values of temperature and stress are below the admissible limit. (orig.)

  9. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

    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

  10. Comparison analysis of fusion breeder blanket concepts

    International Nuclear Information System (INIS)

    Based on the wide survey, the development status and key issues of fusion breeder blanket concepts are summarized. Two types of blanket concepts, i.e. solid and liquid breeder blanket, were compared and assessed in terms of engineering feasibility, tritium recovery and control, economic and safety aspects, etc. The advantages and disadvantages of the two types of blanket concepts were clarified from the viewpoint of technology realization and development potential. This study may act as a valuable reference for fusion blanket concept selection and design. (authors)

  11. 78 FR 2990 - Bear Creek Storage Company, L.L.C.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-01-15

    ... Energy Regulatory Commission Bear Creek Storage Company, L.L.C.; Notice of Request Under Blanket Authorization Take notice that on December 21, 2012, Bear Creek Storage Company, L.L.C. (Bear Creek), 569... the Natural Gas Act, and Bear Creek's blanket certificate issued in Docket No. CP10-28-000 on...

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

    International Nuclear Information System (INIS)

    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)

  13. APT 3He target/blanket. Topical report

    International Nuclear Information System (INIS)

    The 3He target/blanket (T/B) preconceptual design for the 3/8-Goal facility is based on a 1000-MeV, 200-mA accelerator to produce a high-intensity proton beam that is expanded and then strikes one of two T/B modules. Each module consists of a centralized neutron source made of tungsten and lead, a proton beam backstop region made of zirconium and lead, and a moderator made of D2O. Helium-3 gas is circulated through the neutron source region and the blanket to create tritium through neutron capture. The gas is continually processed to extract the tritium with an online separation process

  14. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    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

  15. Compatibility problems in tritium breeding blankets

    International Nuclear Information System (INIS)

    Compatibility between tritium breeding materials (liquid or solid), neutron multiplier and structural steels is a concern for the choice of a tritium breeding blanket for NET. For solid tritium breeding blanket, it seems that the more severe compatibility problem is due to the interaction of beryllium with steel. As for the water-cooled Pb17Li blanket, the first results obtained in experimental conditions closed to the concept have evidenced lower corrosion rates than those measured in thermal convection loops

  16. Fusion blankets for high efficiency power cycles

    International Nuclear Information System (INIS)

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

  17. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

    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

  18. Fusion reactor blanket/shield design study

    International Nuclear Information System (INIS)

    A joint study of tokamak reactor first-wall/blanket/shield technology was conducted by Argonne National Laboratory (ANL) and McDonnell Douglas Astronautics Company (MDAC). The objectives of this program were the identification of key technological limitations for various tritium-breeding-blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium-breeding-blanket concepts were evaluated according to the proposed coolant. The ANL effort concentrated on evaluation of lithium- and water-cooled blanket designs while the MDAC effort focused on helium- and molten salt-cooled designs. A joint effort was undertaken to provide a consistent set of materials property data used for analysis of all blanket concepts. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first-wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented

  19. Test Blanket Module Pipe Forest integration in ITER equatorial port

    International Nuclear Information System (INIS)

    ITER Test Blanket Modules (TBMs) will allow testing Breeding Blanket concepts for a future application in DEMO. IRFM (Institut de Recherche sur la Fusion Magnetique) contribution to this test program consists in the integration of the 2 European TBMs (Helium Cooled Lithium Lead and Helium Cooled Pebble Bed) in a dedicated equatorial port. The two Breeding Blanket concepts use Helium gas as a coolant, liquid PbLi as breeder (for HCLL process) and Helium gas for Tritium extraction (for HCPB process). These materials are passing through the cryostat interspace forming a pipe network called the Pipe Forest. The main structural function of the Pipe Forest is to absorb the thermal expansion due to the Vacuum Vessel and due to the pipe system itself. The Pipe Forest has to cope with several design issues. In this study, the different key parameters of the Pipe Forest design are identified and their relative influence is analysed. Several design options were investigated and compared based on: -Thermo-mechanical finite element calculations -Pipe Forest integration within the cryostat interspace -Interface management -Assembly and maintenance scenarios -Complex pipe routing due to the expansion bends -RCC-MR 2007 requirements The chosen thermal compensation solution (thermal expansion loops) led to a Pipe Forest design. The CAE analysis of this Pipe Forest showed that it fulfills the requirements of the RCC-MR 2007, which is the reference design and construction code selected for the European TBM.

  20. GCFR radial blanket and shield experiment: objectives, preanalysis, and specifications

    International Nuclear Information System (INIS)

    An integral experiment has been designed for the verification of radiation transport methods and nuclear data used for the design of the radial shield for the proposed 300 MW(e) gas-cooled fast breeder reactor (GCFR). The scope of the experiment was chosen to include a thorium oxide radial blanket mockup as well as several shield configurations in order to reduce the uncertainties in the calculated source terms for the radial shield, and to reduce the uncertainties in the calculated radiation damage to the prestressed concrete reactor vessel (PCRV). Additionally, the measurements are intended to bound the uncertainties in calculated gamma-ray heating rates within the blanket and shield. Although designed specifically for the GCFR, the experiment will provide generic data regarding deep penetration in ThO2 and common shield materials, which should also benefit LMFBR designers

  1. ITER blanket, shield and material data base

    International Nuclear Information System (INIS)

    As part of the summary of the Conceptual Design Activities (CDA) for the International Thermonuclear Experimental Reactor (ITER), this document describes the ITER blanket, shield, and material data base. Part A, ''ITER Blanket and Shield Conceptual Design'', discusses the need for ITER of a tritium breeding blanket to supply most of the tritium for the fuel cycle of the device. Blanket and shield combined must be designed to operate at a neutron wall loading of 1MW/m2, and to provide adequate shielding of the magnets to meet the neutron energy fluence goal of 3MWa/m2 at the first wall. After a summary of the conceptual design, the following topics are elaborated upon: (1) function, design requirement, and critical issues; (2) material selection; (3) blanket and shield segmentation; (4) blanket design description; (5) design analysis; (6) shield; (7) radiation streaming analysis; and (8) a summary of benchmark calculations. Part B, ''ITER Materials Evaluation and Data Base'', treats the compilation and assessment of the available materials data base used for the selection of the appropriate materials for all major components of ITER, including (i) structural materials for the first wall, (ii) Tritium breeding materials for the blanket, (iii) plasma facing materials for the divertor and first wall armor, and (4) electric insulators for use in the blanket and divertor. Refs, figs and tabs

  2. Lithium-cooled blankets for advanced tokamaks

    International Nuclear Information System (INIS)

    The main objective of the Tokamak Power System Studies (TPSS) at Argonne National Lab. during fiscal year 1985 was to explore innovative design concepts that have the potential for significant enhancement of the attractiveness of a tokamak-based power plant. Activities in the area of plasma engineering resulted in a reference reactor concept, which served as a model for the impurity control and first-wall/blanket/shield studies. The liquid-metal-cooled first-wall/blanket/shield design activity was centered around the vanadium alloy structure and liquid-lithium coolant leading blanket concept as identified by the Blanket Comparison and Selection Study (BCSS). A ferritic steel structure and a LiPb breeder were considered as backup options. The magnetohydrodynamics (MHD) effects associated with self-cooled liquid-metal blanket/first-wall systems are substantially reduced by the lower magnetic fields required for higher plasmas, the lower neutron wall loading resulting from reduced power output, and the smaller reactor size of the TPSS model reactor. Therefore, improved performance characteristics of self-cooled liquid-metal blanket concepts are achievable mainly because the design constraints are more relaxed compared to the BCSS guidelines. Key aspects of the designs evaluated in the current study include the following: (1) design simplicity; (2) use of the first wall as an impurity control device; (3) modular first-wall/blanket/reflector/shield construction; and (4) integrated first-wall/blanket/reflector/shield

  3. Design and analysis of ITER shield blanket

    International Nuclear Information System (INIS)

    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.)

  4. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

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

  5. Disruption problematics in segmented-blanket concepts

    International Nuclear Information System (INIS)

    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 segmented-blanket design concepts in order to satisfy the disruption problematics.The present paper describes recent investigations on internal blanket reinforcement systems needed in order to improve the first-wall/blanket/shield structural design for next-step and commercial fusion reactors. Particularly in the context of SEAFP and ITER activities, representative 3-D CAD models of the inboard and outboard blanket regions and the related magnetomechanical simulations are illustrated. (orig.)

  6. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

    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....

  7. Development of Solid Breeder Blanket at JAERI

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been performing blanket development based on the long-term research program of fusion blankets in Japan, which was approved by the Fusion Council of Japan in 1999. The blanket development consists of out-pile R and D, In-pile R and D, TBM Neutronics and TPR Tests and Tritium Recovery System R and D. Based on the achievements of element technology development, the R and D program is now stepping to the engineering testing phase, in which scalable mockup tests will be performed for obtaining engineering data unique to the specific structure of the components, with the objective to define the fabrication specification of test blanket modules for ITER. This paper presents the major achievements of the element technology development of solid breeder blanket in JAERI

  8. Fast breeder reactor blanket management: comparison of LMFBR and GCFR blankets

    International Nuclear Information System (INIS)

    The economic performance of the fast breeder reactor blanket, considering different fuel management schemes was studied. To perform this, the investigation started with a standard reactor physics calculation. Then, two economic models for evaluation of the economic performance of the radial blanket were developed. These models formed the basis of a computer code, ECOBLAN, which computes the net economic gain and the levelized fuel cost due to the radial blanket. The net gain in terms of dollars and $/kgHM-y and the levelized fuel cost in mills/kWhe were obtained as a function of blanket thickness and a residence time of the fuel in the blanket. A LMFBR and a GCFR were the reactor models considered in this study. The optimum radial blanket of a GCFR consists of two rows, that of a LMFBR consists of three rows. Regarding the different fuel management schemes, the fixed blanket was found to be more favorable than reshuffled blanket. Out-in and in-out reshuffled blanket offer almost the same net gain. In all the cases, the burnup calculated for the fuel was found to be less than the acceptable limit. There is an optimum residence time for the fuel in the blanket which depends on the position of the fuel in the blanket and the fuel management scheme studied. As expected, except for very short residence times (less than 2.5 years), the radial blanket is a net income producer. There is no significant difference between the economic performance of the blanket of a LMFBR and a GCFR

  9. ITER reference breeding blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Ferrari, M. [ENEA, Frascati (Italy); Bianchi, A. [EFET, Ansaldo Ricerche, Genova (Italy); Celentano, G. [ENEA, ERG-FUS, Centro di Frascati, Via Enrico Fermi, 27, P.O. Box 65, I-00044, Frascati (IT)] [and others

    1999-11-01

    The ITER reference breeding blanket design is water-cooled and is characterised by the use of the neutronic multiplier and breeder materials in the form of pebbles. Besides the achievement, with margin, of the tritium breeding ratio (TBR) minimum requirement, it exhibits an internal layout allowing it to withstand properly electromagnetic loads during plasma disruption and vertical displacement events, and pressure loads in case of rupture of an internal cooling channel (i.e. in-box LOCA). During the first part of 1998, the design has been optimised improving the performance in terms of TBR, enlarging the design margins with respect to the dimensioning loads and investigating in detail the global behaviour of the system during normal and off-normal conditions. (orig.)

  10. ITER reference breeding blanket design

    International Nuclear Information System (INIS)

    The ITER reference breeding blanket design is water-cooled and is characterised by the use of the neutronic multiplier and breeder materials in the form of pebbles. Besides the achievement, with margin, of the tritium breeding ratio (TBR) minimum requirement, it exhibits an internal layout allowing it to withstand properly electromagnetic loads during plasma disruption and vertical displacement events, and pressure loads in case of rupture of an internal cooling channel (i.e. in-box LOCA). During the first part of 1998, the design has been optimised improving the performance in terms of TBR, enlarging the design margins with respect to the dimensioning loads and investigating in detail the global behaviour of the system during normal and off-normal conditions. (orig.)

  11. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

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

  12. Multivariable optimization of fusion reactor blankets

    International Nuclear Information System (INIS)

    The optimization problem consists of four key elements: a figure of merit for the reactor, a technique for estimating the neutronic performance of the blanket as a function of the design variables, constraints on the design variables and neutronic performance, and a method for optimizing the figure of merit subject to the constraints. The first reactor concept investigated uses a liquid lithium blanket for breeding tritium and a steel blanket to increase the fusion energy multiplication factor. The capital cost per unit of net electric power produced is minimized subject to constraints on the tritium breeding ratio and radiation damage rate. The optimal design has a 91-cm-thick lithium blanket denatured to 0.1% 6Li. The second reactor concept investigated uses a BeO neutron multiplier and a LiAlO2 breeding blanket. The total blanket thickness is minimized subject to constraints on the tritium breeding ratio, the total neutron leakage, and the heat generation rate in aluminum support tendons. The optimal design consists of a 4.2-cm-thick BeO multiplier and 42-cm-thick LiAlO2 breeding blanket enriched to 34% 6Li

  13. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

    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 m2) 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.)

  14. 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)

  15. ITER breeding blanket module design and analysis

    International Nuclear Information System (INIS)

    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)

  16. Mechanical and thermal design of hybrid blankets

    International Nuclear Information System (INIS)

    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

  17. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

    Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li17Pb83 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 Li17Pb83 blankets. (author)

  18. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

    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)

  19. Exploratory Study of Blanket Liquid Curtain

    Institute of Scientific and Technical Information of China (English)

    HUGang; HUANGJinhua; FENGKaiming

    2003-01-01

    Blankets and other in-vessel components are easily damaged owing to their circumstance of high radiation and high heat. To protect them, first wall design should be considered. Owing to its high heat removal nd self-refreshing capability, liquid metal first wall has been seen as a potential first wall for a fusion reactor in the future. Blanketliquid curtain is actually a special liquid metal wall to protect blanket.

  20. Advanced high performance solid wall blanket concepts

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    The European Union has been engaged since 1989 in a programme to develop tritium breeding blankets for application in a fusion power reactor. There are four concepts under development, namely two of the solid breeder type and two of the liquid breeder type. At the Forschungszentrum Karlsruhe one blanket concept of each line has been pursued so far with the so-called dual coolant type representing the liquid breeder line. In the dual coolant concept the breeder material (Pb-17Li) is circulated to external heat exchangers to carry away the bulk of the generated heat and to extract the tritium. Additionally, the heavily loaded first wall is cooled by high pressure helium gas. The safety and environmental impact of the dual coolant blanket concept has been assessed as part of the blanket concept selection excercise, a European concerted action, aiming at selecting the two most promising concepts for futher development. The topics investigated are: (a) Blanket materials and toxic materials inventory, (b) energy sources for mobilisation, (c) fault tolerance, (d) tritium and activation products release, and (e) waste generation and management. No insurmountable safety problems have been identified for the dual coolant blanket. The results of the assessment are described in this report. The information collected is also intended to serve as input to the EU 'Safety and Environmental Assessment of Fusion longterm Programme' (SEAL). The unresolved issues pertaining to the dual coolant blanket which would need further investigations in future programmes are outlined herein. (orig.)

  2. Nuclear, thermo-mechanical and tritium release analysis of ITER breeding blanket

    International Nuclear Information System (INIS)

    The design of the breeding blanket in ITER applies pebble bed breeder in tube (BIT) surrounded by multiplier pebble bed. It is assumed to use the same module support mechanism and coolant manifolds and coolant system as the shielding blankets. This work focuses on the verification of the design of the breeding blanket, from the viewpoints which is especially unique to the pebble bed type breeding blanket, such as, tritium breeding performance, tritium inventory and release behavior and thermo-mechanical performance of the ITER breeding blanket. With respect to the neutronics analysis, the detailed analyses of the distribution of the nuclear heating rate and TBR have been performed in 2D model using MCNP to clarify the input data for the tritium inventory and release rate analyses and thermo-mechanical analyses. With respect to the tritium inventory and release behavior analysis, the parametric analyses for selection of purge gas flow rate were carried out from the view point of pressure drop and the tritium inventory/release performance for Li2TiO3 breeder. The analysis result concluded that purge gas flow rate can be set to conventional flow rate setting (88 l/min per module) to 1/10 of that to save the purge gas flow and minimize the size of purge gas pipe. However, it is necessary to note that more tritium is transformed to HTO (chemical form of water) in case of Li2TiO3 compared to other breeder materials. With respect to the thermo-mechanical analyses of the pebble bed blanket structure, the analyses have been performed by ABAQUS with 2D model derived from one of eight facets of a blanket module, based on the reference design. Analyses were performed to identify the temperature distribution incorporating the pebble bed mechanical simulation and influence of mechanical behavior to the thermal behavior. The result showed that the maximum temperature in the breeding material was 617degC in the first row of breeding rods and the minimum temperature was 328degC in

  3. Blanket management method for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    A method for reducing thermal striping in liquid metal fast breeder reactors by reducing temperature gradients between adjacent fuel and blanket assemblies by shuffling blanket assemblies at each refueling outage so as to progressively shuffle the blanket assemblies to the core periphery through multiple moves and to generally locate fresh blanket assemblies adjacent to exposed fuel assemblies and exposed blanket assemblies adjacent to fresh fuel. Additionally, assembly orificing is altered to provide less flow to blanket assemblies needing less flow due to an otherwise decreased temperature gradient and providing additional flow to fuel assemblies which need more flow to sufficiently reduce temperature gradients to prevent thermal striping. (author)

  4. Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Zhu, Qingjun; Li, Jia; Liu, Songlin

    2016-07-01

    In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

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

    International Nuclear Information System (INIS)

    Test blanket module (TBM) is one of important components in ITER. Some of related blanket technologies of future fusion, such as tritium self-sufficiency, the exaction of high-grade heat, design criteria and safety requirements and environmental impacts, will be demonstrated in ITER-TBM. In ITER device, the three equatorial ports have allocated for TBM testing. China had proposed to develop independently the ITER-TBM with helium cooled solid breeder in 12th meeting of test blanket workgroup (TBWG-12). In this work, the preliminary design and analysis for Chinese HCSB TBM will be carried out. The TBM must be contains the function of the first wall, breeding blanket, shield and structure. Finally, in the period of preliminary investigation, HCSB TBM design adopt modularization concept which is helium as coolant and tritium purge gas, ferritic/martensitic steel as structural material, Lithium orthosilicate (Li4SiO4) as tritium breeder, beryllium pebble as neutron multiplier. TBM is allocated in standard vertical frame port. HCSB TBM consist of first wall, backplate, breeding sub-modules, caps, grid and support plate, and breeding sub-modules is arranged by layout of 2 x 6 in blanket box. In this paper, main components of HCSB TBM will be described in detail, also performance analysis of main components have been completed. (authors)

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

    International Nuclear Information System (INIS)

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

  7. Solid breeder blanket design and tritium breeding

    International Nuclear Information System (INIS)

    Thermonuclear D-T power plants will have to be tritium self-sufficient. In addition to recovering the energy carried by the fusion neutrons (about 80% of the fusion energy), the blanket of the reactor will thus have to breed tritium to replace that burnt in the fusion process. This paper is an attempt to cover in a concise way the questions of tritium breeding, and the influence of this issue on the design of, and the material selection for, power reactor blanket relying on the use of solid breeder materials. Tritium breeding requirements - to breed one tritium per fusion neutron - are shown to be quite demanding. To meet them, the blanket must incorporate, in addition to a tritium breeding lithium compound, a neutron multiplier so as to compensate for neutron losses. Presently prefered lithium compounds are Li2O, LiAlO2, Li2ZrO3, Li4SiO4. The neutron multiplier considered in most design concepts is beryllium. Furthermore, the blanket must be designed with a view to minimizing these neutron losses (search for compactness and high coverage ratio of the plasma while minimizing the amount of structures and coolant). The design guidelines are justified and the technological problems which limit their implementation are discussed and illustrated with typical designs of solid breeder blanket. (orig.)

  8. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    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.)

  9. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

    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

  10. Neutronic implications of lead-lithium blankets

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W.R.

    1982-08-01

    Lead-lithium alloys have been proposed for use in several conceptual blanket designs for both inertial and magnetic confinement fusion reactors. In most cases, Pb/sub 83/Li/sub 17/, a eutectic with a melting point of 235/sup 0/C, is the chosen composition. The primary reasons for using Pb/sub 83/Li/sub 17/ instead of Li as the tritium breeding material are the perceived safety advantages, low tritium solubility, and favorable neutronic characteristics. This paper describes the neutronic characteristics of Pb/sub 83/Li/sub 17/ blankets with emphasis on the enhanced neutron leakage through chamber ports and the degradation in blanket performance parameters that occurs as a result of the enhanced leakage.

  11. Fabrication techniques development of test blanket module based on CLAM

    International Nuclear Information System (INIS)

    The Reduced Activation Ferritic/Martensitic steels (RAFMs) are considered as the primary candidate structural material for the DEMO fusion reactor and the first fusion power plant. China Low Activation Martensitic (CLAM) steel, a version of RAFMs, is being developed in ASIPP (Institute of Plasma Physics, Chinese Academy of Sciences), under wide collaboration with many institutes and universities in China and overseas. The designs of FDS (Fusion Design Study) series liquid LiPb blankets for fusion reactors and corresponding Dual Functional Lithium Lead (DFLL) Test Blanket Module (TBM) in International Thermonuclear Experimental Reactor (ITER) are being carried out in ASIPP. And CLAM steel is chosen as the primary candidate structural material in these designs. So the fabrication techniques for DFLL TBM with CLAM are or urgently needed to be studied in detail. The fabrication of DFLL TBM mainly includes the manufacturing of the First Wall (FW), the Cooling Plates (CP) and the joining of the FW and CPs. Currently, solid Hot Isostatic Pressing (HIP) bonding and uniaxial diffusion bonding method are the most promising candidate fabrication method for the FW and CP. Experiments of HIP and unixial diffusion bonding of CLAM/CLAM were carried out and good joints were obtained. As for the joining technique of FW and CPs, the fusion welding techniques such as Tungsten Inert Gas welding, Laser welding and Electron Beam welding are candidates. Preliminary experiments on these welding techniques were performed. The simulation of thermal process by Gleeble 2000 was also carried out. Results of these experiments are summarized and further R and D plan on blanket fabrication techniques is also stated. (authors)

  12. Some new ideas for Tandem Mirror blankets

    International Nuclear Information System (INIS)

    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

  13. Lightweight IMM PV Flexible Blanket Assembly

    Science.gov (United States)

    Spence, Brian

    2015-01-01

    Deployable Space Systems (DSS) has developed an inverted metamorphic multijunction (IMM) photovoltaic (PV) integrated modular blanket assembly (IMBA) that can be rolled or z-folded. This IMM PV IMBA technology enables a revolutionary flexible PV blanket assembly that provides high specific power, exceptional stowed packaging efficiency, and high-voltage operation capability. DSS's technology also accommodates standard third-generation triple junction (ZTJ) PV device technologies to provide significantly improved performance over the current state of the art. This SBIR project demonstrated prototype, flight-like IMM PV IMBA panel assemblies specifically developed, designed, and optimized for NASA's high-voltage solar array missions.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    A preconceptual engineering design has been developed for the 3He Target/Blanket (T/B) System for the Accelerator Production of Tritium Project. This concept uses an array of pressure tubes containing tungsten rods for the neutron spallation source and 3He gas contained in a metal tank and blanket tubes as the tritium production material. The engineering design is based on a physics model optimized for efficient tritium production. Principle engineering consideration were: provisions for cooling all materials including the 3He gas; containment of the gas and radionuclides; remote handling; material compatibility; minimization of 3He, D2O, and activated waste; modularity; and manufacturability. The design provides a basis for estimating the cost to implement the system

  16. Advanced Polymer For Multilayer Insulating Blankets

    Science.gov (United States)

    Haghighat, R. Ross; Shepp, Allan

    1996-01-01

    Polymer resisting degradation by monatomic oxygen undergoing commercial development under trade name "Aorimide" ("atomic-oxygen-resistant imidazole"). Intended for use in thermal blankets for spacecraft in low orbit, useful on Earth in outdoor applications in which sunlight and ozone degrades other plastics. Also used, for example, to make threads and to make films coated with metals for reflectivity.

  17. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

    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.)

  18. Aerogel Blanket Insulation Materials for Cryogenic Applications

    Science.gov (United States)

    Coffman, B. E.; Fesmire, J. E.; White, S.; Gould, G.; Augustynowicz, S.

    2009-01-01

    Aerogel blanket materials for use in thermal insulation systems are now commercially available and implemented by industry. Prototype aerogel blanket materials were presented at the Cryogenic Engineering Conference in 1997 and by 2004 had progressed to full commercial production by Aspen Aerogels. Today, this new technology material is providing superior energy efficiencies and enabling new design approaches for more cost effective cryogenic systems. Aerogel processing technology and methods are continuing to improve, offering a tailor-able array of product formulations for many different thermal and environmental requirements. Many different varieties and combinations of aerogel blankets have been characterized using insulation test cryostats at the Cryogenics Test Laboratory of NASA Kennedy Space Center. Detailed thermal conductivity data for a select group of materials are presented for engineering use. Heat transfer evaluations for the entire vacuum pressure range, including ambient conditions, are given. Examples of current cryogenic applications of aerogel blanket insulation are also given. KEYWORDS: Cryogenic tanks, thermal insulation, composite materials, aerogel, thermal conductivity, liquid nitrogen boil-off

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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 UO2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li2O 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

  1. Design of test blanket system for ITER module testing

    International Nuclear Information System (INIS)

    Test blanket systems to be installed in ITER for developing demo blankets have been investigated. One of the main engineering goals of ITER is to test tritium breeding blankets relevant to a power reactor. The test foreseen on modules include the demonstration of a breeding capability that would lead to tritium self-sufficiency in a reactor and extraction of a high grade heat suitable for electricity generation. To accomplish these goals, several ITER equatorial ports are available to test the test blanket systems, both in the basic performance phase (BPP) and the enhanced performance phase (EPP). Test blanket systems for water-cooled and helium-cooled type DEMO blankets with ceramic breeders, developed in Japan have been designed. The design activities include the neutronics, thermal and hydraulic analyses, and mechanical configuration considering port sharing, cooling systems and tritium recovery systems, and test blanket system compatible with the current ITER design has been developed. (author)

  2. Neutronic analysis of alternative structural materials for fusion reactor blankets

    International Nuclear Information System (INIS)

    The neutronic performance of the International Tokamak Reactor (INTOR) blanket was studied when several alternative structural materials were used instead of the INTOR reference structural material, type 316 stainless steel. The alternative structural materials included: ferritic-, vanadium-, titanium-, long range ordered-, manganese austenitic-, and nimonic-alloys. All were treated both with and without a first-wall coating of beryllium or graphite. The tritium breeding ratio, the nuclear heating, and the gas (hydrogen and helium) production rates in the structural materials were calclated for the possible combinations of structural material and first-wall coating. These parameters were compared with those obtained by using SS-316. The nimonic alloy was the only one with worse neutronic performance than the SS-316. (orig.)

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

    International Nuclear Information System (INIS)

    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.

  4. Water-cooled blanket concepts for the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    The primary goal of the Blanket Comparison and Selection Study (BCSS) was to select a limited number of blanket concepts for fusion power reactors, to serve as the focus for the U.S. Department of Energy blanket research and development program. The concepts considered most seriously by the BCSS can be grouped for discussion purposes by coolant: liquid metals and alloys, pressurized water, helium, and nitrate salts. Concepts using pressurized water as the coolant are discussed. Water-cooled concepts using liquid breeders-lithium and 17Li-83Pb (LiPb)-have severe fundamental safety problems. The use of lithium and water in the blanket was considered unacceptable. Initial results of tests at Hanford Engineering Development Laboratory using steam injected into molten LiPb indicate that use of LiPb and water together in a blanket is a very serious concern from the safety standpoint. Key issues for water-cooled blankets with solid tritium breeders (Li2O, or a ternary oxide such as LiAlO2) were identified and examined: reliability against leaks, control of tritium permeation into the coolant, retention of breeder physical integrity, breeder temperature predictability, determination of allowable temperature limits for breeders, and 6Li burnup effects (for LiAlO2). The BCSS's final rankings and associated rationale for all water-cooled concepts are examined. Key issues and factors for tokamak and tandem mirror reactor versions of water-cooled solid breeder concepts are discussed. The reference design for the top-ranked concept-LiAlO2 breeder, ferritic steel structure, and beryllium neutron multiplier-is presented. Finally, some general conclusions for water-cooled blanket concepts are drawn based on the study's results

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

    International Nuclear Information System (INIS)

    Tritium control issues such as the development of permeation barriers and the choice of the coolant and purge-gas chemistry are of crucial importance for solid breeder blankets. In order to quantify these problems for the helium-cooled ceramic BIT blanket concept, the tritium leakage into the coolant was evaluated and the consequent tritium losses into the steam circuit were determined. Our results indicate that under certain specified conditions the total tritium release from the coolant can be limited to approximately 10 Ci/d, but only on the assumption that experimental data for tritium permeation barriers can be attained under realistic operating conditions. An experimental study on the impact of the gas chemistry on tritium losses is proposed. (orig.)

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

    International Nuclear Information System (INIS)

    Tritium control issues such as the development of permeation barriers and the choice of the coolant and purge-gas chemistry are of crucial importance for solid breeder blankets. In order to quantify these problems for the helium-cooled ceramic breeder-inside-tube (BIT) blanket concept, the tritium leakage into the coolant was evaluated and the consequent tritium losses into the steam circuit were determined. The results indicate that under certain specified conditions the total tritium release from the coolant can be limited to approximately 10 Ci/d, but only on the assumption that experimental data for tritium permeation barriers can be attained under realistic operating conditions. An experimental study on the impact of the gas chemistry on tritium losses is proposed. (author) 8 refs.; 2 figs

  7. Water-cooled lithium-lead blanket

    International Nuclear Information System (INIS)

    The paper is an appendix to a study of the reactor relevance of the NET design concept. The present study examines whether the water-cooled lithium-lead blanket designed for NET can be directly extrapolated to a demonstration (DEMO) reactor. A fundamental requirement of the exercise is that the DEMO design should have a tritium breeding ratio which is higher than that in NET. The water-cooled lithium-lead blanket is discussed with respect to: neutronics design, design parameter survey and thermohydraulics, and engineering design. Results are reported of three-dimensional calculations using the Monte Carlo code MORSE-H to investigate possible neutron leakage between the poloidally disposed breeder tubes, and to determine the global tritium breeding ratio for the final double null machine design. (U.K.)

  8. Diffusive heat blanketing envelopes of neutron stars

    CERN Document Server

    Beznogov, M V; Yakovlev, D G

    2016-01-01

    We construct new models of outer heat blanketing envelopes of neutron stars composed of binary ion mixtures (H - He, He - C, C - Fe) in and out of diffusive equilibrium. To this aim, we generalize our previous work on diffusion of ions in isothermal gaseous or Coulomb liquid plasmas to handle non-isothermal systems. We calculate the relations between the effective surface temperature Ts and the temperature Tb at the bottom of heat blanketing envelopes (at a density rhob= 1e8 -- 1e10 g/cc) for diffusively equilibrated and non-equilibrated distributions of ion species at different masses DeltaM of lighter ions in the envelope. Our principal result is that the Ts - Tb relations are fairly insensitive to detailed distribution of ion fractions over the envelope (diffusively equilibrated or not) and depend almost solely on DeltaM. The obtained relations are approximated by analytic expressions which are convenient for modeling the evolution of neutron stars.

  9. Novel method for sludge blanket measurements.

    Science.gov (United States)

    Schewerda, J; Förster, G; Heinrichmeier, J

    2014-01-01

    The most widely used methods for sludge blanket measurements are based on acoustic or optic principles. In operation, both methods are expensive and often maintenance-intensive. Therefore a novel, reliable and simple method for sludge blanket measurement is proposed. It is based on the differential pressure measurement in the sludge zone compared with the differential pressure in the clear water zone, so that it is possible to measure the upper and the lower sludge level in a tank. Full-scale tests of this method were done in the secondary clarifier at the waste water treatment plant in Hecklingen, Germany. The result shows a good approximation of the manually measured sludge level. PMID:24569276

  10. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

    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

  11. Stellar model atmospheres with magnetic line blanketing

    CERN Document Server

    Kochukhov, O; Shulyak, D

    2004-01-01

    Model atmospheres of A and B stars are computed taking into account magnetic line blanketing. These calculations are based on the new stellar model atmosphere code LLModels which implements direct treatment of the opacities due to the bound-bound transitions and ensures an accurate and detailed description of the line absorption. The anomalous Zeeman effect was calculated for the field strengths between 1 and 40 kG and a field vector perpendicular to the line of sight. The model structure, high-resolution energy distribution, photometric colors, metallic line spectra and the hydrogen Balmer line profiles are computed for magnetic stars with different metallicities and are discussed with respect to those of non-magnetic reference models. The magnetically enhanced line blanketing changes the atmospheric structure and leads to a redistribution of energy in the stellar spectrum. The most noticeable feature in the optical region is the appearance of the 5200 A depression. However, this effect is prominent only in ...

  12. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    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)

  13. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    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. Analysis of Consistency of Printing Blankets using Correlation Technique

    Directory of Open Access Journals (Sweden)

    Balaraman Kumar

    2010-06-01

    Full Text Available This paper presents the application of an analytical tool to quantify material consistency of offset printing blankets. Printing blankets are essentially viscoelastic rubber composites of several laminas. High levels of material consistency are expected from rubber blankets for quality print and for quick recovery from smash encountered during the printing process. The present study aims at determining objectively the consistency of printing blankets at three specific torque levels of tension under two distinct stages; 1. under normal printing conditions and 2. on recovery after smash. The experiment devised exhibits a variation in tone reproduction properties of each blanket signifying the levels of inconsistency also in thickness direction. Correlation technique was employed on ink density variations obtained from the blanket on paper. Both blankets exhibited good consistency over three torque levels under normal printing conditions. However on smash the recovery of blanket and its consistency was a function of manufacturing and torque levels. This study attempts to provide a new metrics for failure analysis of offset printing blankets. It also underscores the need for optimising the torque for blankets from different manufacturers.

  15. Analysis of Consistency of Printing Blankets using Correlation Technique

    Directory of Open Access Journals (Sweden)

    Lalitha Jayaraman

    2010-01-01

    Full Text Available This paper presents the application of an analytical tool to quantify material consistency of offset printing blankets. Printing blankets are essentially viscoelastic rubber composites of several laminas. High levels of material consistency are expected from rubber blankets for quality print and for quick recovery from smash encountered during the printing process. The present study aims at determining objectively the consistency of printing blankets at three specific torque levels of tension under two distinct stages; 1. under normal printing conditions and 2. on recovery after smash. The experiment devised exhibits a variation in tone reproduction properties of each blanket signifying the levels of inconsistency also in thicknessdirection. Correlation technique was employed on ink density variations obtained from the blanket on paper. Both blankets exhibited good consistency over three torque levels under normal printing conditions. However on smash the recovery of blanket and its consistency was a function of manufacturing and torque levels. This study attempts to provide a new metrics for failure analysis of offset printing blankets. It also underscores the need for optimizing the torque for blankets from different manufacturers.

  16. Detection of Breeding Blankets Using Antineutrinos

    Science.gov (United States)

    Cogswell, Bernadette; Huber, Patrick

    2016-03-01

    The Plutonium Management and Disposition Agreement between the United States and Russia makes arrangements for the disposal of 34 metric tons of excess weapon-grade plutonium. Under this agreement Russia plans to dispose of its excess stocks by processing the plutonium into fuel for fast breeder reactors. To meet the disposition requirements this fuel would be burned while the fast reactors are run as burners, i.e., without a natural uranium blanket that can be used to breed plutonium surrounding the core. This talk discusses the potential application of antineutrino monitoring to the verification of the presence or absence of a breeding blanket. It is found that a 36 kg antineutrino detector, exploiting coherent elastic neutrino-nucleus scattering and made of silicon, could determine the presence of a breeding blanket at a liquid sodium cooled fast reactor at the 95% confidence level within 90 days. Such a detector would be a novel non-intrusive verification tool and could present a first application of coherent elastic neutrino-nucleus scattering to a real-world challenge.

  17. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    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/m2 and a particle heat flux of 1 MW/m2. 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

  18. The ITER Blanket System Design Challenge

    International Nuclear Information System (INIS)

    Full text: The blanket system is one of the most technically challenging components of the ITER machine, having to accommodate high heat fluxes from the plasma, large electromagnetic loads during off-normal events and demanding interfaces with many key components (in particular the vacuum vessel and in-vessel coils) and the plasma. Plasma scenarios impose demanding requirements on the blanket in terms of heat fluxes on various areas of the first wall during different phases of operation (inboard and outboard midplane for start-up/shut-down scenarios and the top region close to the secondary X-point during flat top) as well as large electro-magnetic (EM) loads and transient energy deposition during off-normal plasma events (such as disruptions and vertical displacement events (VDE)). The high heat fluxes resulting in some areas have necessitated the use of “enhanced heat flux” panels capable of accommodating an incident heat flux of up to 5 MW/m2 in steady state. The other regions utilize “normal heat flux” panels, which have been developed and tested for a heat flux of the order of 1 — 2 MW/m2. The FW shaping design requires a compromise between the conflicting requirements for accommodation of steady state and transient loads (energy deposition during off-normal events). A shaped surface increases the heat loads which are due to plasma particles following the field lines compared to a perfectly toroidal surface. The blanket provides a major contribution to the shielding of the vacuum vessel and coils. A challenging criterion is the need to limit the integrated heating in the toroidal field coil (TFC) to ∼ 14 kW. This is particularly severe on the inboard leg where approximately 80% of the total nuclear heat on the TFC is deposited. Several design modifications were considered and analyzed to help achieve this, including increasing the inboard blanket radial thickness and reducing the assembly gaps. This paper summarizes the latest progress in the

  19. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    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

  20. Fast-Breeder-Blanket Project: FBBF. Final report

    International Nuclear Information System (INIS)

    This report is the final report for DOE contract DE-AC02-76ET37237 with the Purdue Fast Breeder Blanket Project. The Project was initiated to investigate the uncertainties in Fast Breeder Reactor blanket calculations. Absolute measurements of key neutron reaction rates, neutron spectra, and gamma-ray energy depositions were made in simulated FBF blankets in the Fast Breeder Blanket Facility (FBBF), a Cf-252 driven subcritical facility. Calculation of the spectra and integral reaction rates were made using methods, computer codes, and cross section data typical of those currently used in the design of FBR's. Comparisons of calculated to experimental integral neutron reaction rates give good agreement at the inner portions of the blanket by diverge to C/E ratios of about 0.65 at the outer edge of the blanket for reactions sensitive to the neutron density

  1. The frontiers of research on fusion blanket technology

    International Nuclear Information System (INIS)

    Current topics concerning blanket technology are reviewed. In the chemical engineering/chemistry area, the qualitative and quantitative effects of mass transfer steps of tritium is important in the understanding of the behavior of bred tritium in the solid breeder blanket system. Such phenomena as adsorption, isotope exchange reactions, and water formation reaction at the grain surface produce profound effects on the behavior of the bred tritium in the blanket. Regarding the liquid system, the physical or chemical properties of Li, Li17Pb83 and Flibe as liquid blanket materials were compared. Some recent studies were introduced regarding tritium recovery from the liquid blanket materials, impurity removal from salts, ceramic coating of structural materials, and the vapor pressure of mixtures of metals or salts. Thermal hydraulic topics in relation to several candidate power reactor concepts are summarized. Emphasis is laid on the simultaneous removal of heat and tritium from the blanket and some aspects of forming effective power cycles are developed. (author)

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

    International Nuclear Information System (INIS)

    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

  3. Analysis of deficiencies in fast reactor blanket physics predictions

    International Nuclear Information System (INIS)

    This analysis addresses a deviation between experimental measurements and fast reactor blanket physics predictions. A review of worldwide results reveals that reaction rates in the blanket are underpredicted with the discrepancy increasing with penetration into the blanket. The analysis of this discrepancy involves two parts: quantifying possible error reductions using the most advanced methods and investigating deficiencies in current methodology. The source of these discrepancies was investigated by application of ''state-of-the-art'' group constant generation and flux prediction methodology to flux calculations for the Purdue University Fast Breeder Blanket Facility (FBBF). Refined group constant generation methods yielded a significant reduction in the blanket deviations; however, only about half of the discrepancy can be accounted for in this manner. Transport theory calculations were used to predict the blanket neutron transmission problem. The surprising result is that transport theory predictions utilizing diffusion theory group constants did not improve the blanket results. Transport theory predictions exhibited blanket underpredictions similar to the diffusion theory results. The residual blanket discrepancies not explained using advanced methods require a refinement of the theory. For this purpose an analysis of deficiencies in current methodology was performed

  4. Advanced Acoustic Blankets for Improved Aircraft Interior Noise Reduction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this project advanced acoustic blankets for improved low frequency interior noise control in aircraft will be developed and demonstrated. The improved...

  5. Vibration damage testing of thermal barrier fibrous blanket insulation

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The design of advanced solid breeding blanket in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high fluence, and the development of such as advanced blanket materials has been carried out by the cooperation activities among JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by wet process is a reference material as a tritium breeder, but the stability on high temperature has to be improved for application to DEMO blanket. As one of such the improved materials, TiO2-doped Li2TiO3 pebbles were successfully fabricated and TiO2-doped Li2TiO3 has been studied. For the advanced neutron multiplier, the beryllides that have high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that Be12Ti had lower swelling and tritium inventory than that of beryllium metal. The pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. From these activities, the bright prospect was obtained to realize the DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides. (author)

  7. Systematic methodology for estimating direct capital costs for blanket tritium processing systems

    International Nuclear Information System (INIS)

    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

  8. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    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

  9. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    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

  10. Fusion reactor blanket: neutronic studies in France

    International Nuclear Information System (INIS)

    The problem of effective tritium regeneration is a crucial issue for the fusion reactor, especially 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 analysis. The results of these studies permit us to conclude that it is possible to expect an adequate tritium breeding ratio

  11. ITER solid breeder blanket materials database

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States); Dienst, W. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Material- und Festkoerperforschung; Flament, T. [CEA Centre d`Etudes de Fontenay-aux-Roses (France). Commissariat A L`Energie Atomique; Lorenzetto, P. [NET Team, Garching (Germany); Noda, K. [Japan Atomic Energy Research Inst., Takai, Ibaraki, (Japan); Roux, N. [CEA Centre d`Etudes et de Recherches Les Materiaux (France). Commissariat a L`Energie Atomique

    1993-11-01

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

  12. Steady-state thermal-hydraulic of pebble bed blanket on hybrid reactor

    International Nuclear Information System (INIS)

    This paper gives thermal-hydraulic studies of pebble bed blanket on Hybrid Reactor. The concept of whole pebble bed blanket and the cooling methods are presented. The thermal-hydraulic characteristics of pebble bed are summarized. The theoretical model and code for solving heat transfer and flowing are presented. By using this code the calculation and analysis of thermal hydraulic of pebble bed Blanket of Hybrid Reactor are also given. In order to improve the flexibility, safety and economy, the authors select pebble beds not only to breed Tritium, but also to breed fission material and to multiply neutron. 5 MPa Helium is used as coolant and 0.05 MPa-0.1 MPa Helium is used as Purge gas. The heat transfer mechanisms of pebble bed are very complicated which include conduction, convection and radiation. In order to study the thermal-hydraulic of the bed, the authors just simply consider it as homogeneous and continuous binary phase medium as that used in the porous medium at the condition that the size of the bed is much greater than that of the balls. The coolant or the purge gas flowing through the bed is just considered existing a cooling source in the bed. It also significantly influences the effective conductivity's of the bed. Porous fraction, the main factor of the bed depends on the geometry position and parameters. From this model, one can obtain the thermal-hydraulic governing equations of the bed

  13. Water chemistry control for the target/blanket region of the accelerator production of tritium

    International Nuclear Information System (INIS)

    High-energy particle interactions in the various components of the target/blanket region of the Accelerator Production of Tritium lead to heat generation and deposition. Heavy-water and light-water systems are used to cool the target/blanket system and associated equipment. Structural materials include Inconel alloy 718, aluminum-clad lead rods, aluminum tubes containing helium-3 and tritium gas, and stainless steel components. Proper coolant chemistry is required to maximize neutron production, minimize corrosion of components, and minimize activity buildup. Corrosion-related phenomena and development of coolant and moderator corrosion control for both power and defense fission reactors has been studied extensively over the past 50 years. Less is known, however, about cooling systems for accelerators where a variety of transient chemical species and spallation products may be formed. The following provides a discussion on the issues that need to be addressed for proper water chemistry control for the APT system

  14. An overview on tritium permeation barrier development for WCLL blanket concept

    International Nuclear Information System (INIS)

    The reduction of tritium permeation through blanket structural materials and cooling tubes has to be carefully evaluated to minimise radiological hazards. A strong effort has been made in the past to select the best technological solution for the realisation of tritium permeation barriers (TPB) on complex structures not directly accessible after the completion of the manufacturing process. The best solution was identified in aluminium rich coatings, which form Al2O3 at their surface. Two technologies were selected as reference for the realisation of coating in the WCLL blanket concept: the chemical vapour deposition (CVD) process developed on laboratory scale by CEA, and the hot dipping (HD) process developed by FZK. The results obtained during three years of tests on CVD and HD coated specimens in gas and liquid metal phase are summarised and discussed

  15. Program element 2: Blanket and shield thermal-hydraulic and thermomechanical testing of the first Wall/Blanket/Shield program: Final report

    International Nuclear Information System (INIS)

    Two single-effect scoping tests were performed in Phase 1 of this project. These tests included the measurement of heat transfer contact resistance between a Li2O breeder pellet and stainless steel clad in argon gas - in the temperature range of 5000 to 7000C. The effective heat transfer coefficient was found to lie in the range of 1100 to 1700 W/m2-K which is acceptable for the design of Li2O solid breeder blankets. This effective heat transfer coefficient is expected to improve in helium gas, which has a much higher thermal conductivity than argon. In the second set of tests, the thermomechanical behavior of Li2O pellets was determined under simulated blanket conditions, including thermal cycling, in the presence of a helium gas purge flow of controlled moisture content. The results showed that at a temperature below 8000C and with a reasonable purge flow moisture content of ≅10 ppM, Li2O pellet sintering and pore closure, pellet-clad interaction and vapor phase transport of LiOH are minimal. A multiple-effects solid breeder blanket integral simulation test, followed by an engineering-scale nuclear test are the proposed next steps in this series of tests. Preliminary designs for these experiments were completed. Fabrication techniques for the manufacturing of microspheres as an alternate fuel form for solid breeder fuel were also investigated in the study. Microspheres of LiAlO2 from <30 μm to 180 μm were fabricated by the plasma spray technique. Production of ∼650 μm spheres by a sol-gel process was demonstrated and feasibility was indicated from the production of spheres from ∼100 μm to ∼800 μm in diameter

  16. ITER blanket manifold system: Integration, assembly and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Alex, E-mail: alex.martin@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Dellopoulos, George [F4E, EU ITER Domestic Agency, Barcelona (Spain); Edwards, Paul; Furmanek, Andreas; Gicquel, Stefan; Macklin, Brian [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Martin, Patrick [RÜECKER LYPSA, Carretera del Prat, 65, Cornellá de Llobregat (Spain); Merola, Mario; Norman, Mark; Raffray, Rene [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-10-15

    Highlights: •The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. •-The blanket manifold system has been redesigned to improve leak detection and localization. •-The redesign of the blanket manifold system into a system based on individual pipes has proven to be a major engineering challenge. -- Abstract: The ITER Tokamak Cooling Water System (TCWS) provides coolant for blankets and divertor. The blanket system consists of 440 blanket modules (BMs). The blanket manifold consists of a system of seamless pipes arranged in bundles and routed in poloidal direction from the upper ports of the Vacuum Vessel (VV) to the bottom of the machine. In each of the 18 upper ports there are 20 inlet and 20 outlet pipes, which split at the port exit in two directions, supplying cooling water to either the inboard or the outboard blanket modules. The manifold is routed between the VV and BMs. Branch pipes provide the connection between the manifold and the blanket cooling circuits through a coaxial connector welded to the shield block. A complex, sequential installation sequence has been developed in order to enable the assembly. Once installed the manifold is considered a semi-permanent component, but since failure would prevent ITER operation a maintenance strategy has been planned.

  17. Remote handling demonstration of ITER blanket module replacement

    International Nuclear Information System (INIS)

    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)

  18. Development of the Helium Cooled Lithium Lead blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aiello, G., E-mail: giacomo.aiello@cea.fr [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Aubert, J.; Jonquères, N. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Li Puma, A. [CEA-Saclay, DEN/DANS/DM2S/SERMA/LPEC, 91191 Gif Sur Yvette Cedex (France); Morin, A.; Rampal, G. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France)

    2014-10-15

    Highlights: • The HCLL blanket design has been modified to adapt it to the 2012 EFDA DEMO specifications. • The new design has been developed with the aim to capitalize on TBM experience in ITER. • A new attachment system for the modules has been proposed. - Abstract: The Helium Cooled Lithium Lead (HCLL) blanket is one of the candidate European blanket concepts selected for the DEMOnstration fusion power plant that should follow ITER. In a fusion power plant, the blanket is one of the key components because of its impact on the plant performance, availability, safety and economics. In 2012, the European Fusion Development Agreement (EFDA) agency issued new specifications for DEMO: this paper describes the work performed to adapt the previous 2007 HCLL-DEMO blanket design to those specifications. A new segmentation has been defined assuming straight surfaces for all blanket modules. Following the Multi Module Segment (MMS) option, all modules are attached to a common back supporting structure which also serves as manifold for Helium and PbLi distribution. A detailed CAD design of the central outboard module has been defined. Thermo-hydraulic and thermo-mechanical analyses on of the First Wall and Breeder Zone have been carried out. For the attachment of the modules to the common backplate, a new solution based on the use of Tie Rods, derived from the design of the corresponding HCLL Test Blanket Module for ITER, has been proposed. This paper also identifies the priorities for further development of the HCLL blanket design.

  19. Heterogeneous structure effect on molten salt blanket neutronics

    Energy Technology Data Exchange (ETDEWEB)

    Grebyonkin, K.F.; Kandiev, Ya.Z.; Malyshkin, G.N.; Orlov, A.I. [Inst. of Technical Pysics, Chelyabinsk (Russian Federation). Dept. of Physics

    1997-09-01

    The report presents the results of the molten salt blanket neutronics calculations performed for researchers of a facility for accelerator-driven transmutation of long-lived radioactive wastes and plutonium conversion. Heterogeneous structure effect on molten salt blanket neutronics was studied through computation. 4 refs., 1 fig., 1 tab.

  20. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

    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)

  1. Breeding blanket concepts for fusion and materials requirements

    International Nuclear Information System (INIS)

    This paper summarizes the design and performances of recent breeding blanket concepts and identifies the key material issues associated with them. An assessment of different classes of concepts is carried out by balancing out the potential performance of the concepts with the risk associated with the required material development. Finally, an example strategy for blanket development is discussed

  2. ITER blanket manifold system: Integration, assembly and maintenance

    International Nuclear Information System (INIS)

    Highlights: •The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. •-The blanket manifold system has been redesigned to improve leak detection and localization. •-The redesign of the blanket manifold system into a system based on individual pipes has proven to be a major engineering challenge. -- Abstract: The ITER Tokamak Cooling Water System (TCWS) provides coolant for blankets and divertor. The blanket system consists of 440 blanket modules (BMs). The blanket manifold consists of a system of seamless pipes arranged in bundles and routed in poloidal direction from the upper ports of the Vacuum Vessel (VV) to the bottom of the machine. In each of the 18 upper ports there are 20 inlet and 20 outlet pipes, which split at the port exit in two directions, supplying cooling water to either the inboard or the outboard blanket modules. The manifold is routed between the VV and BMs. Branch pipes provide the connection between the manifold and the blanket cooling circuits through a coaxial connector welded to the shield block. A complex, sequential installation sequence has been developed in order to enable the assembly. Once installed the manifold is considered a semi-permanent component, but since failure would prevent ITER operation a maintenance strategy has been planned

  3. MIT LMFBR blanket research project. Final summary report

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. LMFBR Blanket Physics Project progress report No. 6

    International Nuclear Information System (INIS)

    Progress is summarized in experimental and analytical investigations of the neutronics and photonics of benchmark mockups of LMFBR blankets. During the reporting period work was devoted primarily to a wide range of analytical/numerical investigations, including blanket fuel management/economics studies, evaluation of improved blanket designs, and assessment of state-of-the-art methods for gamma heating calculations. Experimental work included preparations for resumption of MIT Reactor operations, primarily fabrication of improved steel reflector assemblies for blanket mockups, and development of an improved radiophotoluminescent readout device for LiF thermoluminescent detectors. The most significant finding was that the neutronic and economic performance of radial blanket assemblies are essentially independent of core size (rating) for radially-power-flattened cores. Hence the methodology and results of current experiments and calculations should be valid for the large commercial LMFBR's of the future

  5. LMFBR Blanket Physics Project progress report No. 6

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J. (ed.)

    1975-06-30

    Progress is summarized in experimental and analytical investigations of the neutronics and photonics of benchmark mockups of LMFBR blankets. During the reporting period work was devoted primarily to a wide range of analytical/numerical investigations, including blanket fuel management/economics studies, evaluation of improved blanket designs, and assessment of state-of-the-art methods for gamma heating calculations. Experimental work included preparations for resumption of MIT Reactor operations, primarily fabrication of improved steel reflector assemblies for blanket mockups, and development of an improved radiophotoluminescent readout device for LiF thermoluminescent detectors. The most significant finding was that the neutronic and economic performance of radial blanket assemblies are essentially independent of core size (rating) for radially-power-flattened cores. Hence the methodology and results of current experiments and calculations should be valid for the large commercial LMFBR's of the future.

  6. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    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

  7. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

    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 238U 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

  8. Radiolysis Experiments for the Aqueous Self-Cooled Blanket. Final report

    International Nuclear Information System (INIS)

    The results of Fusion Technology Task NAB 1.1 (Radiolytic Experiments for the ASCB), are reported . In the Aqueous Self-Cooled Blanket (ASCB) concept, an aqueous 6Li solution in a metallic structure is used as a shielding-breeding blanket for fusion reactors. Radiolysis could be very important with respect to the design and the use of an ASCB. The objectives of this project were to quantify the radiolytic decomposition of neutron irradiated aqueous lithium solutions and to demonstrate, if possible, the suppression of this decomposition by the initial addition of a small amount of hydrogen. Closed capsules, with the solutions and an inert gas or hydrogen as cover gas, were irradiated with thermal neutrons in a fission reactor. Radiolysis products hydrogen and oxygen (from hydrogen peroxide) as well as tritium were measured after irradiation. Tritium served as an internal dosimeter. The experimental results with LiNO3 , Li2SO4 and LiOH solutions indicate that the radiolytic gas production in an ASCB is proportional to the absorbed radiation energy. The observed radiation chemical yields allow the preliminary estimation of the radiolysis effects for a specific ASCB design. Contrary to the theoretical predictions, the use of hydrogen as a cover gas at up to 1 MPa had no measurable effect on the radiolytic gas production. Probably it will thus not be possible to suppress the radiolytic decomposition of a low-pressure ASCB by the addition of H2. Catalytic recombination will be required

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

    International Nuclear Information System (INIS)

    The International Thermonuclear Experimental Reactor (ITER) project began a new design phase called the Engineering Design Activity (EDA) which started in July 1992. A variety of blanket designs options were analyzed as a part of the U.S. ITER home team blanket option trade-off study (BOTS) which began in May 1993. The options considered were a self-cooled Li/V blanket, a helium cooled Li/V blanket and a water cooled 316 SS nonbreeding shield option. Detailed activation, dose rate and waste disposal rating calculations have been performed for these different ITER blanket design options based on a fluence of 3.0 MWa/m2 and an average neutron wall loading of 2.0 MW/m2. A continuous operation assumption was utilized in the analysis. The results of this work are presented in this conference

  10. 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.

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

    International Nuclear Information System (INIS)

    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 (Li2O) and lithium zirconate (Li2ZrO3) 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

  12. Manufacture of blanket shield modules for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lorenzetto, P. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany)]. E-mail: Patrick.Lorenzetto@tech.efda.org; Boireau, B. [AREVA Centre Technique de Framatome, BP181, F-71200 Le Creusot (France); Boudot, C. [AREVA Centre Technique de Framatome, BP181, F-71200 Le Creusot (France); Bucci, P. [CEA, DTEN/S3ME/LMIC, 17 rue des Martyrs, F-38054 Grenoble (France); Furmanek, A. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Ioki, K. [ITER IT, Boltzmannstr. 2, D-85748 Garching (Germany); Liimatainen, J. [Metso Powdermet, P.O. Box 306, FIN-33101 Tampere (Finland); Peacock, A. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Sherlock, P. [NNC Ltd., Booths Hall, Knutsford, Cheshire WA16 8QZ (United Kingdom); Taehtinen, S. [VTT Industrial Systems, P.O. Box 1704, Espoo, FIN-02044 VTT (Finland)

    2005-11-15

    A research and development programme for the ITER blanket shield modules has been implemented in Europe to provide input for the design and the manufacture of the full-scale production components. It involves in particular the fabrication and testing of mock-ups (small scale and medium scale) and full-scale prototypes of shield blocks (SB) and first wall (FW) panels. The manufacturing feasibility of FW panels has been demonstrated for two copper alloy candidates. Two designs have been developed for the manufacture of the SB, one for a conventional fabrication route and one for a fabrication route based on the hot isostatic press technology. This paper presents the fabrication routes developed in Europe for the manufacture of the ITER Shield modules.

  13. Remote-handling concept for target/blanket modules in the accelerator production of tritium

    International Nuclear Information System (INIS)

    The accelerator production of tritium (APT) has been proposed as the source of tritium for the United States in the next century. The APT will accelerate protons that will strike replaceable tungsten target modules. The tungsten target modules generate neutrons that pass through blanket modules and other modules where He gas is turned into tritium. The target and blanket modules are predicted to require replacement every 1 to 10 yr, depending on their location. The target modules may weigh as much as 78.8 tonnes (85 t) each. All of the modules will be contained in a target/blanket vessel, which is in a shielded facility. The spent modules will be very radioactive so that remote replacement is required. A proposed concept is to use a remotely operated bridge crane and a remotely operated, bridge-mounted manipulator to perform the entire replacement operation. This will require removing/replacing the vessel lid, installing/removing temporary water cooling, closing/opening valves on manifolds and modules, draining of jumpers, removing/replacing jumpers, removing/replacing shielding keys, and removing/replacing the modules. This application is unique because of the size and weight of the modules, the precision required, the type of connectors required, and the complexity of the entire operation. A three-dimensional simulation of the entire module replacement operation has been developed to help understand, communicate, and refine the concepts

  14. AB Blanket for Cities (for continual pleasant weather and protection from chemical, biological and radioactive weapons)

    CERN Document Server

    Bolonkin, Alexander

    2009-01-01

    In a series of previous articles (see references) the author offered to cover a city or other important large installations or subregions by a transparent thin film supported by a small additional air overpressure under the form of an AB Dome. The building of a gigantic inflatable AB Dome over an empty flat surface is not difficult. However, if we want to cover a city, garden, forest or other obstacle course we cannot easily deploy the thin film over building or trees. In this article is suggested a new method which solves this problem. The idea is to design a double film blanket filled by light gas (for example, methane, hydrogen, or helium). Sections of this AB Blanket are lighter then air and fly in atmosphere. They can be made on a flat area (serving as an assembly area) and delivered by dirigible or helicopter to station at altitude over the city. Here they connect to the already assembled AB Blanket subassemblies, cover the city in an AB Dome and protect it from bad weather, chemical, biological and rad...

  15. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

    Aqueous solution blanket using lithium salts such as LiNO3 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)

  16. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    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

  17. Spacecraft thermal blanket cleaning: Vacuum bake of gaseous flow purging

    Science.gov (United States)

    Scialdone, John J.

    1990-01-01

    The mass losses and the outgassing rates per unit area of three thermal blankets consisting of various combinations of Mylar and Kapton, with interposed Dacron nets, were measured with a microbalance using two methods. The blankets at 25 deg C were either outgassed in vacuum for 20 hours, or were purged with a dry nitrogen flow of 3 cu. ft. per hour at 25 deg C for 20 hours. The two methods were compared for their effectiveness in cleaning the blankets for their use in space applications. The measurements were carried out using blanket strips and rolled-up blanket samples fitting the microbalance cylindrical plenum. Also, temperature scanning tests were carried out to indicate the optimum temperature for purging and vacuum cleaning. The data indicate that the purging for 20 hours with the above N2 flow can accomplish the same level of cleaning provided by the vacuum with the blankets at 25 deg C for 20 hours, In both cases, the rate of outgassing after 20 hours is reduced by 3 orders of magnitude, and the weight losses are in the range of 10E-4 gr/sq cm. Equivalent mass loss time constants, regained mass in air as a function of time, and other parameters were obtained for those blankets.

  18. Requirements for a helium-cooled blanket heat removal system development facility for fusion reactor research

    International Nuclear Information System (INIS)

    Existing and potential design problems associated with the helium-cooled blanket assemblies of experimental, demonstration and hybrid reactor designs considered in the Magnetic Fusion Energy (MFE) Program were assessed. It was observed that a balanced program of design, analysis and experimentation would be required to develop, verify and qualify these designs and those of related hardware and equipment. To respond to the potential experimental requirements of the first-generation reactors (the EPRs and possibly the hybrid concept), the need for a helium test facility was identified. It was determined that this facility should have the capacity for recirculating 100,000 kg/hr of helium at 70 atm and 6000C and should have 3 MW of electrical power available for simulating neutron heating. No radioactive material or processes should be used to facilitate ''hands-on'' experimentation and development. The general types of testing anticipated in this facility would include: (1) thermal and coolant flow performance of the blanket and other components in the primary cooling circuit; (2) structural adequacy of the blanket and first wall including vibration considerations; (3) capability for accommodating safety/off-normal conditions. Existing facilities worldwide were surveyed. It was determined that a number of facilities exist in foreign nations for performing the anticipated experiments. However, no large helium gas flow loop exists within the USA. Consequently, it is recommended that a helium thermal-hydraulic blanket test facility be planned and build on a schedule that will meet the unique design development and verification needs of the fusion program. This report provides the rationale and preliminary scoping of the operational characteristics and requirements for such a facility

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

    International Nuclear Information System (INIS)

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

  20. Development of fusion blanket technology for the DEMO reactor.

    Science.gov (United States)

    Colling, B R; Monk, S D

    2012-07-01

    The viability of various materials and blanket designs for use in nuclear fusion reactors can be tested using computer simulations and as parts of the test blanket modules within the International Thermonuclear Experimental Reactor (ITER) facility. The work presented here focuses on blanket model simulations using the Monte Carlo simulation package MCNPX (Computational Physics Division Los Alamos National Laboratory, 2010) and FISPACT (Forrest, 2007) to evaluate the tritium breeding capability of a number of solid and liquid breeding materials. The liquid/molten salt breeders are found to have the higher tritium breeding ratio (TBR) and are to be considered for further analysis of the self sufficiency timing. PMID:22112596

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

    International Nuclear Information System (INIS)

    A description of a fusion breeding blanket concept using draw salt coolant and static 17Li-83Pb is presented. 17Li-83Pb 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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-08-01

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

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

    International Nuclear Information System (INIS)

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

  5. Hybrid reactor blankets for constant energy multiplication and flat power distribution

    International Nuclear Information System (INIS)

    Two blanket design difficulties are usually attributed to the blanket neutronic properties: high peak-to-average power density ratio distribution and the variation of the energy multiplication with burnup. This work shows that blankets can be designed to have a constant energy multiplication and a flat power distribution. These features are illustrated for light water hydrid reactor blankets

  6. 18 CFR 284.402 - Blanket marketing certificates.

    Science.gov (United States)

    2010-04-01

    ... effective for an affiliated marketer with respect to transactions involving affiliated pipelines when an affiliated pipeline receives its blanket certificate pursuant to § 284.284. (2) Should a marketer...

  7. Advanced Acoustic Blankets for Improved Aircraft Interior Noise Reduction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of the proposed Phase II research effort is to develop heterogeneous (HG) blankets for improved sound reduction in aircraft structures. Phase I...

  8. Safety and personnel access aspects of low activation fusion blankets

    International Nuclear Information System (INIS)

    The use of silicon carbide and carbon materials for structural applications in fusion reactor first wall and blanket regions has been proposed and a continuing effort spent on the development of the ceramics technology. The advantages identified are an extremely low induced radioactivity inventory, a high temperature operating capability, abundant raw material resource availability, and minimized plasma impurity effects. One of the unique features of the applications of these materials to fusion reactor blanket designs is that no alloying element is needed in order to assure the specified mechanical properties such as occurs in metal alloys. The major source of long term radioactivity in these materials is impurities. The impurity elements and their concentrations carried over to the blanket structure during fabrication can be minimized by proper fabrication procedures and techniques. The safety and personnel access aspects of such fusion blankets in conjunction with the impurity element concentration are the main subjects of this paper

  9. Lightweight IMM Multi-Junction Photovoltaic Flexible Blanket Assembly Project

    Data.gov (United States)

    National Aeronautics and Space Administration — DSS's recently completed successful NASA SBIR Phase 1 program has established a TRL 3/4 classification for an innovative IMM PV Integrated Modular Blanket Assembly...

  10. 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

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

    International Nuclear Information System (INIS)

    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, Li2TiO3 or Li2O 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

  12. Oxide fuel element and blanket element development programs. Quarterly progress report, April-June 1978

    International Nuclear Information System (INIS)

    Approval-in-principle has been granted for run beyond breach experiment XY-2, which will incorporate an F11A series rod. Fuel microstructures and operating parameters have been tabulated for 118 specimens from the F20 power to melt experiment. Retained gas measurements have been compiled indicating 36-50 μl/gm in this high power fuel. Topical report GEFR-00367 was prepared describing F20 results. Preparation of the Test Design Description for axial blanket experiment AB-1 is proceeding on schedule (for Cycle 2 irradiation). The safety analysis calculations, showing no fuel melting nor sodium boiling in design-basis upsets, have been completed

  13. Impact of prescribed burning on blanket peat hydrology

    OpenAIRE

    Holden, J; Palmer, SM; Johnston, K; Wearing, C.; Irvine, B; Brown, LE

    2015-01-01

    Fire is known to impact soil properties and hydrological flowpaths. However, the impact of prescribed vegetation burning on blanket peatland hydrology is poorly understood. We studied ten blanket peat headwater catchments. Five were subject to prescribed burning, while five were unburnt controls. Within the burnt catchments we studied plots where the last burn occurred ∼2 (B2), 4 (B4), 7 (B7) or greater than 10 years (B10+) prior to the start of measurements. These were compared with plots at...

  14. Blankets for tritium catalyzed deuterium (TCD) fusion reactors

    International Nuclear Information System (INIS)

    The TCD fusion fuel cycle - where the 3He from the D(D,n)3He reaction is transmuted, by neutron capture in the blanket, into tritium which is fed back to the plasma - was recently recognized as being potentially more promising than the Catalyzed Deuterium (Cat-D) fuel cycle for tokamak power reactors. It is the purpose of the present work to assess the feasibility of, and to identify promising directions for designing blankets for TCD fusion reactors

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

    International Nuclear Information System (INIS)

    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

  16. FIRST STEP blanket structure and fuel assembly design

    International Nuclear Information System (INIS)

    FIRST STEP (Fusion, Inertial, Reduced Requirement Systems Test for Special Nuclear Material, Tritium, and Energy Production) is an Inertial Confinement Fusion (ICF) plant designed to produce tritium, SNM, and energy using near-term technology. It is an integrated facility that will serve as a test bed for fusion power plant technology. The design of the blanket structure and blanket fuel assembly for wetted-wall FIRST STEP reactors is presented here

  17. Electrical connectors for blanket modules in ITER

    International Nuclear Information System (INIS)

    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

  18. Preliminary Safety Analysis of Korea HCSB Test Blanket Module

    International Nuclear Information System (INIS)

    A Helium Cooled Solid Breeder (HCSB) blanket has been considered as one of the promising blanket for the fusion power demonstration plant. Therefore HCSB Test Blanket Module (TBM) testing in ITER is the most important milestone for the development of the blanket of the DEMO plant. Korea has developed the HCSB TBM with some features such as graphite reflector and simplified flow passage. The objective of this study was to evaluate the thermal and structural integrity of the HCSB TBM under the hypothetical accidental conditions such as cooling pipe break in TBM. The safety analysis was performed under conservative conditions based on the TBM design, which can be assumed by the similarity of the safety analysis of the ITER shielding blanket. Transient analysis model was used to calculate the temperature distribution for Loss of Coolant Accident (LOCA). Simplified analysis conditions were a) simultaneous plasma shutdown and LOCA b) LOCA and then after FW temperature reaches 1150 deg. plasma shutdown. Helium circuit behavior during the different LOCA scenarios was also evaluated. Finally the design modifications based on the analysis result and the related R-and-D of the HCSB blanket design for the application in a DEMO reactor were mentioned. (author)

  19. Thermal hydraulics and mechanics research on fusion blanket system

    International Nuclear Information System (INIS)

    In-vessel components such as Blanket and Divertor in a fusion reactor have a function of exhausting high heat and particle loads in order to maintain the structural soundness of the reactor. In the International Thermonuclear Experimental Reactor called ITER, build by ITER Organization under the framework of collaboration of seven parties including Japan, there are two kinds of blanket systems will be install. One is a shield blanket, which consists of a first wall (FW) and a block module shielding against neutron flux to a vacuum chamber and a superconducting magnet system. The other blanket system is called as a Test Blanket Module (TBM). TBM is a kind of prototype blanket for a fusion power plant and has functions of breeding of tritium (T) and extraction of energy from fusion plasma. TBM consists of FW and T-breeding / neutron (n)-multiplier zone. A concept of TBM developed by JAEA is water-cooled pebble-bed type, which means that FW and other structures are cooled by pressurized high temperature water and T-breeding / n-multiplier zone consists of multiple layers of pebble bed made of T-breeding and n-multiplier material. This paper describes the status of R and Ds on FW and pebble beds from the view of thermo-hydraulics and mechanics. (author)

  20. Canadian fusion breeder blanket program: Irradiation facilities at chalk river*1

    Science.gov (United States)

    Hastings, I. J.; Burton, D. G.; Celli, A.; Delaney, R. D.; Fehrenbach, P. J.; Howe, L. M.; Larson, L. L.; MacEwen, S. R.; Miller, J. M.; Naeem, T. A.; Sawicki, J. A.; Swanson, M. L.; Verrall, R. A.; Zee, R. H.

    1986-11-01

    The major irradiation facility at Chalk River Nuclear Laboratories (CRNL) is the NRU research reactor. Both unvented and vented capsule experiments on candidate blanket ceramics can be performed. In the unvented tests, tritium release data (HT-to-HTO ratio, tritium retention) are obtained by post-irradiation heating of the breeder ceramic in the presence of a sweep gas. Four tests have been completed on Li 2O and LiAlO 2. Effects of sweep gas composition, extraction vessel material and ceramic properties have been determined. Two unvented irradiations under the BEATRIX international breeder exchange program have been completed; analysis is underway. The vented tests involve long-term irradiation of candidate blanket materials. CRITIC-I, scheduled for mid-1986 under BEATRIX, will examine ANL-fabricated Li 2O in a six-month irradiation at 700-1200 K, varying sweep gas composition, with on-line HT/HTO measurement. Additionally, accelerator simulation techniques are available, using 70 kV and 2.0 MV mass separators, a 2.5 MV Van de Graaff accelerator and a tandem accelerator super-conducting cyclotron, the latter allowing irradiation with protons, deuterons or helium at 18-20 MeV.

  1. The integrated-blanket-coil concept applied to the poloidal field and blanket systems of a tokamak reactor

    International Nuclear Information System (INIS)

    A novel concept is proposed for combining the blanket and coil functions of a fusion reactor into a single component. This concept, designated the ''integrated-blanket-coil'' (IBC) concept, is applied to the poloidal field and blanket systems of a tokamak reactor. An examination of resistive power losses in the IBC suggests that these losses can be limited to 10% of the fusion thermal power. By assuming a sandwich construction for the IBC walls, magnetohydrodynamic (MHD)-induced pressure drops and associated pressure stresses are shown to be modest and well below design limits. For the stainless steel reference case examined, the MHD-induced pressure drop was estimated to be about 1/3 MPa and the associated primary membrane stress was estimated to be about 47 MPa. The preliminary analyses indicate that the IBC concept offers promise as a means for making fusion reactors more compact by combining blanket and coil functions in a single component

  2. Degrading the Plutonium Produced in Fast Breeder Reactor Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jor-Shan; Kuno, Yusuke [Tokyo University, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 113-8656 (Japan)

    2009-06-15

    Plutonium quality, defined as the plutonium isotopic composition, is an important measure for proliferation-resistance (PR) of a nuclear energy system. The quality of the plutonium produced in the blanket assemblies of a fast breeder reactor could be as good as or better than the weapons-grade (WG). The presence of such good quality plutonium is a proliferation concern. There are various options to degrade the plutonium produced in the breeder blanket. The obvious one is to blend the blanket plutonium with those produced from the reactor core during reprocessing. Other options try to prevent the generation of good quality plutonium (Pu). The Protected Plutonium Production (P{sup 3}) Project proposed by Tokyo Institute of Technology (TIT)1,2,3 advocates the doping of certain amount of neptunium (Np), or americium (Am) in fresh blanket fuel for irradiation. The increased production of {sup 238}Pu, {sup 240}Pu and {sup 242}Pu by neutron capture in {sup 237}Np and Am would degrade the blanket plutonium. However, as {sup 237}Np is a controlled material according to IAEA, its use as doping material in fresh blanket fuel presents a concern for nuclear proliferation. In addition, the fabrication of fresh blanket fuel with inclusion of americium would be complicated due to the emission of intense low-energy gamma radiation from {sup 241}Am. Am is normally accompanied by Cm since the separation of those 2 elements is very difficult. Fuel containing both Am and Cm may make Safeguards measurement difficult. A variation would be doping the fresh blanket fuel with minor actinide (e.g., a group of neptunium, americium, and curium), or with separated reactor-grade (RG) plutonium. The drawback of such schemes would be the need for glove boxes in fresh blanket fuel fabrication. It is possible to fuel the breeder blankets with recycled (reprocessed) uranium oxide. The recycled uranium, recovered from reprocessing, contains {sup 236}U, which when irradiated in the blanket would

  3. Safety Evaluation of the EVOLVE Blanket Concept

    International Nuclear Information System (INIS)

    This article summarizes the results of the safety evaluation of the Evaporation of Lithium and Vapor Extraction (EVOLVE) W-alloy first wall (FW) and blanket concept. We have analyzed the EVOLVE design response during a confinement bypass accident. A confinement bypass accident was chosen because, based on previous safety studies, this accident can produce environmental releases by breaching the primary radioactive confinement boundary of EVOLVE, which is the EVOLVE vacuum vessel (VV). As a consequence of a bypass accident, air from a room adjoining the reactor enters the plasma chamber by way of a failed VV port. This air reacts with the high temperature metals inside of the VV to release energy in the case of a lithium spill, or to mobilize radioactive material by oxidation, and then transport this material to the environment by natural convection airflow through the failed VV port. We use the MELCOR code to analyze the response of EVOLVE during this accident. Based on these results, the EVOLVE concept can meet the no-evacuation dose goal set by the DOE Fusion Safety Standard if the EVOLVE confinement building ventilation system is closed within two hours of the onset of this accident

  4. Current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

    Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phase of the Li-LiH, Li-LiD, and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li3N, Li2O, and Li2C2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g., Li--Al and Li--Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li--M alloys can be estimated from lithium activity data for these alloys

  5. Flow characteristics of the Cascade granular blanket

    International Nuclear Information System (INIS)

    Analysis of a single granule on a rotating cone shows that for the 350 half-angle, double-cone-shaped Cascade chamber, blanket granules will stay against the chamber wall if the rotational speed is 50 rpm or greater. The granules move axially down the wall with a slight (5-mm or less) sinusoidal oscillation in the circumferential direction. Granule chute-flow experiments confirm that two-layered flow can be obtained when the chute is inclined slightly above the granular material angle of repose. The top surface layer is thin and fast moving (supercritical flow). A thick bottom layer moves more slowly (subcritical flow controlled at the exit) with a velocity that increases with distance from the bottom of the chute. This is a desirable velocity profile because in the Cascade chamber about one-third of the fusion energy is deposited in the form of x rays and fusion-fuel-pellet debris in the top surface (inner-radius) layer

  6. Target/Blanket Design for the Accelerator Production of Tritium Plant

    International Nuclear Information System (INIS)

    The Accelerator Production of Tritium Target/Blanket (T/B) system is comprised of an assembly of tritium-producing modules supported by safety, heat removal, shielding, and retargeting systems. The T/B assembly produces tritium using a high-energy proton beam, a tungsten/lead spallation neutron source and 3He gas as the tritium-producing feedstock. The supporting heat removal systems remove the heat deposited by the proton beam during both normal and off-normal conditions. The shielding protects workers from ionizing radiation, and the retargeting systems remove and replace components that have reached their end of life. All systems reside within the T/B building, which is located at the end of a linear accelerator. For the nominal production mode, protons are accelerated to an energy of 1030 MeV at a current of 100 mA and are directed onto the T/B assembly. The protons are expanded to a 0.19- x 1.9-m beam spot before striking a centrally located tungsten neutron source. A surrounding lead blanket produces additional neutrons from scattered high-energy particles. A total of 27 neutrons are produced per incident proton. Tritium is produced by neutron capture in 3He gas that is contained in aluminum tubes throughout the blanket. The 3He/tritium mixture is removed on a semi-continuous basis for purification in an adjacent Tritium Separation Facility. Systems and components are designed with safety as a primary consideration to minimize risk to the workers and the public. Materials and component designs were chosen based on the experiences of operating spallation neutron sources that have been designed and built for the neutron science community. An extensive engineering development and demonstration program provides detailed information for the completion of the design

  7. Impact of blanket tritium against the tritium plant of fusion reactor

    International Nuclear Information System (INIS)

    The breeder blanket and the blanket tritium recovery system are tested using test blanket modules during ITER campaign. And then, these are integrated with the tritium plant for the first time at a prototype reactor after ITER. In this work, impact to the tritium plant by integration of the solid breeder blanket was discussed. The method of tritium extraction from the blanket and the choice of the process for breeder blanket interface should be discussed not only from the viewpoint of tritium release but also from the viewpoint of the load of processing. (author)

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

    International Nuclear Information System (INIS)

    This report presents the conceptual design and preliminary feasibility assessment for the hybrid blanket and power conversion system of the Mirror Hybrid Fusion-Fission Reactor. Existing gas-cooled fission reactor technology is directly applicable to the Mirror Hybrid Reactor. There are a number of aspects of the present conceptual design that require further design and analysis effort. The blanket and power conversion system operating parameters have not been optimized. The method of supporting the blanket modules and the interface between these modules and the primary loop helium ducting will require further design work. The means of support and containment of the primary loop components must be studied. Nevertheless, in general, the conceptual design appears quite feasible

  9. Recent MHD activities for blanket analysis at UPC

    International Nuclear Information System (INIS)

    In the frame of fusion reactor design definition, the detailed analysis of main flow parameters in liquid metal blankets is of utmost interest. Critical aspects are (1) tritium inventories and permeation rates, (2) heat extraction and maximum temperatures for material specifications and (3) MHD pressure drops. The aim of GREENER research group at UPC is to develop a CFD code, based on the OpenFOAM toolbox, able to deal with the main phenomena occurring at blanket channels (MHD coupling, heat transfer and tritium transport) and capable to quantify the above mentioned critical aspects. In parallel, CIMNE research group is developing its own MHD code, mainly focused on algorithm optimisation. The paper summarises the developing tools at each research group and compares their behaviour in a validation step using analytical solutions. In order to expose the applicability of the codes, some simulation results related with the HCLL-ITER/TBM blanket are exposed. Special focus is made on buoyancy flows in U-shaped channels and multi channel effect. Moreover, a preliminary flow analysis related with vertical banana-shape liquid metal channels is discussed, related with a new blanket design that is being considered as a progress of conceptual design refinement of dual-coolant liquid metal blankets (DEMO specifications).

  10. Neutronics experiments for DEMO blanket at JAERI/FNS

    International Nuclear Information System (INIS)

    In nuclear fusion DEMO reactor, the blanket is required to provide the tritium breeding ratio (TBR) of more than unity by the neutron induced reaction in lithium in the blanket. To provide the TBR of more than unity is critical issue in the development of the blanket. Also in order to develop the blanket with low activation level, the evaluation of the induced activity with high precision is required by taking into account the sequential reactions induced by secondary charged particles. In order to evaluate these issues experimentally, neutronics experiments have been performed by using DT neutrons at JAERI/FNS. From the results of TBR experiment by using the mockup relevant to the DEMO blanket with multilayered structure composed of Be, Li2TiO3 and F82H, it was clarified that the TPR can be evaluated within 10 % uncertainty by using the Monte Carlo calculation. From the results of sequential reactions experiment for the test specimens simulating the cooling water pipe, it was found that the effective cross-sections due to the sequential reactions were increased in a form close to an exponential curve in the cooling water pipe with reducing the distance to the water. (author)

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

    International Nuclear Information System (INIS)

    The European Union has been engaged since 1989 in a programme to develop tritium breeding blankets for application in a fusion power reactor. There are four blanket concepts under development. Two of them use lithium ceramics, the other two concepts employ an eutectic lead-lithium alloy (Pb-17Li) as breeder material. The two most promising concepts were to select in 1995 for further development. In order to prepare the selection, a Blanket Concept Selection Exercise (BCSE) has been inititated by the participating associations under the auspices of the European Commission. This BCSE has been performed in 14 working groups which, in a comparative evaluation of the four blanket concepts, addressed specific fields. The working group safety addressed the safety implications. This report describes the methodology adopted, the safety issues identified, their comparative evaluation for the four concepts, and the results and conclusions of the working group to be entered into the overall evaluation. There, the results from all 14 working groups have been combined to yield a final ranking as a basis for the selection. In summary, the safety assessment showed that the four European blanket concepts can be considered as equivalent in terms of the safety rating adopted, each concept, however, rendering safety concerns of different quality in different areas which are substantiated in this report. (orig.)

  12. Thermomechanics analysis and optimization for high power density blanket

    International Nuclear Information System (INIS)

    Thermomechanics analysis, i.e. steady thermal analysis and steady thermal stress analysis have been carried out for a high power density blanket. The Fusion Experimental Breeder (FEB) is adopted as the reference reactor. The parts for the blanket module in Pro/ENGINEER were created, then turn to Pro/MECHANICA functionality for thermomechanics analysis. During analysis, the distribution of the power density in the blanket was optimized to be more flat, the arched curvature and rounds of the cooling tube panels were optimized to less stiffness, and the boundary condition at the interface of helium cooling tube panel and manifold chamber was optimized, which is reasonable by using advanced welding processes with electron beam or laser beam in a single pass. To the end, a maximum temperature Tm 350 degree C and a maximum shear stress τm 80 MPa for the helium cooling panels have been shown in the calculations. (authors)

  13. Direct Lit Electrolysis In A Metallic Lithium Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Colon-Mercado, H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Babineau, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Elvington, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Garcia-Diaz, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Teprovich, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Vaquer, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-10-13

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

  14. Development of the water cooled lithium lead blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Aiello, G.; Jonquères, N. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Li Puma, A. [CEA-Saclay, DEN/DANS/DM2S/SERMA/LPEC, 91191 Gif Sur Yvette Cedex (France); Morin, A.; Rampal, G. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France)

    2014-10-15

    Highlights: • The WCLL blanket design has been modified to adapt it to the 2012 EFDA DEMO specifications. • Preliminary CAD design of the equatorial outboard module of the WCLL blanket has been developed for DEMO. • Finite elements analyses have been carried out in order to assess the module thermal behavior in the straight part of the module. - Abstract: The water cooled lithium lead (WCLL) blanket, based on near-future technology requiring small extrapolation from present-day knowledge both on physical and technological aspect, is one of the breeding blanket concepts considered as possible candidates for the EU DEMOnstration power plant. In 2012, the EFDA agency issued new specifications for DEMO: this paper describes the work performed to adapt the WCLL blanket design to those specifications. Relatively small modules with straight surfaces are attached to a common Back Supporting Structure housing feeding pipes. Each module features reduced activation ferritic-martensitic steel as structural material, liquid Lithium-Lead as breeder, neutron multiplier and carrier. Water at typical Pressurized Water Reactors (PWR) conditions is chosen as coolant. A preliminary design of the equatorial outboard module has been achieved. Finite elements analyses have been carried out in order to assess the module thermal behavior. Two First Wall (FW) concepts have been proposed, one favoring the thermal efficiency, the other favoring the manufacturability. The Breeding Zone has been designed with C-shaped Double-Walled Tubes in order to minimize the Water/Pb-15.7Li interaction likelihood. The priorities for further development of the WCLL blanket concept are identified in the paper.

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

    International Nuclear Information System (INIS)

    The potential sources of tritium contamination of the helium coolant of ceramic breeder blankets have previously been evaluated for the specific case of the European BIT DEMO blanket. This confirmed that the control of tritium losses to the steam circuit is a critical issue which demands development concerning (a) permeation barriers, (b) tritium recovery processes maintaining a very low tritium activity in the coolant, and (c) control of the coolant chemistry. The specifications of these developments required the evaluation of the tritium losses through the steam generators, and includes the definition of their operating conditions by thermodynamic cycle calculations, and their thermal-hydraulic design. For both tasks, specific computer tools were developed. The geometry obtained, the surface area and the temperature profiles along the heat-exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam-oxidized Incoloy 800 austenitic stainless steel was identified as the best-suited existing material. Our results indicate that in nominal steady-state operation the tritium escape into the steam cycle could be restricted to less than 10 Ci per day. The conditions for this are specified, but their feasibility demands, in particular, the resolution of certain gas chemistry problems, and their validation in the more stringent environment of an operating blanket. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) was identified as bearing a large tritium release potential. The problems associated with such transients are discussed and possible solutions are proposed. (orig.)

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

    International Nuclear Information System (INIS)

    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·m2 for steels and 5,0x10-6 - 5,0x10-5 Ohm·m2 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)

  17. Nuclear Performance Analyses for HCPB Test Blanket Modules in ITER

    International Nuclear Information System (INIS)

    The Helium-Cooled Pebble Bed (HCPB) blanket is one of two breeder blanket concepts developed in the framework of the European Fusion Technology Programme for performance tests in ITER. The related efforts currently focus on the design optimisation of suitable Test Blanket Modules (TBM) and associated R-and-D activities. Four different HCPB TBM types are considered for addressing issues related to (i) electromagnetic transients (EM), (ii) neutronics and Tritium (NT), (iii) thermo-mechanical properties of the pebble beds (TM), and (iv) the integral performance of the blanket module (Plant Integration, PI). The lay-out of the NT and the PI modules has been entirely revised to represent the latest HCPB breeder blanket concept for fusion power reactors. A HCPB TBM consists of a steel box with an internal stiffening grid and small breeder units. The stiffening grid forms radially running open cells accommodating the breeder units (BU). The BU consists of a back plate with attached breeder canisters providing space for the breeder pebble beds. The space between the canisters and the stiffening plates is filled with Beryllium pebbles for the neutron multiplication. The latest design assumes two vertically arranged breeder containers per BU with a toroidal bed height of 10 and 24 mm, for NT and PI modules, respectively. Li4SiO4 is assumed as breeder material at 6Li enrichment levels between 40 at % (NT) and 90 at % (PI). This work is devoted to the neutronic, shielding and activation analyses performed recently for NT and PI variants of the HCPB TBM in ITER. The analyses are based on three-dimensional neutronic and activation calculations making use of a 20 degree torus sector model of ITER developed for Monte Carlo calculations with the MCNP code. The model includes a proper representation of the horizontal ITER test blanket port, the water cooled support frame with two integrated HCPB blanket test modules, the radiation shield and the port environment. Monte Carlo

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

    International Nuclear Information System (INIS)

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

  19. APT Blanket Safety Analysis: Counter Current Flow Limitation for Cavity Spaces

    International Nuclear Information System (INIS)

    The thermal-hydraulic modeling aspects for the APT blanket system have been broken up into two basic modeling components: (1) the blanket system and (2) the cavity flood system. In most cases these systems are modeled separately. This separate study for the coolability of the blanket modules can also be used to establish/evaluate a functional design requirement on gap size between the blanket modules

  20. Maximizing fluence rate and field uniformity of light blanket for intraoperative PDT

    OpenAIRE

    LIANG, XING; Kundu, Palak; Finlay, Jarod; Goodwin, Michael; Zhu, Timothy C.

    2012-01-01

    A light blanket is designed with a system of cylindrically diffusing optical fibers, which are spirally oriented. This 25×30 cm rectangular light blanket is capable of providing uniform illumination during intraoperative photodynamic therapy. The flexibility of the blanket proves to be extremely beneficial when conforming to the anatomical structures of the patient being treated. Previous tests of light distribution from the blanket have shown significant loss of intensity with the length of ...

  1. Neutronic analysis of a dual He/LiPb coolant breeding blanket for DEMO

    OpenAIRE

    Catalán, J.P.; Ogando Serrano, Francisco; Sanz Gonzalo, Javier; Palermo, I.; Veredas, G.; Gómez Ros, J. M.; Sedano, L

    2010-01-01

    A conceptual design of a DEMO fusion reactor is being developed under the Spanish Breeding Blanket Technology Programme: TECNO_FUS based on a He/LiPb dual coolant blanket as reference design option. The following issues have been analyzed to address the demonstration of the neutronic reliability of this conceptual blanket design: power amplification capacity of the blanket, tritium breeding capability for fuel self-sufficiency, power deposition due to nuclear heating in superconducting coils ...

  2. 78 FR 30295 - Constellation Energy Commoditiesgroup, Inc., ENI USA Gas Marketing LLC, Sequent Energy Canada...

    Science.gov (United States)

    2013-05-22

    ..., and 13-32-LNG] Constellation Energy Commoditiesgroup, Inc., ENI USA Gas Marketing LLC, Sequent Energy........... Constellation Order vacating blanket Energy authority to import/export Commodities natural gas from/to...

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

    International Nuclear Information System (INIS)

    Full text: One of the main requirements of advanced nuclear-power engineering is inherent safety of power installations. It initiates R and D of heavy liquid metals (lead, lead- bismuth eutectic) application in fission reactors as substitute of sodium. The same requirement makes advisable R and D of the lead and lead-bismuth eutectic application in blanket of fusion reactors as substitute of lithium. High magnetohydrodynamic pressure drop is an important issue for liquid metal blanket concepts. To decrease MHD-resistance authors propose to form electro-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 electro-insulating coatings characteristics rd (r - specific resistance of coatings, d - thickness) is ∼ 10-5Ω·m2 for steels and 5, 0x10-6 - 5, 0x10-5Ω·m2 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 electro-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, steam generators and another 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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. ANL ITER high-heat-flux blanket-module heat transfer experiments. Fusion Power Program

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. First-wall/blanket materials selection for STARFIRE tokamak reactor

    International Nuclear Information System (INIS)

    The development of the reference STARFIRE first-wall/blanket design involved numerous trade-offs in the materials selection process for the breeding material, coolant structure, neutron multiplier, and reflector. The major parameters and properties that impact materials selection and design criteria are reviewed

  8. Technical issues for beryllium use in fusion blanket applications

    International Nuclear Information System (INIS)

    Beryllium is an excellent non-fissioning neutron multiplier for fusion breeder and fusion electric blanket applications. This report is a compilation of information related to the use of beryllium with primary emphasis on the fusion breeder application. Beryllium resources, production, fabrication, properties, radiation damage and activation are discussed. A new theoretical model for beryllium swelling is presented

  9. Fusion blanket testing in MFTF-α + T

    International Nuclear Information System (INIS)

    The Mirror Fusion Test Facility-α + T (MFTF-α + T) is an upgraded version of the current MFTF-B test facility at Lawrence Livermore National Laboratory, and is designed for near-term fusion-technology-integrated tests at a neutron flux of 2 MW/m2. Currently, the fusion community is screening blanket and related issues to determine which ones can be addressed using MFTF-α + T. In this work, the minimum testing needs to address these issues are identified for the liquid-metal-cooled blanket and the solid-breeder blanket. Based on the testing needs and on the MFTF-α + T capability, a test plan is proposed for three options; each option covers a six to seven year testing phase. The options reflect the unresolved question of whether to place the research and development (R and D) emphasis on liquid-metal or solid-breeder blankets. In each case, most of the issues discussed can be addressed to a reasonable extent in MFTF-α+T

  10. Burnable Poison optimization for Seed-Blanket Cores

    International Nuclear Information System (INIS)

    The main objective of the Seed and Blanket Units (SBU) core designs is to reduce production of Pu and long-term toxicity of the spent PWR fuel. The SBU concept assumes a heterogeneous seed-blanket fuel assembly, with spatial separation of the U and Th parts of the fuel. In the SBU assembly the Seed fuel is an alloy of 20% enriched metallic Uranium and zircaloy, the Blanket fuel is ThO2 mixed with about 13% of 12.2% enriched UO2. The Uranium is included in the mix in order to increase the BOL power in the Th pins and dilute the bred U233 isotope to avoid proliferation concerns. The 108 Seed fuel rods are located in the central region of the assembly and surrounded by 156 blanket rods. The use of metallic fuel in the Seed enables high density of fissile material. The U-Zr alloy also has a higher thermal conductivity than UO2, although this advantage is partially offset by the low melting temperature of metallic fuels. More importantly, compared with oxide fuel, the radiation induced creep and swelling phenomena are more pronounced in metallic fuel at elevated temperatures due to the loss of its crystallographic state. This loss occurs at a rather low temperature of 6600 C, and in U-Zr alloys at 6160 C. For this reason, reduction of the local pin power peaking in the assembly is particularly important

  11. Activation analysis of coolant water in ITER blanket and divertor

    International Nuclear Information System (INIS)

    Coolant water in blankets and divertor cassettes will be activated by neutrons during ITER operation. 16N and 17N are determined to be the most important activation products in the coolant water in terms of their impact on ITER design and performance. In this study, the geometry of cooling channels in blanket module 4 was described precisely in the ITER neutronics model ‘Alite-4’ based on the latest CAD model converted using MCAM developed by FDS Team. The 16N and 17N concentration distribution in the blanket, divertor cassette and their primary heat transport systems were calculated by MCNP with data library FENDL2.1. The activation of cooling pipes induced 17N decay neutrons was analyzed and compared with that induced by fusion neutrons, using FISPACT-2007 with data library EAF-2007. The outlet concentration of blanket and divertor cooling systems was 1.37 × 1010 nuclide/cm3 and 1.05 × 1010 nuclide/cm3 of 16N, 8.93 × 106 nuclide/cm3 and 0.33 × 105 nuclide/cm3 of 17N. The decay gamma-rays from 16N in activated water could be a problem for cryogenic equipments inside the cryostat. Near the cryostat, the activation of pipes from 17N decay neutrons was much lower than that from fusion neutrons.

  12. 18 CFR 33.1 - Applicability, definitions, and blanket authorizations.

    Science.gov (United States)

    2010-04-01

    ... the outstanding voting securities; or (iii) Any security of a subsidiary company within the holding... company subsidiary in connection with such acquisition. (4) A holding company granted blanket... subsidiaries, or associate companies within the holding company system has captive customers in the...

  13. LMFBR blanket assembly heat transfer and hydraulic test data evaluation

    International Nuclear Information System (INIS)

    The USA Test Program for characterization of breeder reactor blanket T and H performance is providing a data base for improved confidence in the design tools employed. Pressure drop tests with wire wrapped rod bundles having a 1.08 triangular pitch to diameter ratio and 4 inch (10 cm) wire wrap lead using water, sodium and air have defined a smooth, continuous, single-valued friction factor versus Reynolds number correlation. This eliminates a possible source of flow instability. The rod bundle temperature rise profiles measured in the heat transfer tests using a prototypic blanket rod bundle agrees in magnitude and shape with the predictions of the marching type sub-channel codes currently employed in blanket subchannel analysis. The low flow test data demonstrates increasing buoyancy induced flows in the lower Reynolds number flow regime. This and the remaining test data will supply a base for calibration of the mixing momentum exchange and conduction factors employed in the subchannel analysis codes; which will contribute to the confidence of the blanket design predictions and reduce the uncertainties which are commonly expressed as hot channel/spot factors

  14. Nuclear analyses of Indian LLCB test blanket system in ITER

    International Nuclear Information System (INIS)

    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 development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radio-active waste management, equipments maintenance and replacement strategies and nuclear safety. To predict the nuclear behaviour of LLCB test blanket module in ITER, nuclear responses like tritium production, nuclear heating, neutron fluxes and radiation damages are estimated. As a part of ITER machine, LLCB TBS has to follow certain nuclear shielding requirements i.e. shutdown dose rates should not exceed the defined limits in ITER premises (inside bio-shield ∼100 μSv/hr after 12 days cooling and outside bio-shield ∼10 μSv/hr after 1 day cooling). Hence nuclear analyses are performed to assess and optimize the shielding capability of LLCB TBS inside and outside bio-shield. To state the radio-activity level of LLCB TBS components which support the rad-waste and safety assessment, nuclear activation analyses are executed. 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.1). The paper describes comprehensive nuclear performance of LLCB TBS in ITER. (author)

  15. Development of simulator for remote handling system of ITER blanket

    International Nuclear Information System (INIS)

    The maintenance activity in the ITER has to be performed remotely because 14 MeV neutron caused by fusion reaction induces activation of structural material and emission of gamma ray. In general, it is one of the most critical issues to avoid collision between the remote maintenance system and in-vessel components. Therefore, the visual information in the vacuum vessel is required strongly to understand arrangement of these devices and components. However, there is a limitation of arrangement of viewing cameras in the vessel because of high intensity of gamma ray. It is expected that enough numbers of cameras and lights are not available because of arrangement restriction. Furthermore, visibility of the interested area such as the contacting part is frequently disturbed by the devices and components, thus it is difficult to recognize relative position between the devices and components only by visual information even if enough cameras and lights are equipped. From these reasons, the simulator to recognize the positions of each devices and components is indispensable for remote handling systems in fusion reactors. The authors have been developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robot simulation software ''ENVISION''. The simulator is connected to the control system of the manipulator which was developed as a part of the blanket maintenance system in the EDA and can reconstruct the positions of the manipulator and the blanket module using the position data of the motors through the LAN. In addition, it can provide virtual visual information, such as the connecting operation behind the blanket module with making the module transparent on the screen. It can be used also for checking the maintenance sequence before the actual operation. The developed simulator will be modified further adding other necessary functions and finally completed as a prototype of the actual simulator for the blanket remote handling system

  16. Integrated-blanket-coil (IBC) concept applied to the poloidal field and blanket systems of a tokamak reactor

    International Nuclear Information System (INIS)

    A novel concept is proposed for combining the blanket and coil functions of a fusion reactor into a single component. This concept, designated the integrated-blanket-coil (IBC) concept, is applied to the poloidal field and blanket systems of a Tokamak reactor. An examination of resistive power losses in the IBC suggests that these losses can be limited to less than or equal to 10% of the fusion thermal power. By assuming a sandwich construction for the IBC walls, MHD-induced pressure drops and associated pressure stresses are shown to be modest and well below design limits. For the stainless steel reference case examined in this paper, the MHD-induced pressure drop was estimated to be approx. 1/3 MPa and the associated primary membrane stress was estimated to be approx. 47 MPa. The preliminary analyses presented in this paper indicate that the IBC concept offers promise as a means for making fusion reactors more compact by combining blanket and coils functions in a single component

  17. Conceptual design of a water cooled breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Pu, Yong; Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Li, Jia; Peng, ChangHong [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Ma, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Chen, Lei [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2014-10-15

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by {sup 6}Li(n,α)T reaction. Li{sub 2}TiO{sub 3} pebbles and Be{sub 12}Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li{sub 2}TiO{sub 3} and Be{sub 12}Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be{sub 12}Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option

  18. Conceptual design of a water cooled breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by 6Li(n,α)T reaction. Li2TiO3 pebbles and Be12Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li2TiO3 and Be12Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be12Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into

  19. Hydraulics of laboratory and full-scale upflow anaerobic sludge blanket (UASB) reactors.

    Science.gov (United States)

    Batstone, D J; Hernandez, J L A; Schmidt, J E

    2005-08-01

    Laboratory-scale upflow anaerobic sludge blanket (UASB) reactors are often used as test platforms to evaluate full-scale applications. However, for a given volume specific hydraulic loading rate and geometry, the gas and liquid flows increase proportionally with the cube root of volume. In this communication, we demonstrate that a laboratory-scale reactor had plug-flow hydraulics, while a full-scale reactor had mixed flow hydraulics. The laboratory-scale reactor could be modeled using an existing biochemical model, and parameters identified, but because of computational speed with plug-flow hydraulics, mixed systems are instead recommended for parameter identification studies. Because of the scaling issues identified, operational data should not be directly projected from laboratory-scale results to the full-scale design. PMID:15977253

  20. Double ionization chamber as neutron flux monitor and for tritium breeding studies in fusion blanket experiments

    International Nuclear Information System (INIS)

    A new method for direct determination of tritium breeding, specially suited to thermal blankets is presented. The method can provide true tritium events even in the presence of a reasonable fraction of high energy neutrons. There exists no transfer or recovery losses and also the method exhibits good efficiency, in addition to being an on-line one. The detector consist of two identical chambers separated by a common earthing ring which supports a thin nickel foil loaded with Li6F covered with a thin gold layer acting as a conducting electrode. Two nickel discs are held symmetrically on either side of the central electrode as charge collectors. The chambers can be filled with a suitable inert gas. The system response to thermal neutrons was calculated by both analytical as well Monte Carlo method and is in good agreement with experimental results

  1. Tritium transport analysis in HCPB DEMO blanket with the FUS-TPC Code (KIT Scientific Reports ; 7642)

    OpenAIRE

    Franza, Fabrizio

    2013-01-01

    In thermonuclear fusion reactors, the fuel is an high temperature deuterium-tritium plasma, in which tritium is bred by lithium isotopes present inside solid ceramic breeder (e.g. Li-Orthosilicate) or inside liquid eutectic alloys (e.g. Pb-16Li alloy). In the breeding areas a significant fraction of the tritium produced is extracted out from the Breeding Zone by the He gas purging the breeding ceramic in the Helium Cooled Pebble Bed (HCPB) blanket concept or transported in solution by the owi...

  2. Engineering test station for TFTR blanket module experiments

    International Nuclear Information System (INIS)

    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

  3. Deposition and properties of blanket-W using silane reduction

    International Nuclear Information System (INIS)

    The authors have studied the deposition and properties of silane-reduced blanket tungsten films, and their suitability as interconnects. Their reflectance and surface morphology on Si substrates are higher than those of H2 - reduced films. This is suited for submicron lithography. Resistivity becomes 8 μΩ · cm, which is the same as that of H2-reduced films. Surface morphology and resistivity degrade when β-W is formed. Since step coverage of interconnects in half-micron contact holes is improved with selective deposition, a combination of silane-reduced selective and blanket tungsten deposition allows both contact and wiring metallization. This combination is also desirable in multilevel interconnects on ULSIs

  4. Nuclear performance analyses for HCPB test blanket modules in ITER

    International Nuclear Information System (INIS)

    Neutronic, shielding and activation analyses have been performed for recent design variants of the Helium Cooled Pebble Bed (HCPB) test blanket module (TBM) in ITER on the basis of 3D Monte Carlo calculations. The main objective has been to assess and optimise the nuclear performance of the HCPB test blanket modules in terms of the tritium generation, the nuclear heating and the radiation shielding and provide, among others, the data required for the engineering design of the test modules. The shielding efficiency of the TBM system was shown to be sufficient to allow access of work personnel to the port extension after a waiting time of 10 days after shut down as required by ITER. The activation analyses provided the afterheat and activation data for quality assured safety analyses assuming a representative irradiation scenario

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

    International Nuclear Information System (INIS)

    In the frame of the recent CEA studies aiming at the development, evaluation and comparison of solid breeder blanket concepts in view of their adaptation to NET, and the evaluation of specific questions related to the first wall design, the present paper examines first the performances of a helium cooled toroidal blanket design for NET, based on innovative Beryllium/Ceramics breeder rod elements. Neutronic and thermo-mechanical optimisation converges on a concept featured by a breeding capability in excess of 1.2, a reasonnable pumping power of 1 % and a narrow breeder temperature range (470 +- 30 0C of the breeder), the latter being largely independent of the power level. The final section of the paper is devoted to the evaluation of the heat load poloidal distribution and to the irradiation effects on first wall structural materials

  6. Markov Blanket Ranking using Kernel-based Conditional Dependence Measures

    OpenAIRE

    Strobl, Eric V.; Visweswaran, Shyam

    2014-01-01

    Developing feature selection algorithms that move beyond a pure correlational to a more causal analysis of observational data is an important problem in the sciences. Several algorithms attempt to do so by discovering the Markov blanket of a target, but they all contain a forward selection step which variables must pass in order to be included in the conditioning set. As a result, these algorithms may not consider all possible conditional multivariate combinations. We improve on this limitati...

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

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

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

  8. Helium-3 blankets for tritium breeding in fusion reactors

    Science.gov (United States)

    Steiner, Don; Embrechts, Mark; Varsamis, Georgios; Vesey, Roger; Gierszewski, Paul

    1988-01-01

    It is concluded that He-3 blankets offers considerable promise for tritium breeding in fusion reactors: good breeding potential, low operational risk, and attractive safety features. The availability of He-3 resources is the key issue for this concept. There is sufficient He-3 from decay of military stockpiles to meet the International Thermonuclear Experimental Reactor needs. Extraterrestrial sources of He-3 would be required for a fusion power economy.

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

    International Nuclear Information System (INIS)

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li2O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N2) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concepts are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li2O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

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

    International Nuclear Information System (INIS)

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li2O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N2) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li2O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

  11. Development of insulating coatings for liquid metal blankets

    International Nuclear Information System (INIS)

    It is shown that self-cooled liquid metal blankets are feasible only with electrically insulating coatings at the duct walls. The requirements on the insulation properties are estimated by simple analytical models. Candidate insulator materials are selected based on insulating properties and thermodynamic consideration. Different fabrication technologies for insulating coatings are described. The status of the knowledge on the most crucial feasibility issue, the degradation of the resisivity under irradiation, is reviewed

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

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

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

  14. MFTF-B Upgrade for blanket-technology testing

    International Nuclear Information System (INIS)

    Based on preliminary studies at Lawrence Livermore National Laboratory (LLNL), we believe the Mirror Fusion Test Facility (MFTF-B) could be upgraded for operation in a hot-ion Kelley mode in a portion of the central cell to provide fusion nuclear engineering data, particularly blanket technology information, by the end of the decade. Cost of this mode of operation would be modest compared with that of the other fusion devices considered in the last few years for such purposes

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

    International Nuclear Information System (INIS)

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li2O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N2) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li2O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concepts are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

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

    International Nuclear Information System (INIS)

    An 80-K Multi Layer Insulation (MLI) blanket in the interconnect region between magnets DD0019 and DD0027 in the Fermi National Accelerator Laboratory (FNAL) ER string was instrumented with temperature sensors to obtain the steady-state temperature gradient through the blanket after string cooldown. A thermal model of the 80-K blanket assembly was constructed to analyze the steady-state temperature gradient data. Estimates of the heat flux through the 80-K MLI blanket assembly and predicted temperature gradients were calculated. The thermal behavior of the heavy polyethylene terapthalate (PET) cover layers separating the shield and inner blanket and inner and outer blankets was derived empirically from the data. The results of the analysis predict a heat flux of 0.363--0.453 W/m2 based on the 11 sets of data. These flux values are 33--46% below the 80-K MLI blanket heat leak budget of 0.676 W/m2. The effective thermal resistance of the two heavy PET cover layers between the shield and inner blanket was found to be 2.1 times that of a single PET spacer layer, and the effective resistance of the two heavy PET cover layers between the inner blanket and outer blanket was found to be 7 times that of a single PET spacer layer. Based on these results, the 80-K MLI blanket assembly appears to be performing more than adequately to meet the 80-K static IR heat leak budget. However, these results should not be construed as a verification of the 80-K static IR heat leak, since no actual heat leak was measured. The results have been used to improve the empirically based model data in the 80-K MLI blanket thermal model, which has previously not included the effects of heavy PET cover layers on 80-K MLI blanket thermal performance

  17. Investigation of aqueous slurries as fusion reactor blankets

    International Nuclear Information System (INIS)

    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

  18. Test apparatus for ITER blanket pebble packing behavior

    International Nuclear Information System (INIS)

    Current Japanese design for ITER Driver Blanket consists of three breeder layers, nine multiplier layers and five cooling panels. The breeder layers and the multiplier layers contain 1 mm diameter spheres of Li2O and Be, respectively. The heat transfer in such 'Pebble Layered Blanket' is largely affected by the packing fraction of the pebbles which can be easily changed by the vibration during the operation. The packing fraction of the pebbles are expected to be as high as possible on the view point of nuclear heat design to maintain the optimum temperature of the breeder layer. Thus, it is necessary to establish the stable packed bed of the breeder and multiplier. The present experimental apparatus was fabricated for the engineering tests with the partial model of Japanese blanket. Test apparatus consists of stainless steel test panels, transparent plastic test panels, vibrators and measurement instruments. The apparatus can examine various parameters of sphere packed beds such as packing fraction, panels deformation, loading weight at the bottom of the panels and so on under various vibrating conditions. (author)

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

    International Nuclear Information System (INIS)

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

  20. Enhanced plasma current collection from weakly conducting solar array blankets

    Science.gov (United States)

    Hillard, G. Barry

    1993-05-01

    Among the solar cell technologies to be tested in space as part of the Solar Array Module Plasma Interactions Experiment (SAMPIE) will be the Advanced Photovoltaic Solar Array (APSA). Several prototype twelve cell coupons were built for NASA using different blanket materials and mounting techniques. The first conforms to the baseline design for APSA which calls for the cells to be mounted on a carbon loaded Kapton blanket to control charging in GEO. When deployed, this design has a flexible blanket supported around the edges. A second coupon was built with the cells mounted on Kapton-H, which was in turn cemented to a solid aluminum substrate. A final coupon was identical to the latter but used germanium coated Kapton to control atomic oxygen attack in LEO. Ground testing of these coupons in a plasma chamber showed considerable differences in plasma current collection. The Kapton-H coupon demonstrated current collection consistent with exposed interconnects and some degree of cell snapover. The other two coupons experienced anomalously large collection currents. This behavior is believed to be a consequence of enhanced plasma sheaths supported by the weakly conducting carbon and germanium used in these coupons. The results reported here are the first experimental evidence that the use of such materials can result in power losses to high voltage space power systems.

  1. Two dimensional distribution of tritium breeding ratio and induced activity in Japanese water cooled and helium cooled test blanket modules

    International Nuclear Information System (INIS)

    Solid breeder blankets are regarded as a near-at-hand blanket concept for a fusion power demonstration plant in Japan. Test blanket module (TBM) to be tested in ITER is the most important milestone to establish the fusion demonstration blanket. For the candidate TBM's, two types of TBM, water cooled solid breeder TBM, and a helium gas cooled solid breeder TBM have been proposed and designed in JAERI. For detailed performance study under operation and after shut down, detailed neutronics analysis gives the most important design conditions, such as, distribution of tritium breeding ratio, nuclear heating rate during operation, and induced activation and decay heat after termination of irradiation. In the analysis, neutron and gamma transportation was calculated by two dimensional analysis code, DOT3.5, for two TBMs. Nuclear reaction rate and induced activation rate were evaluated by APPLE-3 and ACT-4, respectively. The analysis model included configurations of thermo-mechanical test modules and surrounding common frames for both of He cooled and water cooled TBMs. By the neutronics analysis, TBR and contact dose rate by induced activation till one year after termination of the module testing have been evaluated. For the evaluation of induced activation level change and decay heat change, the transient decreases in one year after termination of the module testing have been calculated. The time duration of the module testing before termination of testing is assumed to be 133 continuous days of full power operation. The result of TBR analysis showed that TBR distribution in the toroidal direction of TBM is not significant, however, the neutron flux decreases in the region of sidewall of common frame made of SS and water. This result shows that there is relatively large neutron loss from the TBM to the common frame. Thus, it is considered that the TBR value observed in the TBM testing may be smaller than the estimation by one dimensional neutronics analysis which does

  2. Exchange reaction of hydrogen isotopes on proton conductor ceramic of hydrogen pump for blanket tritium recovery system

    International Nuclear Information System (INIS)

    Electrochemical hydrogen pump using ceramic proton conductor has been investigated to discuss its application for the blanket tritium recovery system of the nuclear fusion reactor. As the series of those work, the transportation experiments of H2-D2 mixture via ceramic proton conductor membrane have been carried out. Then, the phenomenon that was caused by the exchange reaction between the deuterium in the ceramic and the hydrogen in the gas phase has been observed. So, the ceramic proton conductor which doped deuterium was exposed to hydrogen under the control of zero current, and the effluent gas was analyzed. It is considered that the hydrogen in the gas phase is taken as proton to the ceramic by isotope exchange reaction, and penetrates to the ceramic by diffusion with replacement of deuteron. (author)

  3. Continuous fine pattern formation by screen-offset printing using a silicone blanket

    Science.gov (United States)

    Nomura, Ken-ichi; Kusaka, Yasuyuki; Ushijima, Hirobumi; Nagase, Kazuro; Ikedo, Hiroaki; Mitsui, Ryosuke; Takahashi, Seiya; Nakajima, Shin-ichiro; Iwata, Shiro

    2014-09-01

    Screen-offset printing combines screen-printing on a silicone blanket with transference of the print from the blanket to a substrate. The blanket absorbs organic solvents in the ink, and therefore, the ink does not disperse through the material. This prevents blurring and allows fine patterns with widths of a few tens of micrometres to be produced. However, continuous printing deteriorates the pattern’s shape, which may be a result of decay in the absorption abilities of the blanket. Thus, we have developed a new technique for refreshing the blanket by substituting high-boiling-point solvents present on the blanket surface with low-boiling-point solvents. We analyse the efficacy of this technique, and demonstrate continuous fine pattern formation for 100 screen-offset printing processes.

  4. Axial blanket fuel design and demonstration. First semi-annual progress report, January-September 1980

    International Nuclear Information System (INIS)

    The axial blanket fuel design in this program, which is retrofittable in operating pressurized water reactors, involves replacing the top and bottom of the enriched fuel column with low-enriched (less than or equal to 1.0 wt % 235U) fertile uranium. This repositioning of the fissile inventory in the fuel rod leads to decreased axial leakage and increased discharge burnups in the enriched fuel. Various axial blanket fuel designs, with blanket thicknesses from 0 to 10 inches and blanket enrichments from 0.2 to 1.0 wt % 235U, were investigated to determine the relationship between uranium utilization and power peaking. Analyses were preformed to assess the nuclear, mechanical, and thermal-hydraulic effects arising from the use of axial blankets. Four axial blanket lead test assemblies are being fabricated for scheduled irradiation in cycle 5 of Sacramento Municipal Utility District's Rancho Seco pressurized water reactor. Analyses to support licensing cycle 5 are in progress

  5. Studies of a self-cooled lead lithium blanket for HiPER reactor

    International Nuclear Information System (INIS)

    Within the frame of the HiPER reactor, we propose and study a Self Cooled Lead Lithium blanket with two different cooling arrangements of the system First Wall - Blanket for the HiPER reactor: Integrated First Wall Blanket and Separated First Wall Blanket. We compare the two arrangements in terms of power cycle efficiency, operation flexibility in out-off-normal situations and proper cooling and acceptable corrosion. The Separated First Wall Blanket arrangement is superior in all of them, and it is selected as the advantageous proposal for the HiPER reactor blanket. However, it still has to be improved from the standpoint of proper cooling and corrosion rates. (authors)

  6. MIT LMFBR blanket research project. Quarterly progress report, January 1, 1976--March 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, M.J.

    1976-01-01

    Progress in the experimental and theoretical investigation of LMFBR breeding blanket design parameters is reported. State-of-the-art approaches for the calculation of gamma heating in the core, blanket, and reflector regions of LMFBR's have been evaluated, with particular emphasis on coupled neutron-gamma methods cross section sets. The effects of heterogeneity on resonance self-shielding were examined for the blanket region and the capture reaction in /sup 238/U. (DG)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

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

    International Nuclear Information System (INIS)

    Tritium separation from inorganic lithium compounds: Li2O, LiAlO2, Li2SiO3, Li4SiO4, LiF, LiBeF3, Li2BeF4 irradiated with a beam of a gamma facility and a nuclear reactor, has been studied. In the first case the gas phase is absent. In the latter one- the tritium amount in the gas does not exceed 1-2% of its total amount in the salt. Based on the EPR spectra of irradiated salts the concentrations of paramagnetic centres are calculated. It is shown that during thermal annealing the main portion of tritium in the gas phase is in the form of oxide (HTO, T2O). Tritium is separated from lithium fluoroberyllates in the form of hydrogen (HT, T2). The kinetics of tritium oxide isolation from irradiated lithium oxide aluminate, metha- and orthosilicates, lithium sulphate has been studied. The activation energies of tritium oxide separation process are presented. A supposition is made that chemical reaction of the HTO (T2O) or HT(T2) or HF(TF) formation is a limiting stage. Clarification of the process stage limiting the rate of tritium recovery will permit to evaluate conditions for the optimum work of lithium material in the blanket, lithium zone to select the lithium element structure and temperature regime of irradiation

  10. Cold trapping of traces of tritiated water from the helium loops of a fusion breeder blanket

    International Nuclear Information System (INIS)

    The ITER Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) will comprise three helium loops designed for: tritium extraction from the breeder zone, heat removal, and purification of the coolant. The process step envisaged for tritium extraction as well as for coolant purification includes a cryogenic cold trap as main component for the removal of tritiated water vapour (mainly HTO, H2O). The concentrations of water in the gas streams are expected to be extremely small, i.e. of the order of 10 ppm by volume. In this paper, we describe first runs with a cold trap using helium as the carrier gas at flow rates of 0.1 and 1.0 m3/h. The range of water vapour concentration in the helium carrier gas was 0.5 to >200 ppmv. The experiments have demonstrated the ability of the cold trap to remove water vapour efficiently from the He stream down to concentrations of less than 0.02 ppmv when the inlet water concentration is in the range of 300-650 ppmv or higher

  11. Structural optimization of the blanket first wall to reduce thermal stress using the Taguchi method

    International Nuclear Information System (INIS)

    The first wall of a fusion reactor blanket faces the core plasma directly. The first wall endures high heat loads that lead to high thermal stresses. To ensure the reliability of the first wall structure, it is desirable to reduce the thermal stress. In this study, structural optimization of the blanket first wall was carried out using the Taguchi method. The finite element method was used to conduct a numerical simulation to investigate the thermo-mechanical responses of the blanket first wall. The optimal configuration of the blanket first wall was derived. (author)

  12. Survey on experimental neutron physics of CTR blankets in the KFA

    International Nuclear Information System (INIS)

    The research program of the KFA on neutronic CTR-blanket problems is described by examples of experiments on blanket-model assemblies. The work is in strong correlation with theoretical evaluations of the problems. The main objective is the experimental test of nuclear data and calculational methods by comparison of the results. The properties of the experimental blanket models used are derived from the existing CTR conceptual design studies and from the theory of neutron fields. The limitations of the present blanket experiments are discussed. The development of appropriate measuring techniques is also described

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Full text: JAEA has conducted the conceptual design study of blanket for a fusion DEMO reactor SlimCS. Considering DEMO specific requirements, we place emphasis on a blanket concept with durability to severe irradiation, ease of fabrication for mass production, operation temperature of blanket materials, and maintainability using remote handling equipment. This paper present a promising concept satisfying these requirements, which is characterized by minimized welding lines near the front, a simplified blanket interior consisting of cooling tubes and a mixed pebble bed of breeder and neutron multiplier, and approximately the same outlet temperature for all blanket modules. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in tritium breeding ratio even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production, which will facilitate the access of remote handling equipment for replacement of the blanket modules and improve the access of diagnostics. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  16. A European proposal for an ITER water-cooled solid breeder blanket

    International Nuclear Information System (INIS)

    The water-cooled solid breeder blanket concept proposed here aims to replace the shielding blanket for the enhanced performance phase of the international thermonuclear experimental reactor (ITER). The nominal performances are as follows: an average neutron wall load of 1 MW m-2 which corresponds to a fusion power of about 1.5 GW, and an average neutron fluence of 1 MWy m-2. The proposed blanket concept has been designed to accept a power increase of about 30% and power transients up to 3-5 GW for a short time. This blanket concept is based on a breeder inside tube (BIT)-type blanket with poloidal breeding elements made of 316 L-type stainless steel and filled with lithium metazirconate and beryllium pebbles. Inlet and outlet water temperatures of 160 and 200 C have been considered with a medium-pressure cooling system during plasma burn. The diameters of the breeding elements are compatible with the space available in test fission reactor core channels, making in-pile testing, required for blanket development and qualification, easier. A conservative approach using qualified materials, a blanket concept easily testable in fission reactors and on-going mock-up testing, which can be qualified using the blanket test module during the basic performance phase of ITER, will allow the blanket reliability required for the enhanced performance phase to be achieved. (orig.)

  17. European reference design of the water-cooled lithium-lead blanket for a demonstration reactor

    International Nuclear Information System (INIS)

    The water-cooled Pb-17Li blankets represent one of the blanket lines selected within the European Union for DEMO-relevant design and R ampersand D activities. This paper gives a presentation of the reference conceptual design for water-cooled Pb-17Li DEMO blankets and an overview on the results of its performance assessments. Moreover, a critical discussion about the technical aspects requiring further improvements and/or modifications is performed taking into account the present status of the associated R ampersand D. This concept appears to be a very promising candidate for a DEMO reactor breeding blanket

  18. Preliminary results of the MARIOS experiment on minor actinide bearing blanket concept

    International Nuclear Information System (INIS)

    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 241Am is, therefore, an option for the reduction of radiotoxicity, residual power packages as well as the repository area. One of the most attractive possibilities to burn MA is represented by the minor actinides bearing blanket (MABB) concept. In this option, MA is diluted in a UO2 matrix and irradiated for a long time (4 100 EFPD) into radial blankets at the periphery of the outer core of a sodium fast reactor. The MARIOS irradiation experiment is the latest of a series of experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS). Moreover, the MARIOS experiment, together with the DIAMINO experiment, which is in progress at the OSIRIS reactor in France, is dealing with the study of the in-pile behaviour of UO2 containing minor actinides (MA) in order to gain knowledge on the role of the microstructure and of the temperature on gas release and on fuel swelling for the MABB concept. The MARIOS experiment is carried out in the framework of the 4-year project FAIRFUELS of the EURATOM 7. Framework Programme (FP7). The MARIOS experiment was conducted at the HFR (high flux reactor) in Petten (The Netherlands) and started in March 2011. It has been planned that the experiment will last 11 cycles, corresponding approximately to 11 months. This paper covers the description and the objectives of the experiment, as well as the preliminary results in terms of first assessment and temperature recording. (authors)

  19. 18 CFR 157.210 - Mainline natural gas facilities.

    Science.gov (United States)

    2010-04-01

    ... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false Mainline natural gas... COMMISSION, DEPARTMENT OF ENERGY REGULATIONS UNDER NATURAL GAS ACT APPLICATIONS FOR CERTIFICATES OF PUBLIC... GAS ACT Interstate Pipeline Blanket Certificates and Authorization Under Section 7 of the Natural...

  20. 78 FR 21349 - Orders Granting Authority To Import and Export Natural Gas, To Export Liquefied Natural Gas, To...

    Science.gov (United States)

    2013-04-10

    ... Irving Oil Commercial Order granting blanket GP. authority to export natural gas to Canada. 3214 01/02/13... Granting Authority To Import and Export Natural Gas, To Export Liquefied Natural Gas, To Export Compressed.... INC 12-113-LNG MAIN PASS ENERGY HUB, LLC 12-114-LNG IRVING OIL COMMERCIAL GP 12-164-NG XPRESS...

  1. The impact of blanket design on activation and thermal safety

    International Nuclear Information System (INIS)

    Activation and thermal safety analyses for experimental and power reactors are presented. The effects of a strong neutron absorber, B4C, on activation and temperature response of experimental reactors to Loss-of-Cooling Accidents are investigated. Operational neutron fluxes, radioactivities of elements and thermal transients are calculated using the codes ONEDANT, REAC and THIOD, respectively. The inclusion of a small amount of B4C in the steel blanket of an experimental reactor reduces its activation and the post LOCA temperature escalation significantly. Neither the inclusion of excessive amounts of B4C nor enriched 10B in the first walls of an experimental reactor bring much advantage. The employment of a 2 cm graphite tile liner before the first wall helps to limit the post LOCA escalation of first wall temperature. The effect of replacing a 20 cm thick section of a steel shield of a fusion power reactor with B4C is also analyzed. The first wall temperature peak is reduced by 100 degree C in the modified blanket. The natural convection effect on thermal safety of a liquid lithium cooled blanket are investigated. Natural convection has no impact at all, unless the magnetic field can be reduced. If magnets can be shut off rapidly after the accident, then the temperature escalation of the first wall will be limited. Upflow of the coolant is better than the initial downflow design from a thermal safety point of view. Activities of three structural materials, OTR stainless steel, SS-316 and VCrTi are compared. Although VCrTi has higher activity for a period of two hours after the accident, it has one to two orders of magnitude less activity than those of the steels in the mid- and long-terms. 29 refs., 42 figs., 9 tabs

  2. Thermal cycle test of elemental mockups of ITER breeding blanket

    International Nuclear Information System (INIS)

    Thermal cycle tests for mockups of breeder pebble beds of ITER breeding blanket have been carried out to investigate their thermo-mechanical behavior with the interaction between a pebble bed and a breeder rod containing the breeder pebbles. The mockups have been designed to demonstrate a part of the Breeder Inside Tube (BIT)' structure of ITER breeding blanket. Candidate material pebbles of Li2TiO3 was applied as breeder specimen, and Al pebbles were applied for simulating the neutron multiplier of Be pebbles. These pebbles have been packed in test tubes by using a vibration machine. Tested configurations were single layer mockups with Li2TiO3 single diameter packing and binary packing beds, and double layer mockups with Li2TiO3/Al single diameter packing and binary packing beds. In order to clarify the deformation performance of breeder tube, two different thickness of the breeder rod were also tested: one for nominal condition and another for acceleration test. Pebble bed of Li2TiO3 is heated with an electric heater, which is equipped at the center of the breeder rod, simulating the temperature profile by volumetric heating of breeder pebbles. The outside of a breeder rod in a single layer mockup and the outside of the outer tube in case of double layer mockup is cooled by water. Temperature of the breeder beds has been controlled by a power input of the heater. After the thermal cycle tests, the internal dimensions and local packing fraction of mockups have been examined by using an X-ray CT device. As the result, no significant change of packing fraction was observed after five thermal cycles with maximum heater temperature of 600degC. Any bulging of the breeder rod or any cracking of the pebble has not been observed. A soundness of the typical structure and breeder pebble bed of ITER breeding blanket against thermal cycles was confirmed. (author)

  3. Impact of prescribed burning on blanket peat hydrology

    Science.gov (United States)

    Holden, Joseph; Palmer, Sheila M.; Johnston, Kerrylyn; Wearing, Catherine; Irvine, Brian; Brown, Lee E.

    2015-08-01

    Fire is known to impact soil properties and hydrological flow paths. However, the impact of prescribed vegetation burning on blanket peatland hydrology is poorly understood. We studied 10 blanket peat headwater catchments. Five were subject to prescribed burning, while five were unburnt controls. Within the burnt catchments, we studied plots where the last burn occurred ˜2 (B2), 4 (B4), 7 (B7), or greater than 10 years (B10+) prior to the start of measurements. These were compared with plots at similar topographic wetness index locations in the control catchments. Plots subject to prescribed vegetation burning had significantly deeper water tables (difference in means = 5.3 cm) and greater water table variability than unburnt plots. Water table depths were significantly different between burn age classes (B2 > B4 > B7 > B10+) while B10+ water tables were not significantly different to the unburnt controls. Overland flow was less common on burnt peat than on unburnt peat, recorded in 9% and 17% of all runoff trap visits, respectively. Storm lag times and hydrograph recession limb periods were significantly greater (by ˜1 and 13 h on average, respectively) in the burnt catchments overall, but for the largest 20% of storms sampled, there was no significant difference in storm lag times between burnt and unburnt catchments. For the largest 20% of storms, the hydrograph intensity of burnt catchments was significantly greater than those of unburnt catchments (means of 4.2 × 10-5 and 3.4 × 10-5 s-1, respectively), thereby indicating a nonlinear streamflow response to prescribed burning. Together, these results from plots to whole river catchments indicate that prescribed vegetation burning has important effects on blanket peatland hydrology at a range of spatial scales.

  4. The current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

    The available thermodynamic information is reviewed for three categories of materials that meet essential criteria for use as breeding blankets in D-T fuelled fusion reactors: liquid lithium, solid lithium alloys, and lithium-containing ceramics. The leading candidate, liquid lithium, which also has potential for use as a coolant, has been studied more extensively than have the solid alloys or ceramics. Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries, etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phases of the Li-LiH, Li-LiD and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li3N, Li2O, and Li2C2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g. Li-Al and Li-Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li-M alloys can be estimated from lithium activity data for these alloys. There is essentially no refined thermodynamic information on the prospective ceramic blanket materials. The kinetics of tritium release from these materials is briefly discussed. Research areas are pointed out where additional thermodynamic information is needed for all three material categories. (author)

  5. Reddening and blanketing of RR-Lyrae stars, ch. 3

    International Nuclear Information System (INIS)

    The effects of metal line blanketing and interstellar reddening upon the colours of the RR-Lyrae Stars are discussed. Due to the faintness of these stars in the ultraviolet W channel (at lambda 3720 A) the photometry is in most cases reduced to a four-colour VBLU photometry, i.e. there are only three colour indices available for the determination of the four quantities: interstellar reddening, effective temperature, atmospheric pressure (or effective gravity), and metal line strength which determine the energy distribution that was measured

  6. Recovery of tritium from a liquid lithium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Talbot, J.B.

    1981-01-01

    The sorption of tritium on yttrium from liquid lithium and the subsequent release of tritium from yttrium by thermal regeneration of the metal sorbent were investigated to study such a tritium-recovery process for a fusion reactor blanket of liquid lithium. Recent static sorption experiments have shown the effects of lithium temperature and possible impurities on the sorption of tritium. Diffusivity data, obtained from previous tritium recovery experiments, were evaluated to show the importance of the yttrium surface condition in controlling the release of tritium.

  7. Electronic measurement of tritium breeding in a fusion blanket assembly

    International Nuclear Information System (INIS)

    The tritium production rate (TPR) distribution in a fusion blanket assembly previously determined from measurements of tritium beta activity was remeasured using an independent electronic method. The results agreed within the experimental errors and confirmed the previously reported discrepancies with predictions based on three-dimensional Monte Carlo calculations and multigroup cross sections. The experimental agreement reduced the possibility that results based on the conventional chemical separation of the tritium produce could be subject to a common systematic error and confirmed the validity of the electronic method for TPR measurement

  8. Magnetohydrodynamic research in fusion blanket engineering and metallurgical processing

    International Nuclear Information System (INIS)

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

  9. Thermal analysis of the ITER blanket first wall

    International Nuclear Information System (INIS)

    The 3D temperature distribution in the First Wall (FW) of the International Thermonuclear Experimental Reactor (ITER) blanket is studied. The effect of FW exposure to different heat fluxes and heat generation rates on the temperature distribution inside the wall is also examined. The design of FW adopted by ITER council in 2001 is taken as a reference design for the FW through the analysis. The study reveals that the maximum and minimum temperatures increase linearly along the poloidal direction according to the specified incident heat flux and heat generation. The study also indicates a linear variation for the coolant temperature along the cooling channels throughout the poloidal direction

  10. Neutronic study on seed-blanket type reduced-moderation water reactor fuel assembly

    International Nuclear Information System (INIS)

    Parametric studies have been done for a PWR-type reduced-moderation water reactor (RMWR) with seed-blanket fuel assemblies to achieve a high conversion ratio, a negative void reactivity coefficient and a high burnup by using MOX, metal (Pu+U+Zr) or T-MOX (PuO2+ThO2) fuels. From the result of the assembly burnup calculation, it has been seen that 50% to 60% of seed in a seed-blanket (MOX-UO2) assembly has higher conversion ratio compared to the other combinations of seeds and blankets. And the recommended number of seed-blanket layers is 20, in which the number of seed layers is 15 (S15) and that of blanket layers is 5 (B5). It was found that the conversion ratio of a seed-blanket assembly decreases, when seed and blanket are arranged so as to look like a flower shape (Hanagara). By the optimization of different parameters, the S15B5 fuel assembly with the height of seed of 1,000x2 mm, internal blanket of 150 mm and axial blanket of 400x2 mm is recommended for a high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and that of blanket fuel rod is 0.4 mm. In the S15B5 assembly, the conversion ratio is 1.0 and the average burnup in (seed + internal blanket + outer blanket) region is 38 GWd/t. The cycle length of the core is 16.5 effective full power in month (EFPM) by 6 batches refuelling scheme and the enrichment of fissile Pu is 14.6 wt%. The void coefficient is +22 pcm/%void, though, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use the S15B5 fuel assembly as a high burnup reactor to achieve 45 GWd/t in (seed + internal blanket + outer blanket) region, but, it is necessary to decrease the height of seed to 500x2 mm to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +21 pcm/%void. The fuel temperature coefficient is negative for both of the cases. It is possible to improve the conversion

  11. R and D activities of the liquid breeder blanket in Korea

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer MARS and GAMMA were developed for He coolant and liquid breeder analysis. Black-Right-Pointing-Pointer FMS/FMS and Be/FMS joining methods were developed and verified with high heat flux test. Black-Right-Pointing-Pointer High temperature and pressure nitrogen and He loops were constructed for heat transfer experiment for developed codes validation. Black-Right-Pointing-Pointer A PbLi breeder loop was constructed for components, MHD, and corrosion tests. Black-Right-Pointing-Pointer A chamber for tritium extraction with a gas-liquid contact method was constructed. - Abstract: A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the common fields of a solid TBM such as design tools, structural material, fabrication methods, and He cooling technology to support this concept for the ITER. Also, other fields such as a liquid breeder technology and tritium extraction have been developed from the designed liquid TBM. For design tools, system codes for safety analysis such as Multi-dimensional Analysis of Reactor Safety (MARS) and GAs Multi-component Mixture Analysis (GAMMA) were developed for He coolant and liquid breeder. For the fabrication methods, Ferritic Martensitic Steel (FMS) to FMS and Be to FMS joinings with a Hot Isostatic Pressing (HIP) were developed and verified with a high heat flux test of up to 0.5-1.0 MW/m{sup 2}. Moreover, three mockups were successfully fabricated and a 10-channel prototype is being fabricated to make a rectangular channel FW. For the integrity of the joining, two high heat flux test facilities were constructed, and one using an electron beam has been constructed. With the 6 MPa nitrogen loop, a basic heat transfer experiment for code validation was performed. From the verification of the components such as preheater and

  12. On the relation between plasma and neutral gas profiles in a cold gas-blanket system

    International Nuclear Information System (INIS)

    A solution for the neutral density profile using the measured plasma density and temperature gradients is presented. The fluid model is used. It is found that the penetration length for neutrals is underestimated in the situation where the integrated profiles are used. The ionization rate need not be inferred in the present calculation, because the ionization of neutrals diffusing into the plasma is implicitly included in measured profiles. This calculation is advantageous in the low temperature range where the ionization rate is a strongly varying function of temperature. Finally the presented solution indicates that the temperature gradient plays the essential role in the determination of the neutral density profile. (Auth.)

  13. Investigation of neutronics for CH DEMO blanket with helium-cooled ceramics breeding concepts

    International Nuclear Information System (INIS)

    ITER TBM provides the strong support for the design, materials and technology of DEMO blanket. However, ITER TBM is quite different from a DEMO blanket in aspects of boundary conditions and neutron wall loading. It is very important to further clarify relations between ITER TBM and DEMO blanket. Neutronics of the blanket is theory basis for development of fusion reactor. In order to further identify the outline design for China ITER helium-cooled solid breeder (HCSB) test blanket module (TBM) in view of Chine DEMO goal, investigation of neutronics for a DEMO reactor with helium-cooled ceramics breeding blanket is investigated by means of three dimensional MCNP code. In this paper, the author attempts to explore the pathway from ITER TBM to DEMO blanket in view of neutronics design. (1) One-dimensional neutronics of three types of breeding blanket with 4 BZ (breeding zones) (Case 1), 2 BZ (Case 2) and 3 BZ (Case 3), respectively, are studied when neutron wall loading is assumed to be 0.78 MW/m2 on ITER and 2.64 MW/m2 on DEMO. Results show that TBR (tritium breeding ratio) of Case 1 is the smallest one that is adopted by CH ITER HCSB TBM. TBR of Case 3 is 1.43 and the largest one. Case 2 has the most simplified structure and the highest power density of 20.58 MW/m3 (ITER Wn=0.78 MW/m2), which is approach to 23.49 MW/m3 (DEMO: Wn=2.64 MW/m2), that is, TBM on ITER (Case 2) is probably used to test the characterization of DEMO blanket (Case 1). ITER TBM can approach DEMO blanket in a certain engineering parameters although ITER TBM cannot approach to DEMO blanket in overall engineering conditions. Three types of HCSB blankets are all useful according to different requirements of DEMO blanket. (2) 3D neutronics calculation is much necessary for defining tritium self-sufficiency of DEMO blanket. When Case1, Case 2, Case 3 is applied to CH HCSB DEMO blanket, TBR is 0.95, 1.04 and 1.11, respectively. In three cases, case 3 has the largest TBR more than 1.0 and is proposed

  14. Stability of LMR oxide pins and blanket rods during run-beyond-cladding-break (RBCB) operation

    International Nuclear Information System (INIS)

    Since 1981, the U.S. Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan have collaborated on an operational reliability testing program in the Experimental Breeder Reactor II. The tests were designed to determine the irradiation behavior of liquid-metal reactor (LMR) oxide pins and blanket rods during steady-state, transient, and run-beyond-claddin-breach (RBCB) operation. Phase I tests completed in 1987 involved current LMR oxide designs and claddings; the phase II tests begun in 1988 concentrate on advanced LMR designs, large-diameter pins (7.5 mm), and advance cladding alloys. The cladding breaches in these tests have been readily detected by fission-gas and delayed-neutron (DN) precursor release. The condition of the fuel pin has been monitored by these releases during RBCB operation. A variety of failures have been intentionally studied in the RBCB portion of the program for operating times of up to 142 full-power days; also, several failure types have been incidentally experienced during the transient tests. Types of failure have included those induced by gas-pressure loading either naturally or by prethinning of the cladding defects, and fuel-cladding mechanical interaction (FCMI)-induced failures or secondary failures caused by the formation of low-density fuel-sodium reaction product (FSRP). This paper summarizes this experience with regard to LMR oxide fuel stability during RBCB operation

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

    International Nuclear Information System (INIS)

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

  16. Annual report of the CTR blanket engineering research facility in 1993

    International Nuclear Information System (INIS)

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1993. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (author)

  17. Annual report of the CTR Blanket Engineering research facility in 1992

    International Nuclear Information System (INIS)

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1992. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (J.P.N.)

  18. Annual report of the CTR Blanket Engineering research facility in 1996

    International Nuclear Information System (INIS)

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1996. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (J.P.N.)

  19. Annual report of the CTR Blanket Engineering research facility in 1994

    International Nuclear Information System (INIS)

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor(CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1994. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (author)

  20. Activation analyses for the different options considered in the US ITER blanket trade-off study

    International Nuclear Information System (INIS)

    Detailed activation analyses were performed for the different blanket design options considered in the ITER blanket option trade-off study. The options considered included a self-cooled Li/V option, a helium-cooled Li/V option and a water-cooled 316 SS non-breeding shield option. A vacuum vessel made of double-wall Inconel 625 and water-cooled 316 SS balls was used with all options. The He-cooled blanket activity is higher than that of the self-cooled blanket due to the larger structure content. Meanwhile, the vacuum vessel activity is lower for the He-cooled blanket option due to the larger neutron attenuation in the blanket. The shield activity and decay heat of the 316 SS/H2O option are higher than those of the Li/V blankets due to the large amount (80%) of 316 SS used. In both Li/V options the blanket qualifies as class C low-level waste. On the other hand, the 316 SS/H2O shield does not qualify for disposal as low- level waste. The 316 SS/H2O option produces the highest off-site doses in the case of accidental release of 100% of its radioactive inventory. Only remote maintenance would be allowed for all options. (orig.)

  1. Blanket Cooling Concepts and Heat Conversion Cycles for Controlled Thermonuclear Reactors

    International Nuclear Information System (INIS)

    The neutron energy generated in fusion plasmas produces heat in solid and liquid blanket regions which confine the fusion plasma. This heat has to be removed by cooling the blanket, and is then converted into electric energy by thermodynamic processes. Blanket cooling can directly be achieved by cycling the liquid blanket material through an outside cooler, or indirectly by leading gaseous or liquid coolants through pipes in the blanket. For the conversion of heat into electricity, single Rankine or Brayton cycles can be applied as well as binary Rankine cycles using potassium and steam as cycle fluid. In this paper, the special features of different systems for blanket cooling and for heat conversion are described and discussed. The thermodynamic requirements for favourable operation of the different heat conversion cycles, and those for the heat removal system from the CTR ate pointed out. Also the pumping power in magnetically or unmagnetically coolant flows is considered. Selected solutions for combining the systems of blanket cooling and of heat conversion are compared with respect to plant efficiency, lost-volume fraction in the lithium blanket region, and the expected constructional and safety features. The comparisons are made for thermal powers of the reactor between 5000 and 10 000 MW and toroidal reactor configurations. Solutions to be preferred are pointed out. (author)

  2. Development and analysis of fusion breeder blanket neutronics. Progress report, November 1, 1983-October 31, 1984

    International Nuclear Information System (INIS)

    The following activities are briefly described: (a) the IBM versions of the computer codes FORSS, PUFF-II, ONETRAN, TWOTRAN-II, and DOT4.3 were obtained from the Radiation Shielding Information Center (RSIC) and have been implemented on the UCLA local computer, the IBM 3033; (b) mathematical and computational models to describe the time-dependent transport and inventory of tritium in individual components of a fusion reactor system have been developed; (c) extensive cross-section sensitivity and uncertainty analysis was carried out to evaluate an estimate for the uncertainty associated with the TBR (both from 6Li and 7Li, individually) in four of the leading blanket concepts (the Li2O/HT-9 helium-cooled blanket, the 17Li-83Pb/PCA self-cooled blanket, the LiAlO2/He/FS/Be blanket, and the flibe/He/FS/Be blanket); (d) as far as the TBR obtain able in various blanket concepts is concerned, a comparative analysis was carried out to estimate the change in TBR in a particular blanket module when placed in a tokamak machine [R (first wall) approx. 2 m] as opposed to adopting the same blanket in a mirror machine [R (first wall) approx. 50 cm] with the same wall loading

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

    International Nuclear Information System (INIS)

    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)

  4. Beryllium usage in fusion blankets and beryllium data needs

    International Nuclear Information System (INIS)

    Increasing numbers of designers are choosing beryllium for fusion reactor blankets because it, among all nonfissile materials, produces the highest number (2.5 neutron in an infinite media) of neutrons per 14-MeV incident neutron. In amounts of about 20 cm of equivalent solid density, it can be used to produce fissile material, to breed all the tritium consumed in ITER from outboard blankets only, and in designs to produce Co-60. The problem is that predictions of neutron multiplication in beryllium are off by some 10 to 20% and appear to be on the high side, which means that better multiplication measurements and numerical methods are needed. The n,2n reactions result in two helium atoms, which cause radiation damage in the form of hardening at low temperatures (300/degree/C). The usual way beryllium parts are made is by hot pressing the powder. A lower cost method is to cold press and then sinter. There is no radiation damage data on this form of beryllium. The issues of corrosion, safety relative to the release of the tritium built-up inside beryllium, and recycle of used beryllium are also discussed. 10 figs

  5. Fission power flattening in hybrid blankets using mixed fuel

    International Nuclear Information System (INIS)

    In a source-driven fissionable blanket, a flat fission power density (FPD) is achieved by using a mixed fuel (ThO2 and natural UO2) with the thoriumuranium ratio changing from front to back in the ten fuel rows along the radial direction. A straightforward graphic method is used. The temporal behavior of the FPD has been observed for an operation period of 6 months and for a plant load factor of 75% by applying a fusion driver neutron flux of 1014 14-MeV neutrons(cm2 . s) at the first wall, corresponding to --2.25 MWm2. To keep the power density flat, it is necessary to replace the fuel in rows 1, 2, and 3, close to the first wall. The time intervals for this operation increase, counting from initial start-up, typically, 2 months, 6 months, etc. One result of this study is that plutonium produced in such a hybrid blanket contains very low amounts of even isotopic components even over very long operation times of --3 yr. Hence, if fusion reactors are introduced into the energy market, special regulations are needed for international safeguarding

  6. First wall/blanket/shield design and optimization system

    International Nuclear Information System (INIS)

    First wall/blanket/shield design and optimization system (BSDOS) has been developed to provide a state-of-the-art design tool for fast accurate analysis. In addition, it has been designed to perform several other functions: (1) allowing comparison and evaluation studies for different concepts using the same data bases and ground rules, (2) permitting the use of any figure of merit in the evaluation studies, (3) optimizing the first wall/blanket/shield design parameters for any figure of merit under several design constraints, (4) permitting the use of different reactor parameters in the evaluation and optimization analyses, (5) allowing the use of improved eingineering data bases to study the impact on the design performance for planning future research and development, and (6) evaluating the effect of the data base uncertainties on the design performance. BSDOS is the first design and optimization system to couple the highly interacting neutronics, heat transfer, thermal hydraulics, stress analysis, radioactivity and decay-heat analyses, tritium balance, and capital cost. A brief description of the main features of BSDOS is given in this paper. Also, results from using BSDOS to perform design analysis for several reactor components are presented. 17 refs., 1 fig., 2 tabs

  7. Damage Identification of Continuum Structures Using Blanketing Effect

    International Nuclear Information System (INIS)

    A damage identification method for the continuum structures with making use of the blanketing effect is proposed in this paper. The conventional damage indicator methods such as flexibility matrix method are difficult to deal with the damage identification of complex continuum structure such as multi-span beam. The presented method takes the change rates according to certain diagonal elements of the damaged structure flexibility matrix as identification indicator function. The 'blanketing effect' of the indicator function with respect to the damage factors makes it easy to identify the damage of multi-span beam and plate structures. Some useful formulas are derived and the sensitivity matrix of the identification indicator function is studied. The numerical simulation is performed and the identification results of several damage status are compared. Both beam and plate structures are used for numerical simulation. The results show that the presented indicator function method is very brevity and effective. It is especially suitable for the damage identification of the multi-span beam and plate structures such as bridges and floor

  8. Fusion blanket materials development and recent R and D activities

    International Nuclear Information System (INIS)

    Development of structural materials plays an important role in the feasibility of fusion power plant. The candidate structural materials for future fusion reactors are Reduced Activation Ferritic Martensitic (RAFM) steel, nano structured ODS Steel, vanadium alloys and SiC/SiCf composite etc. RAFM steel is presently considered as the structural material for Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) because of its high void swelling resistance and improved thermal properties compared to austenitic steel. Development of RAFM steel in India is being carried out in full swing in collaboration with various research laboratories and steel industries. This paper presents an overview of the Indian activities on fusion blanket materials and describes in brief the efforts made to develop IN-RAFM steel as structural material for the LLCB TBM. In future, due to enhanced properties of vanadium base alloy and nano structured materials like ODS RAFMS, RAFM steel may be replaced by these materials for its application in DEMO relevant fusion reactor. Future R and D activities will be specifically towards the development of these structural materials for fusion reactor

  9. Ferritic steels for the first generation of breeder blankets

    International Nuclear Information System (INIS)

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

  10. Wash resistance and repellent properties of Africa University mosquito blankets against mosquitoes

    Directory of Open Access Journals (Sweden)

    N. Lukwa

    2013-04-01

    Full Text Available The effect of permethrin-treated Africa University (AU mosquito blankets on susceptible female Anopheles gambiae sensu lato mosquitoes was studied under laboratory conditions at Africa University Campus in Mutare, Zimbabwe. Wash resistance (ability to retain an effective dose that kills ≥80% of mosquitoes after a number of washes and repellence (ability to prevent ≥80% of mosquito bites properties were studied. The AU blankets were wash resistant when 100% mortality was recorded up to 20 washes, declining to 90% after 25 washes. Untreated AU blankets did not cause any mortality on mosquitoes. However, mosquito repellence was 96%, 94%, 97.9%, 87%, 85% and 80.7% for treated AU blankets washed 0, 5, 10, 15, 20 and 25 times, respectively. Mosquito repellence was consistently above 80% from 0-25 washes. In conclusion, AU blankets washed 25 times were effective in repelling and killing An. gambiae sl mosquitoes under laboratory conditions.

  11. Development of a virtual reality simulator for the ITER blanket remote handling system

    International Nuclear Information System (INIS)

    The authors developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robotic simulation software, ENVISION. The simulator is connected to the control system of the manipulator, which was developed as part of the blanket maintenance system during the Engineering Design Activity (EDA), and can reconstruct the positions of the manipulator and blanket module using position data transmitted from motors through a LAN. In addition, it can provide virtual visual information (e.g., about the interface structures behind the blanket module) by making the module transparent on the screen. It can also be used for confirming a maintenance sequence before the actual operation. The simulator will be modified further, with addition of other necessary functions, and will finally serve as a prototype of the actual simulator for the blanket remote handling system, which will be procured as part of an in-kind contribution

  12. Resonance self-shielding in the blanket of a hybrid reactor

    International Nuclear Information System (INIS)

    Three sets of energy group cross sections were obtained using various approximations for resonance self shielding. The three models used in obtaining the cross sections were: (a) infinitely dilute model, (b) homogeneous-medium resonance self shielding, and (c) heterogeneous-medium resonance self shielding. The effects on the blanket performance of fusion--fission hybrid reactors, and in particular, on the performance of the current reference Westinghouse Demonstration Tokamak Hybrid Reactor blanket, were compared and analyzed for a variety of fuel-coolant combinations. It has been concluded that (1) the infinitely dilute cross sections can be used to produce preliminary crude estimates for beginning-of-life (BOL) only, (2) the resonance absorber finite dilution should be considered for BOL, poorly moderated blankets and well moderated blankets with low fissile material content situations, and (3) the spacial details should be considered in high fissile content, well moderated blanket situations

  13. Blanket design and performance for the LOTUS fusion-fission hybrid test facility

    International Nuclear Information System (INIS)

    This report summarizes the results of studies performed during 1982 to design an optimized blanket for the initial series of experiments to be conducted in the LOTUS test facility at the Swiss Federal Institute of Technology in Lausanne (EPFL). The experiments are expected to begin in early 1984. An Overview of different hybrid blanket design concepts proposed to date is first given. The technological and economic implications of the different blanket design philosophies are discussed to provide the basis and rationale for the thorium fast-fission blanket design concept selected for the first series of experiments. Detailed description, dimensions, and characteristics of the selected blanket design are given. The neutronic optimization studies on which the design is based are described in detail. Instrumentation and measurement techniques to be used in LOTUS are described elsewhere

  14. Multi-dimensional neutronics analysis of the 'canister blanket' for NET

    International Nuclear Information System (INIS)

    At KfK a design of a helium-cooled ceramic breeder blanket, called 'canister blanket', has been developed for the NET fusion test reactor. In this report a detailed neutronic analysis of the 'canister blanket', based on one-, two- and three-dimensional Monte-Carlo calculations in the NET-III double null configuration, is presented. The main object refers to the three-dimensional analysis of a complete sector of the NET-reactor containing the 'canister blanket'. This concerns the poloidal distribution of the neutron wall load and the neutron fluxes at the first wall, the spatial distribution of the power density, the total power production and global effects on the tritium breeding ratio. It is shown that, in case of the 'canister blanket', a global tritium breeding ratio beyond 1.0 seems to be feasible for NET. (orig.)

  15. Preliminary thermo-mechanical analysis of ITER breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-01-01

    Thermo-mechanical analysis has been conducted on ITER breeding blanket taking into account thermo-mechanical characteristics peculiar to pebble beds. The features of the analysis are to adopt an elasto-plastic constitutive model for pebble beds and to take into account spatially varying thermal conductivity and heat transfer coefficient, especially in the Be pebble bed, depending on the stress. ABAQUS code and COUPLED TEMPERATURE-DISPLACEMENT procedure of the code are selected so that thermal conductivity is automatically calculated in each calculation point depending on the stress. The modified DRUCKER-PRAGER/Cap plasticity model for granular materials of the code is selected so as to deal with such mechanical features of pebble bed as shear failure flow and hydrostatic plastic compression, and capability of the model is studied. The thermal property-stress correlation used in the analysis is obtained based on the experimental results at FZK and the results of additional thermo-mechanical analysis performed here. The thermo-mechanical analysis of an ITER breeding blanket module has been performed for four conditions: case A; nominal case with spatial distribution of thermal conductivity and heat transfer coefficient in Be pebble bed depending on the stress, case B; constant thermal conductivity, case C; thermal conductivity = -20% of nominal case, and case D; thermal conductivity = +20% of nominal case. In the nominal case the temperature of breeding material (Li{sub 2}ZrO{sub 3}) ranges from 317degC to 554degC and the maximum temperature of Be pebble bed is 446degC. It is concluded that the temperature distribution is within the current design limits. Though the analyses performed here are preliminary, the results exhibit well the qualitative features of the pebble bed mechanical behaviors observed in experiments. For more detail quantitative estimates of the blanket performance, further investigation on mechanical properties of pebble beds by experiment

  16. Preliminary Analysis for K-DEMO Water Cooled Breeding Blanket Using MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    In the present study, thermal-hydraulic analyses for the blanket concept are being conducted using the Multidimensional Analysis of Reactor Safety (MARSKS) code, which has been used for the safety analysis of a pressurized water reactor. The purposes of the analyses are to verify the applicability of the code for the proposed blanket system, to investigate the departure of nucleate boiling (DNB) occurrence during the normal and transient conditions, and to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. In this paper, the thermal analysis results of the proposed blanket design using the MARS-KS code are presented for the normal operation and an accident condition of a reduced coolant flow rate. Afterwards, the plan for the whole blanket system analysis using MARSKS is introduced and the result of the first trial for the multiple blanket module analysis is summarized. In the present study, thermal-hydraulic analyses for the blanket concept were conducted using the MARS-KS code for a single blanket module. By comparing the MARS calculation results with the CFD analysis results, it was found that MARS-KS can be applied for the blanket thermal analysis with less number of computational meshes. Moreover, due to its capability on the two-phase flow analysis, it can be used for the transient or accident simulation where a phase change may be resulted in. In the future, the MARS-KS code will be applied for the anticipated transient and design based accident analyses. The investigation of the DNB occurrence during the normal and transient conditions will be of special interest of the analysis using it. After that, a methodology to simulate the entire blanket system was proposed by using the DLL version of MARS-KS. A supervisor program, which controls the multiple DLL files, was developed for the common header modelling. The program explicitly determines the flow rates of each module which can equalize

  17. Oxidation behavior of SIC/SIC composites for helium cooled solid breeder blanket

    International Nuclear Information System (INIS)

    SiC/SiC composite is one of the candidate structural materials for a fusion reactor blanket because of its low induced radioactivity, excellent high temperature mechanical properties and excellent radiation resistance. Helium (He) gas cooled blanket (HCSB) has been considered as one of the blanket design concepts using the SiC/SiC composite for relatively high temperature plant operation. Chemical stability, especially an oxidation resistance, is a key issue to be solved for the HCSB structural material because He gas in the HCSB might include partial oxygen. The desired strength of SiC/SiC composite can be given by an optimized interface layer between the fiber and matrix (F/M interface). In order to improve its mechanical properties, several advanced F/M interfaces such as an SiC/C multilayer (ML) and a porous SiC have been developed. However, SiC/SiC composites have a possibility of F/M interface degradation by oxidation at fusion reactor operating condition, for example by the reaction of C+O2→CO2. The purpose of this study is to evaluate the oxidation behavior of SiC/SiC composites with conventional pyrolitic carbon interface (PyC-SiC/SiC) and advanced multilayer interface (ML-SiC/SiC) in He+O2 environments at 1273 K. The SiC/SiC composites used in this work were fabricated at ORNL. Reinforced SiC fiber was 1D Hi-Nicalon Type-S fiber. SiC matrix was beta-SiC fabricated by a forced chemical vapor infiltration (FCVI) process. The average thickness of F/M interface (pyrolitic carbon and SiC/C multilayer) was 1000nm. Samples were machined into 2 mm x 1.5 mm x 2 mm blocks and the surface of them was mechanically polished. Oxidation tests were carried out using a thermal gravimetric analysis (TGA) equipment. Mixtures of He with 1500ppm O2 were used. Samples were heated from room temperature to the test temperature (1273 K) at 40 K/min. and then held at the test temperature for 100 h. Experimental conditions included 100 sccm flow rate and system pressure of 1 atm

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-02

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

  19. Tailorable advanced blanket insulation using aluminoborosilicate and alumina batting

    Science.gov (United States)

    Calamito, Dominic P.

    1989-01-01

    Two types of Tailorable Advanced Blanket Insulation (TABI) flat panels for Advanced Space Transportation Systems were produced. Both types consisted of integrally woven, 3-D fluted core having parallel faces and connecting ribs of Nicalon yarns. The triangular cross section flutes of one type was filled with mandrels of processed Ultrafiber (aluminoborosilicate) stitchbonded Nextel 440 fibrous felt, and the second type wall filled with Saffil alumina fibrous felt insulation. Weaving problems were minimal. Insertion of the fragile insulation mandrels into the fabric flutes was improved by using a special insertion tool. An attempt was made to weave fluted core fabrics from Nextel 440 yarns but was unsuccessful because of the yarn's fragility. A small sample was eventually produced by an unorthodox weaving process and then filled with Saffil insulation. The procedures for setting up and weaving the fabrics and preparing and inserting insulation mandrels are discussed. Characterizations of the panels produced are also presented.

  20. Heating performances of a IC in-blanket ring array

    Science.gov (United States)

    Bosia, G.; Ragona, R.

    2015-12-01

    An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is due to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing, at the highest power level, high RF standing and DC rectified voltages at the plasma periphery, with frequent voltage breakdowns and enhanced local wall loading. In a previous work [1] the concept for an Ion Cyclotron Heating & Current Drive array (and using a different wave guide technology, a Lower Hybrid array) based on the use of periodic ring structure, integrated in the reactor blanket first wall and operating at high input power and low power density, was introduced. Based on the above concept, the heating performance of such array operating on a commercial fusion reactor is estimated.

  1. Heating performances of a IC in-blanket ring array

    Energy Technology Data Exchange (ETDEWEB)

    Bosia, G., E-mail: gbosia@to.infn.it [Department of Physics, University of Turin (Italy); Ragona, R. [Laboratory for Plasma Physics-LPP-ERM/KMS, Brussels (Belgium)

    2015-12-10

    An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is due to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing, at the highest power level, high RF standing and DC rectified voltages at the plasma periphery, with frequent voltage breakdowns and enhanced local wall loading. In a previous work [1] the concept for an Ion Cyclotron Heating & Current Drive array (and using a different wave guide technology, a Lower Hybrid array) based on the use of periodic ring structure, integrated in the reactor blanket first wall and operating at high input power and low power density, was introduced. Based on the above concept, the heating performance of such array operating on a commercial fusion reactor is estimated.

  2. Experimental Investigation of Ternary Alloys for Fusion Breeding Blankets

    International Nuclear Information System (INIS)

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

  3. Ward blanket assembly pre-test predictions by superenergy

    International Nuclear Information System (INIS)

    A number of WARD sodium cooled blanket geometry experiments in the mixed convection range were prepredicted by the computer code SUPERENERGY in the original version of this report. Of these tests one was actually run and is re-examined in Revision I. When account is taken of the difference between the assumed and actual test operating conditions and the temperature-induced variation of the Reynolds number with height in the bundle, the code results and test data compare very favorably. Of the other actual test cases examined in Revision I, the comparison with predictions was poor at low flows. Since these low flows however were predicted to be outside the mixed convection regime per Khan's criteria, the validity of the criteria is suspect and will be reassessed

  4. Movement of tritiated water injected into blanket peat

    Directory of Open Access Journals (Sweden)

    R.S. Clymo

    2016-04-01

    Full Text Available In 1966, tritiated water was injected at five sites at depths between 25 and 100 cm into blanket bog at Moor House National Nature Reserve. The distribution of tritium activity on a logarithmically spaced grid around these sites was sampled in 1990, 24 years after placement. The proportions of tritium accounted for ranged from 80 % for the injection at 100 cm deep, to 20 % for the injection at 25 cm deep. Both 80 and 20 should be considered as ± 10 %. Results imply that diffusion close to the injection may have played a part in movement of tritium; evapotranspiration is not inconsistent with the losses inversely proportional to depth of placement; but the main process of movement is probably bulk (mass flow of water through the peat.

  5. Experimental Investigation of Ternary Alloys for Fusion Breeding Blankets

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-26

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

  6. Cosmetic wastewater treatment by upflow anaerobic sludge blanket reactor

    International Nuclear Information System (INIS)

    Anaerobic treatment of pre-settled cosmetic wastewater in batch and continuous experiments has been investigated. Biodegradability tests showed high COD and solid removal efficiencies (about 70%), being the hydrolysis of solids the limiting step of the process. Continuous treatment was carried out in an upflow anaerobic sludge blanket reactor. High COD and TSS removal efficiencies (up to 95% and 85%, respectively) were achieved over a wide range of organic load rate (from 1.8 to 9.2 g TCOD L-1 day-1). Methanogenesis inhibition was observed in batch assays, which can be predicted by means of a Haldane-based inhibition model. Both COD and solid removal were modelled by Monod and pseudo-first order models, respectively.

  7. Heating performances of a IC in-blanket ring array

    International Nuclear Information System (INIS)

    An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is due to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing, at the highest power level, high RF standing and DC rectified voltages at the plasma periphery, with frequent voltage breakdowns and enhanced local wall loading. In a previous work [1] the concept for an Ion Cyclotron Heating & Current Drive array (and using a different wave guide technology, a Lower Hybrid array) based on the use of periodic ring structure, integrated in the reactor blanket first wall and operating at high input power and low power density, was introduced. Based on the above concept, the heating performance of such array operating on a commercial fusion reactor is estimated

  8. Upflow anaerobic sludge blanket reactor--a review.

    Science.gov (United States)

    Bal, A S; Dhagat, N N

    2001-04-01

    inorganic matter in the absence of molecular oxygen. Complex polymeric materials such as polysaccharides, proteins, and lipids (fat and grease) are first hydrolyzed to soluble products by extracellular enzymes, secreted by microorganisms, so as to facilitate their transport or diffusion across the cell membrane. These relatively simple, soluble compounds are fermented or anaerobically oxidized, further to short-chain fatty acids, alcohols, carbon dioxide, hydrogen, and ammonia. The short-chain fatty acids (other than acetate) are converted to acetate, hydrogen gas, and carbon dioxide. Methanogenesis finally occurs from the reduction of carbon dioxide and acetate by hydrogen. The initial stage of anaerobic degradation, i.e. acid fermentation is essentially a constant BOD stage because the organic molecules are only rearranged. The first stage does not stabilize the organics in the waste. However this step is essential for the initiation of second stage methane fermentation as it converts the organic material to a form, usable by the methane producing bacteria. The second reaction is initiated when anaerobic methane forming bacteria act upon the short chain organic acids produced in the 1st stage. Here these acids undergo methane fermentation with carbon dioxide acting as hydrogen acceptor and getting reduced to methane. The methane formed, being insoluble in water, escapes from the system and can be tapped and used as an energy source. The production and subsequent escape of methane causes the stabilization of the organic material. The methane-producing bacteria consist of several different groups. Each group has the ability to ferment only specific compounds. Therefore, the bacterial consortia in a methane producing system should include a number of different groups. When the rate of bacterial growth is considered, then the retention time of the solids becomes important parameter. The acid fermentation stage is faster as compared to the methane fermentation stage. This

  9. Lithium ceramic of blankets intend for Russian fusion reactors and an influence of the ceramic properties on parameters of reactor tritium systems

    International Nuclear Information System (INIS)

    Russian Controlled Fusion Program involves development of a DEMO design and participation in ITER Project. A solid breeder blanket in DEMO contains a ceramic orthosilicate lithium breeder and a beryllium multiplier. Test Modules of the blanket are developed in a frame of ITER activities. Experimental models of tritium breeding zones (TBZ) for the Modules, materials and technology fabrication of the TBZ, tritium reactor systems to control and treat of gases released from lithium ceramic being developed. Two models of tritium breeding and neutron multiplying elements of the TBZ were designed, manufactured and have been tested already in IVV-2M nuclear reactor. The first model consists of lithium orthosilicate ceramic sphere pebbles (1-1.5 mm diameter) and beryllium sphere (0.1 and 1.0 mm diameter). Ceramic cylindrical pellets (11 mm diameter and 10 mm height) and porous beryllium (20% porosity) are in the second model. Some properties and microstructure of the ceramic elements are performed. Initial results of some changes of ceramic structure and in-pile experimental ratio of hydrogen and oxygen form of tritium release in helium/neon purge gas are presented. These results and outcome of irradiated LiAlO2, Li4SiO4 and Li2SiO3 ceramics in a water-graphite nuclear reactor are considered to be a DATE BASE for development of the Test Modules and the DEMO blanket and influence of the kinetic tritium release parameters on DEMO tritium systems are discussed. (author)

  10. Neutronics experiments for DEMO blanket at JAERI/FNS

    Science.gov (United States)

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

    2003-07-01

    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, Li2TiO3 blocks with a 6Li enrichment of 40% and 95%, and beryllium blocks. Sample pellets of Li2TiO3 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 needed to calculate the formation of the radioactive nuclei (56Co, 184Re, 48V, 206Bi, 65Zn and 51Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections increased remarkably while coming closer to polyethylene board, which was a substitute for water. As a result of this present study, it has become clear that the sequential reaction rates are important factors in the accurate evaluation of induced activity in fusion reactor design.

  11. Possible association of mucous blanket integrity with postirradiation colonization resistance

    International Nuclear Information System (INIS)

    Radiation-induced infections can be associated with changes in colonization potential of the intestine. Since the mucous blanket, which overlays the epithelium, is a major mucosal structure and is heavily colonized by microorganisms, we examined the status of the mucus after radiation and evaluated susceptibility to intestinal challenge with bacteria. A downward shift (2.5 X 10(8) cells/g to 5.3 X 10(5)) of total facultatively anaerobic bacteria of the ileum of C3HeB/FeJ mice was detected by 3 days post exposure to 10 Gy 60Co. Numbers of flora returned to normal by 11 days after radiation. Scanning electron microscopy was used to show that the loss of bacteria could be associated with major disruptions of the continuity of the mucous blanket. The pathogen Pseudomonas aeruginosa adhered to mouse mucous films used in in vitro assays. When irradiated mice were challenged orally with 1 X 10(5) P. aeruginosa on days 1, 2, or 3 after irradiation, a progressive increase in susceptibility was seen, but no animals died before Day 4 postirradiation. Sensitivity to subcutaneous (sc) challenge with Pseudomonas also increased by Day 3 and was probably due largely to the profound neutropenia observed. Immunoglobulin G (Gamimmune), which protected burned mice infected with Pseudomonas, was ineffectual in treatment of 7 or 10 Gy irradiated mice challenged either orally or sc with the organism. The ileal mucosal barrier was compromised after radiation in ways which could facilitate epithelial colonization, an event which combined with other immunological and physiological decrements in this model can compromise the effectiveness of therapeutic modalities

  12. Neutronic studies of fissile and fusile breeding blankets

    International Nuclear Information System (INIS)

    In light of the need of convincing motivation substantiating expensive and inherently applied research (nuclear energy), first a simple comparative study of fissile breeding economics of fusion fission hybrids, spallators and also fast breeder reactors has been carried out. As a result, the necessity of maximization of fissile production (in the first two ones, in fast breeders rather the reprocessing costs should be reduced) has been shown, thus indicating the design strategy (high support ratio) for these systems. In spite of the uncertainty of present projections onto further future and discrepancies in available data even quite conservative assumptions indicate that hybrids and perhaps even earlier - spallators can become economic at realistic uranium price increase and successfully compete against fast breeders. Then on the basis of the concept of the neutron flux shaping aimed at the correlation of the selected cross-sections with the neutron flux, the indications for the maximization of respective reaction rates has been formulated. In turn, these considerations serve as the starting point for the guidelines of breeding blanket nuclear design, which are as follows: 1) The source neutrons must face the multiplying layer (of proper thickness) of possibly low concentration of nuclides attenuating the neutron multiplication (i.e. structure materials, nongaseous coolants). 2) For the most effective trapping of neutrons within the breeding zone (leakage and void streaming reduction) it must contain an efficient moderator (not valid for fissile breeding blankets). 3) All regions of significant slow flux should contain 6Li in order to reduce parasite neutron captures in there. (orig./HP)

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

    International Nuclear Information System (INIS)

    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 Li2O test region, the rotating neutron target and the Li2CO3 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 Li2O 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)

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

    International Nuclear Information System (INIS)

    A trade-off study of liquid-metal self-cooled blankets was carried out to define the performance of these blankets with respect to the main functions in a fusion reactor, and to determine the potential to operate at the maximum possible values of the performance parameters. The main purpose is to improve the reactor economics by maximizing the blanket energy multiplication factor, reduce the capital cost of the reactor, and satisfy the design requirements. The main parameters during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the 6Li enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, the impact of different reactor design choices on the performance parameters was analyzed. The effect of the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, the coolant choice for the nonbreeding inboard blanket, and the neutron source distribution were part of the trade-off study. In addition, tritium breeding benchmark calculations were performed to study the impact of the use of different transport codes and nuclear data libraries. The importance and the negative effect of high TBR on the energy multiplication motivated the benchmark calculations

  15. Eddy current induced electromagnetic loads on shield blankets during plasma disruptions in ITER: A benchmark exercise

    International Nuclear Information System (INIS)

    According to recent updates of ITER shield blanket design, electromagnetic loads during the plasma disruption are being evaluated to verify the mechanical confidence and reliability. As a course of such evaluations, a benchmark activity for the electromagnetic analysis, coordinated by ITER Organization, is underway between ITER parties to compare the calculation results for disruption loads on the blankets. In this paper, we present calculation results for the electromagnetic loads on the simplified but practical model of ITER shield blankets with respect to six representative disruption scenarios of which ITER distributes simulation results based on the DINA code as a reference of the design and analysis. Commercial finite element method software, ANSYS/EmagTM, was employed to evaluate the eddy current on the blanket modules with the 40o sector model for major conducting structure of the tokamak including double-walled vacuum vessel, triangular support, and vertical targets of divertors. An interface between ANSYS/EmagTM and plasma simulator was implemented with a conversion tool assigning the plasma current density on the ANSYS elements corresponding to the current filaments in DINA outputs. Discussions are made of the possible improvement of the blanket model taking more realistic blanket configuration into account at the cost of the moderate increase in computational time. A final remark is given of the possibility of incorporating halo currents into ANSYS disruption simulations, which are major sources of electromagnetic loads on in-vessel components including blankets.

  16. Range of blanket concepts from near term solutions to advanced concepts

    International Nuclear Information System (INIS)

    Breeding blankets are key components of fusion power plants and determine to a large degree their attractiveness. There is a large range of possible blanket concepts, characterized by breeding material, structural material, coolant, and geometrical arrangement. On one extreme side there are concepts requiring only a modest extrapolation of the present technology, but with limited attractiveness. On the other side there are concepts with the potential for very attractive plants, but involving a considerable development risk. Therefore, the selection of blanket concepts depends on the overall strategy for fusion power plant development. To ensure sufficient support over the long development time, it has to be shown in a credible way, that (a) there are feasible blanket concepts for fusion power plants which can be developed with a high confidence of success, and (b) the final product will be an attractive power plant which can be operated safely, with minimum impact on the environment, and at an acceptable cost of electricity. This requires an evaluation of the entire range of blanket concepts, starting from near term solutions and reaching up to very advanced designs with a potential for exceptionally high performance. Typical examples of candidate blanket concepts are compared in this study. The blanket development strategy developed in the EU is described as an example

  17. Synthesis and Characterization of Fibre Reinforced Silica Aerogel Blankets for Thermal Protection

    Directory of Open Access Journals (Sweden)

    S. Chakraborty

    2016-01-01

    Full Text Available Using tetraethoxysilane (TEOS as the source of silica, fibre reinforced silica aerogels were synthesized via fast ambient pressure drying using methanol (MeOH, trimethylchlorosilane (TMCS, ammonium fluoride (NH4F, and hexane. The molar ratio of TEOS/MeOH/(COOH2/NH4F was kept constant at 1 : 38 : 3.73 × 10−5 : 0.023 and the gel was allowed to form inside the highly porous meta-aramid fibrous batting. The wet gel surface was chemically modified (silylation process using various concentrations of TMCS in hexane in the range of 1 to 20% by volume. The fibre reinforced silica aerogel blanket was obtained subsequently through atmospheric pressure drying. The aerogel blanket samples were characterized by density, thermal conductivity, hydrophobicity (contact angle, and Scanning Electron Microscopy. The radiant heat resistance of the aerogel blankets was examined and compared with nonaerogel blankets. It has been observed that, compared to the ordinary nonaerogel blankets, the aerogel blankets showed a 58% increase in the estimated burn injury time and thus ensure a much better protection from heat and fire hazards. The effect of varying the concentration of TMCS on the estimated protection time has been examined. The improved thermal stability and the superior thermal insulation of the flexible aerogel blankets lead to applications being used for occupations that involve exposure to hazards of thermal radiation.

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

    International Nuclear Information System (INIS)

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m2. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ''beryllium-joint'' concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket

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

    International Nuclear Information System (INIS)

    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

  20. Development of the breeding blanket and shield model for the fusion power reactors system SYCOMORE

    International Nuclear Information System (INIS)

    SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory. Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper. For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available. Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE. In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006–2007 studies) is also presented

  1. Development of the breeding blanket and shield model for the fusion power reactors system SYCOMORE

    Energy Technology Data Exchange (ETDEWEB)

    Li-Puma, Antonella, E-mail: antonella.lipuma@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Jaboulay, Jean-Charles, E-mail: Jean-Charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Martin, Brunella, E-mail: brunella.martin@gmail.com [Incka, 19-21 Rue du 8 mai 1945, F-94110 Arcueil (France)

    2014-10-15

    SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory. Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper. For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available. Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE. In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006–2007 studies) is also presented.

  2. Phase III experiments of the JAERI/USDOE collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    A pseudo-line source has been realized by using an accelerator based D-T point neutron source. The pseudo-line source is obtained by time averaging of continuously moving point source or by superposition of finely distributed point sources. The line source is utilized for fusion blanket neutronics experiments with an annular geometry so as to simulate a part of a tokamak reactor. The source neutron characteristics were measured for two operational modes for the line source, continuous and step-wide modes, with the activation foil and the NE213 detectors, respectively. In order to give a source condition for a successive calculational analysis on the annular blanket experiment, the neutron source characteristics was calculated by a Monte Carlo code. The reliability of the Monte Carlo calculation was confirmed by comparison with the measured source characteristics. The shape of the annular blanket system was a rectangular with an inner cavity. The annular blanket was consist of 15 mm-thick first wall (SS304) and 406 mm-thick breeder zone with Li2O at inside and Li2CO3 at outside. The line source was produced at the center of the inner cavity by moving the annular blanket system in the span of 2 m. Three annular blanket configurations were examined; the reference blanket, the blanket covered with 25 mm thick graphite armor and the armor-blanket with a large opening. The neutronics parameters of tritium production rate, neutron spectrum and activation reaction rate were measured with specially developed techniques such as multi-detector data acquisition system, spectrum weighting function method and ramp controlled high voltage system. The present experiment provides unique data for a higher step of benchmark to test a reliability of neutronics design calculation for a realistic tokamak reactor. (J.P.N.)

  3. Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

    International Nuclear Information System (INIS)

    A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

  4. Utilization of fusion neutrons very high energy effects in the design of a fusion reactor tritium breeding blanket

    International Nuclear Information System (INIS)

    Tritium needed for ITER fusion reactions can be regenerated in a blanket located in the tokamak. Such breeding has to be achieved through special interactions between very high energy (14.1 MeV) fusion neutrons and the blanket materials, like the lithium-7 tritium breeding effect and the thorium-232 neutron multiplication effect, all this while occupying the smallest possible space. Five blankets were designed and investigated; three of them were purely composed of lithium materials, while the two others were designed by adding a thorium layer before the lithium layer. A 3-D modeling was created using a Monte Carlo N-particle Code (MCNP) to simulate the fusion neutrons histories through the tritium breeding blankets, with a blanket thickness ranging from 1 cm to 200 cm. The minimum blanket thickness necessary to obtain a tritium breeding ratio (TBR) greater than one ranges from 20 cm to 45 cm. In particular, the Lithium Oxide Mono-Layer Blanket (LO-MLB) achieves a TBR greater than one while allowing blanket thickness to stay under 25 cm, thus making it the most efficient blanket in this sample. Second, the maximum TBR for thick blankets ranges from 1.5 to 2. In particular, the Natural Lithium Mono-Layer Blanket (NL-MLB) displays the highest maximum TBR thanks to its perfect combination of the lithium-7 and lithium-6 tritium breeding capacity. (author)

  5. A neutronic study on advanced sodium cooled fast reactor cores with thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    Highlights: • SFR burner core configurations are explored and analyzed for effective use of thorium blankets. • Thorium blankets can significantly improve SFR burner core performances. • No recycling or partial recycling of Th blankets with multi-batches is very effective. - Abstract: In this paper, new design concepts of sodium cooled fast reactor (SFR) cores having thorium blanket are suggested for pursuing effective burning of TRU (transuranics) nuclides from LWR spent fuels and their neutronic performances are analyzed. Several core configurations having different arrangements of thorium blankets are explored to improve the core performances and safety-related parameters including sodium void worth which is one of main concerns on safety of SFR cores. Specifically, axial and radial thorium blankets are considered for two type cores. The first one is the typical annular type cores having two different fuel regions where axial thorium blankets are placed in the axially central regions while the second one is the single fuel region cores having central non-fuel region where the axial blanket and radial blankets are considered. Also, the effects of the recycling options and fuel management schemes of the used thorium blanket on the core performances are analyzed. The core performance analyses show that thorium blankets with no recycling option and multi-batch fuel management schemes are very effective to improve the core performances including burnup reactivity swing, sodium void worth and TRU consumption rate

  6. 聚变堆液态包层提氚鼓泡器的概念设计%Conceptual design of tritium bubbler for fusion reactor liquid blanket

    Institute of Scientific and Technical Information of China (English)

    谢波; 翁葵平; 侯建平; 古梅

    2015-01-01

    The conceptual design of liquid blanket tritium bubbler (LBTB) with the gas-liquid exchange column as core was proposed, based on the works of hydrogen extraction from liquid lithium alloys by gas-liquid contact method. LBTB consists of the gas sample purifier, gas-liquid exchange column system, saturator-desorption and auxiliary system. The LBTB was Ar-H2 as carrier, and would on line monitor the tritium behavior of liquid blanket main loop, and test the tritium recovery efficiency whether or not reaching 90%after multi-column cascade.%在气-液接触法提取液态锂合金中的氢的实验基础上,提出了以气-液交换柱为核心的提氚鼓泡器(LBTB)的概念设计。LBTB 主要由气体进样纯化器、气-液交换柱系统、饱和器-解吸器和辅助系统构成。LBTB以氩氢混合气为吹洗气,其主要功能是在线监测液态包层主回路中的氚行为,并检验多柱级联后的氚回收率是否可以达到90%的期望值。

  7. Realization of a flat fission power density in a hybrid blanket over long operation periods

    International Nuclear Information System (INIS)

    A straightforward numerical graphical method is applied to achieve a flat fission power density (FPD) in a hybrid blanket by using a mixed fuel (ThO2 and natural UO2) with variable fractions of the fuel components in the radial direction. The neutronic analysis is carried out on a blanket with a hard neutron spectrum in the fissionable zone by simply omitting the moderating beryllium neutron multiplier. Mainly due to this precaution in the blanket design, the FPD could be kept quasi-constant over a relatively long plant lifetime

  8. APT Blanket Thermal Analysis of Cavity Flood Cooling with a Beam Window Break; FINAL

    International Nuclear Information System (INIS)

    The cavity flood system is designed to be the primary safeguard for the integrity of the blanket modules and target assemblies during loss of coolant accidents, LOCA''s. In the unlikely event that the internal flow passages in a blanket module or a target assembly dryout, decay heat in the metal structures will be dissipated to the cavity flood system through the module or assembly walls. This study supplements the two previous studies by demonstrating that the cavity flood system can adequately cool the blanket modules when the cavity vessel beam window breaks

  9. Studies on increasing tritium breeding ratios of JXFR and INTOR-J blankets

    International Nuclear Information System (INIS)

    Increasing the tritium breeding ratios of JXFR (JAERI Experimental Fusion Reactor) and INTOR-J (Japanese proposal for INTOR) blankets by loading PbO or Pb38Li62 as the neutron multiplyer and condensing 6Li has been studied. The breeding ratio of JXFR blanket increases from 1.09 to 1.17 in using PbO and condensing 6Li. In INTOR-J blanket the ratio increases from 1.10 to 1.20 in replacing Li2O by Pb38Li62 and condensing 6Li. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, B.J.

    1980-01-01

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

  12. Preliminary Analysis of Liquid Metal MHD Pressure Drop in the Blanket for the FDS

    Institute of Scientific and Technical Information of China (English)

    王红艳; 吴宜灿; 何晓雄

    2002-01-01

    Preliminary analysis and calculation of liquid metal Li17Pb83 magnetohydrodynamic (MHD) pressure drop in the blanket for the FDS have been presented to evaluate the significance of MHD effects on the thermal-hydraulic design of the blanket. To decrease the liquid metal MHD pressure drop, Al2O3 is applied as an electronically insulated coating onto the inner surface of the ducts. The requirement for the insulated coating to reduce the additional leakage pressure drop caused by coating imperfections has been analyzed. Finally, the total liquid metal MHD pressure drop and magnetic pump power in the FDS blanket have been given.

  13. STARTUP OF UPELOW ANAEROBIC SLUDGE BLANKET REACTOR FOR INDUSTRIAL WASTEWATER TREATMENT

    Directory of Open Access Journals (Sweden)

    A.R. Mesdaghinia

    1994-06-01

    Full Text Available Up flow anaerobic sludge blanket (UASB reactors have been increasingly used for industrial wastewater treatment. Because of existing problems in startup step of these reactors, in this research the startup of a UASB in pilot scale and room temperature condition was studied. The total height of UASB reactor was 270 cm and effective height was 240 cm. Diameter of the reactor in lower part was 20 cm (reaction zone and 40 cm in upper part (solid-gas-liquid separator five sampling ports with interval of 32 cm were provided and the effective volume of the reactor was 100 liters. Septic tank digested sludge and cow manure were used for the seeding of UASB reactor. In the startup step of the reactor, volumetric loading was increased step by step. After 155 days granule formation was observed and after 215 days of the study the removal rate increased to 4.62 kg COD/m/ day. More than 98% of soluble COD removal was achieved in lower 160 cm of reactor.

  14. Thermophilic degradation of phenolic compounds in lab scale hybrid up flow anaerobic sludge blanket reactors

    International Nuclear Information System (INIS)

    This Study describes the feasibility of anaerobic degradation of United States Environmental Protection Agency (USEPA) listed 4-chloro-2-nitrophenol (4C-2-NP), 2-chloro-4-nitrophenol (2C-4-NP), 2-chloro-5-methylphenol (2C-5-MP) from a simulated wastewater using four identical 7L bench scale hybrid up flow anaerobic sludge blankets (HUASBs) at five different hydraulic retention times (HRTs) under thermophilic condition (55 ± 3 deg. C). The substrate to co-substrate ratios were maintained between 1:33.3 and 1:166.6. Continuous monitoring of parameters like pH, volatile fatty acids (VFAs) accumulation, oxidation reduction potential, chemical oxygen demand (COD), alkalinity, gas productions, methane percentages were carried out along with compound reduction to asses the efficiency of biodegradation. The compound reduction was estimated by using spectrophotometric methods and further validated with high-performance liquid chromatography (HPLC). Optimum HRT values were observed at 24 h. Optimum ratio of substrate (phenolic compounds) to co-substrate (glucose) was 1:100. Scanning electron micrographs show that the granules were composed of thermophilic Methanobrevibacter and thermophilic Methanothrix like bacteria.

  15. R and D activities of the liquid breeder blanket in Korea

    International Nuclear Information System (INIS)

    Highlights: ► MARS and GAMMA were developed for He coolant and liquid breeder analysis. ► FMS/FMS and Be/FMS joining methods were developed and verified with high heat flux test. ► High temperature and pressure nitrogen and He loops were constructed for heat transfer experiment for developed codes validation. ► A PbLi breeder loop was constructed for components, MHD, and corrosion tests. ► A chamber for tritium extraction with a gas–liquid contact method was constructed. - Abstract: A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the common fields of a solid TBM such as design tools, structural material, fabrication methods, and He cooling technology to support this concept for the ITER. Also, other fields such as a liquid breeder technology and tritium extraction have been developed from the designed liquid TBM. For design tools, system codes for safety analysis such as Multi-dimensional Analysis of Reactor Safety (MARS) and GAs Multi-component Mixture Analysis (GAMMA) were developed for He coolant and liquid breeder. For the fabrication methods, Ferritic Martensitic Steel (FMS) to FMS and Be to FMS joinings with a Hot Isostatic Pressing (HIP) were developed and verified with a high heat flux test of up to 0.5–1.0 MW/m2. Moreover, three mockups were successfully fabricated and a 10-channel prototype is being fabricated to make a rectangular channel FW. For the integrity of the joining, two high heat flux test facilities were constructed, and one using an electron beam has been constructed. With the 6 MPa nitrogen loop, a basic heat transfer experiment for code validation was performed. From the verification of the components such as preheater and circulator, a 9 MPa He loop was constructed, and it supplies high temperature (500 °C) and pressure (8 MPa) He to the high

  16. Status of R and D on tritium permeation barrier coatings for tritium breeding blanket of fusion reactor

    International Nuclear Information System (INIS)

    The paper overviews the recent progress in the application of several typical tritium permeation barrier (TPB) coatings and their corresponding fabrication technologies for tritium breeding blanket of fusion reactor. According to the design requirements of CN HCCB-TBM and DFLL-TBM. China Institute of Atomic Energy (CIAE) and cooperation u nits have made great progresses in TPB coatings on the Reduced Activation Ferritic/Martensitic (RAFM) steel substrate. Some typical TPB coatings, such as FeAl/Al2O3, Er2O3, Y2O3 and Al2O3, have been prepared by pack aluminizing plus pre-oxidizing, Sol-Gel and MOCVD processes. The results showed that all major properties, including micro- properties, gas permeation reduction factor, electrical insulation properties etc., meet basically the requirements. (authors)

  17. Test blanket module maintenance operations between port plug and ancillary equipment unit in ITER

    International Nuclear Information System (INIS)

    In collaboration between the FZK and KFKI-RMKI, in the frame of the activities of the EU Breeder Blanket Programme a concept for test blanket module (TBM) integration, maintenance schedules and all required special purpose equipments has been developed. During the first 10 years of ITER operation four different plasma scenarios will be used. Hence it will be possible to investigate the characteristics (e.g. tritium breeding performance) of different TBM concepts which will be installed during operation for the different phases of ITER operation in the equatorial ports 2, 16 and 18. In every port two TBMs will be accommodated, in the port 16 will be the European helium-cooled pebble bed blanket. In different phases of ITER operation different TBMs will be used. Therefore a complex maintenance process is necessary for the exchange of TBMs. Two TBMs are mounted onto one common frame, into a port plug (PP), which offers a standardised interface to the vacuum vessel (VV). It is cantilevered with a flange to VV port extension. This attachment system is the same in every equatorial port, so the exchange process of this structure with the TBMs is also the standard operation of ITER. Several components of the helium cooling system of the EU breeder modules, valves, pipes, gas mixers, thermal sleeves, pipes for tritium extraction, measurement system are integrated into the ancillary equipment unit (AEU), which during the operation will connect the port plug to the subsystems. The bigger part of the AEU is accommodated in the port cell and the rest part of it is penetrated into the interspace inside the bioshield and reach the back plane of the installed PP. The remote handling operations for connection/disconnection of an interface between the PP of the EU-TBMs and the AEU are investigated with the goal to reach a quick and simple TBM exchange procedure. The current design of the EU-TBMs foresees up to 18 supply lines for both TBMs. These lines have to be connected here. A

  18. Test blanket module maintenance operations between port plug and ancillary equipment unit in ITER

    International Nuclear Information System (INIS)

    In collaboration between the FZK and KFKI-RMKI, in the frame of the activities of the EU Breeder Blanket Programme a concept for Test Blanket Module (TBM) integration, maintenance schedules and all required special purpose equipments has been developed. During the first 10 years of ITER operation 4 different plasma scenarios will be used. Hence it will be possible to investigate the characteristics (e.g. tritium breeding performance) of different TBM concepts which will be installed during operation for the different phases of ITER operation in the equatorial ports 2, 16 and 18. In every port will be two TBMs accomodated, in the port 16 will be the the European Helium Cooled Pebble Bed blanket. In the different phases of ITER operation different TBMs will be used. Therefore a complex maintenance process is necessary for exchange the TBMs. Two TBMs are mounted into one common frame, into a Port Plug (PP), which offers a standardised interface to the Vacuum Vessel (VV). It is cantilevered with a flange to VV Port Extension. This attachment system is the same in every equatorial port, so the exchange process of this structure with the TBMs are also standard operation of ITER. Several components of the Helium cooling system of the EU breeder modules, valves, pipes, gas mixers, thermal sleeves, pipes for tritium extraction, measurement system, etc. All of them is integrated into the Ancillary Equipment Unit (AEU) which during operation will connect the port plug to the sub systems. The bigger part of the AEU is accomodated in the Port Cell and the rest part of it is penetrate to the interspace inside the bioshield and reach the back plane of the installed PP. The remote handling operations for connection / disconnection of an interface between the PP of the EU-TBMs and the AEU are investigated with the goal to reach a quick and simple TBM exchange procedure. The current design of the EU-TBMs foresees up to 18 supply lines for both TBMs. These lines have to be connected

  19. In plain sight: the Chesapeake Bay crater ejecta blanket

    Directory of Open Access Journals (Sweden)

    D. L. Griscom

    2012-02-01

    Full Text Available The discovery nearly two decades ago of a 90 km-diameter impact crater below the lower Chesapeake Bay has gone unnoted by the general public because to date all published literature on the subject has described it as "buried". To the contrary, evidence is presented here that the so-called "upland deposits" that blanket ∼5000 km2 of the U.S. Middle-Atlantic Coastal Plain (M-ACP display morphologic, lithologic, and stratigraphic features consistent with their being ejecta from the 35.4 Ma Chesapeake Bay Impact Structure (CBIS and absolutely inconsistent with the prevailing belief that they are of fluvial origin. Specifically supporting impact origin are the facts that (i a 95 %-pure iron ore endemic to the upland deposits of southern Maryland, eastern Virginia, and the District of Columbia has previously been proven to be impactoclastic in origin, (ii this iron ore welds together a small percentage of well-rounded quartzite pebbles and cobbles of the upland deposits into brittle sheets interpretable as "spall plates" created in the interference-zone of the CBIS impact, (iii the predominantly non-welded upland gravels have long ago been shown to be size sorted with an extreme crater-centric gradient far too large to have been the work of rivers, but well explained as atmospheric size-sorted interference-zone ejecta, (iv new evidence is provided here that ~60 % of the non-welded quartzite pebbles and cobbles of the (lower lying gravel member of the upland deposits display planar fractures attributable to interference-zone tensile waves, (v the (overlying loam member of the upland deposits is attributable to base-surge-type deposition, (vi several exotic clasts found in a debris flow topographically below the upland deposits can only be explained as jetting-phase crater ejecta, and (vii an allogenic granite boulder found among the upland deposits is deduced to have been launched into space and sculpted by hypervelocity air friction

  20. Objectives and progress of the ITER Test Blanket Working Group activities

    International Nuclear Information System (INIS)

    The ITER Test Blanket Working Group (TBWG) has restarted its activity in October 2003. Its objectives are to define the test program of the breeding blankets in ITER, to verify its feasibility and compatibility with ITER operation, and to identify the necessary collaboration on R and D taking into account the needs of the six ITER Parties and the progress of breeding blanket technology. Despite the large required extrapolations to be made from the ITER conditions to those of a potential DEMO reactor, all parties agree on the extreme importance of breeding blanket tests in ITER. The main expected outcomes of the work of the TBWG are the establishment of a meaningful and coordinated testing program and a complete definition of the interfaces between the three equatorial ports devoted to the testing and the ITER machine and buildings. (author)

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

    International Nuclear Information System (INIS)

    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 UO2 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 UO2 has improved the agreement between the calculations and experiment, but does not account for all of the differences

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

    International Nuclear Information System (INIS)

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

  3. Acoustic contributions of a sound absorbing blanket placed in a double panel structure: Absorption Versus Transmission

    CERN Document Server

    Doutres, Olivier; 10.1121/1.3458845

    2010-01-01

    The objective of this paper is to propose a simple tool to estimate the absorption vs. transmission loss contributions of a multilayered blanket unbounded in a double panel structure and thus guide its optimization. The normal incidence airborne sound transmission loss of the double panel structure, without structure-borne connections, is written in terms of three main contributions; (i) sound transmission loss of the panels, (ii) sound transmission loss of the blanket and (iii) sound absorption due to multiple reflections inside the cavity. The method is applied to four different blankets frequently used in automotive and aeronautic applications: a non-symmetric multilayer made of a screen in sandwich between two porous layers and three symmetric porous layers having different pore geometries. It is shown that the absorption behavior of the blanket controls the acoustic behavior of the treatment at low and medium frequencies and its transmission loss at high frequencies. Acoustic treatment having poor sound ...

  4. Integrated-blanket-coil (IBC) concept applied to the OH-coil for spherical tori

    International Nuclear Information System (INIS)

    This concept combines blanket and coil functions into a single component. The objectives of the concept are to: (1) provide design options, (2) simplify overall configuration, (3) enhance compactness, and (4) reduce costs. Some drawings of the system are given

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

    International Nuclear Information System (INIS)

    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.)

  6. A high tritium breeding ratio (TBR) blanket concept and requirements for nuclear data relating to TBR

    International Nuclear Information System (INIS)

    Significance of developing a blanket having a sufficiently larger tritium breeding ratio (TBR) than 1.0 is discussed. For this purpose, a high TBR blanket with a front breeder zone just before the multiplier is introduced together with conventional blankets. From discussion of TBR characteristics in these blankets, the necessity of improving on nuclear data, i.e. reducing uncertainties is presented as follows; σs, σnp and σnα of structural and coolant materials, and σn2n of the multiplier at higher energies above several MeV, and σnγ of these materials and σnαT of 6Li at energies from several hundred keV to thermal energy. (author)

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

    International Nuclear Information System (INIS)

    A prototype (∼900H x 1700W x 350T mm) of the ITER shielding blanket module has been fabricated following the previous successful fabrication of a small-scale (∼500H x 400W x 150T mm) and mid-scale (∼800H x 500W x 350T 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 Al2O3 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 minimize thermal effects on the mechanical properties and to reduce the number

  8. Neutronics Optimization of LiPb-He Dual-Cooled Fuel Breeding Blanket for the Fusion-Driven sub-critical System

    Institute of Scientific and Technical Information of China (English)

    郑善良; 吴宜灿

    2002-01-01

    The concept of the liquid Li17Pb83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR >1.05) and annual output of 100 kg or more fissile 239Pu (FBR>0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimizated calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio ( BR = TBR + FBR ) is listed corresponding to different cases.

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

    International Nuclear Information System (INIS)

    The concept of the liquid Li17Pb83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR > 1.05) and annual output of 100 kg or more fissile 239Pu (FBR > 0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimized calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio (Br = Tbr + Fbr) is listed corresponding to different cases

  10. Wash resistance and repellent properties of Africa University mosquito blankets against mosquitoes

    OpenAIRE

    Lukwa, N; A. Makuwaza; T. Chiwade; S.L. Mutambu; M. Zimba; P. Munosiyei

    2013-01-01

    The effect of permethrin-treated Africa University (AU) mosquito blankets on susceptible female Anopheles gambiae sensu lato mosquitoes was studied under laboratory conditions at Africa University Campus in Mutare, Zimbabwe. Wash resistance (ability to retain an effective dose that kills ≥80% of mosquitoes after a number of washes) and repellence (ability to prevent ≥80% of mosquito bites) properties were studied. The AU blankets were wash resistant when 100% mortality was recorded up t...

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

    International Nuclear Information System (INIS)

    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

  12. Study of the tritium production in a 1-D blanket model with Monte Carlo methods

    OpenAIRE

    Cubí Ricart, Álvaro

    2015-01-01

    In this work a method to collapse a 3D geometry into a mono dimensional model of a fusion reactor blanket is developed and tested. Using this model, neutron and photon uxes and its energy deposition will be obtained with a Monte Carlo code. This results will allow to calculate the TBR and the thermal power of the blanket and will be able to be integrated in the AINA code.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2007-08-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mattas, R.F.; Smith, D.L. [Argonne National Lab., IL (United States); Wu, C.H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany). NET Team; Koroda, T. [Japan Atomic Energy Research Inst., Ibaraki-ken (Japan); Shatalov, G. [Kurchatov Inst. of Atomic Energy, Moscow (USSR)

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mattas, R.F.; Smith, D.L. (Argonne National Lab., IL (United States)); Wu, C.H. (Max-Planck-Institut fuer Plasmaphysik, Garching (Germany). NET Team); Koroda, T. (Japan Atomic Energy Research Inst., Ibaraki-ken (Japan)); Shatalov, G. (Kurchatov Inst. of Atomic Energy, Moscow (USSR))

    1992-01-01

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

  17. Application Effect’s Research of Vetiver Eco Blanket in Pubugou Reservoir Fluctuating Zone

    OpenAIRE

    Lan Huijuan; Zheng Kaiyuan; Ni Fuquan; Deng Yu; Liu Dengyu; Yang Xinwei; Wang Tao

    2015-01-01

    To solve the ecological disasters in Pubugou Reservoir Fluctuating Zone, ecological blanket governance model is proposed in this paper, which may provide good early environment for plants’ survival in fluctuation zone, and then play the function of greening and sustainable development to ensure the slopes’ stability. Meanwhile, based on the result of vetiver ecological blanket in Hanyuan experimental zone, we find that three kinds of typical Fluctuating Zone slope’s greening effect is good, w...

  18. Implementation of two-phase tritium models for helium bubbles in HCLL breeding blanket modules

    OpenAIRE

    Fradera, Jordi; Sedano, L.A.; Mas de les Valls Ortiz, Elisabet; Batet Miracle, Lluís

    2011-01-01

    Tritium self-sufficiency requirement of future DT fusion reactors involves large helium production rates in the breeding blankets; this might impact on the conceptual design of diverse fusion power reactor units, such as Liquid Metal (LM) blankets. Low solubility, long residence-times and high production rates create the conditions for Helium nucleation, which could mean effective T sinks in LM channels. A model for helium nano-bubble formation and tritium conjugate transport phen...

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2010-06-01

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

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

    International Nuclear Information System (INIS)

    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 Li2CO3 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 Li2CO3 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 analysis

  1. US DCLL test blanket module design and relevance to DEMO

    International Nuclear Information System (INIS)

    Full text: In the design of Test Blanket Modules (TBMs) for ITER, it is required to provide a design concept that is demonstration power reactor (DEMO) relevant. It should be noted that in the US, DEMO is defined to be a good representation of the first generation fusion power reactor. In order to evaluate the potential of the US TBM design for DEMO, a system evaluation of DEMO design was performed with an improved GA system code, and the physics results were benchmarked to ITER. With the selection of ferritic steel as the structural material, the maximum neutron wall loading is limited to 3 MW/m2. When designed to a 3 GW fusion device the optimum aspect ratio is found to be in the range of 2.5 to 3. Results show that the US dual coolant lead-lithium (DCLL) blanket can satisfy all the DEMO design requirements. On the chamber wall material, for the ITER-TBM design, the design guidance is to apply a 2 mm Be layer onto the plasma facing surface. When extrapolated to the DEMO design, the Be layer will not be suitable due to radiation damage. Similarly, a carbon surface will not be suitable due to high physical and chemical sputtering rates, radiation damage of the material and potential large retention of tritium. Unfortunately, the remaining commonly proposed material, tungsten (W), would suffer radiation damage from alpha particle implantation and, with blistering, W transport to the plasma core could severely limit the core performance. To resolve this potential impasse, different innovative options were evaluated. All high performance tokamak experiments presently use boron or silicon to condition the first wall. To use boron in DEMO, it is found that in-situ boronization will be required in order to maintain a boronized layer on the chamber wall. This boronized layer could also protect the W substrate, while retaining low-Z wall characteristics. Further innovative ideas are being evaluated to handle transient events like ELMs and disruptions. TOPICS: (PPCA) Power

  2. Preliminary neutronics design and analysis of helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Zhongliang; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Chen, Chong; Li, Min; Zhou, Guangming

    2015-06-15

    Highlights: • Neutronics design of a helium cooled solid breeder blanket for CFETR was presented. • The breeding zones parallel to FW and perpendicular to FW were optimized. • A series of neutronics analyses for the proposed blanket were shown. - Abstract: Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. Tritium self-sufficiency is one of the most important issues for CFETR and the tritium breeding ratio (TBR) is recommended not less than 1.2. As one of the candidates, a helium cooled solid breeder blanket for CFETR superconducting tokamak option was proposed. In the concept, radial arranged U-shaped breeding zones are adopted for higher TBR and simpler structure. In this work, three-dimensional neutronics design and analysis of the blanket were performed using the Monte Carlo N-Particle transport code MCNP with IAEA data library FENDL-2.1. Tritium breeding capability of the proposed blanket was assessed and the breeding zones parallel to first wall (FW) and perpendicular to FW were optimized. Meanwhile, the nuclear heating analysis and shielding performance were also presented for later thermal and structural analysis. The results showed that the blanket could well meet the tritium self-sufficiency target and the neutron shield could satisfy the design requirements.

  3. Advanced sodium cooled reactor cores having thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    In this paper, a design concept of 400 MWe sodium cooled fast reactor (SFR) cores having thorium blankets for effective burning of TRU (Transuranics) from LWR spent fuel is described. Specifically, we considered two recycling options of thorium blankets : 1) no recycling and 2) fully recycling. The thorium blankets are loaded in the axially central regions of the core regions and their axial heights are adjusted so as to increase TRU burning rate and to reduce burnup reactivity swing. Also, we analyzed the performances of the cores having different fuel management batch sizes and different recycling options for the searched core configuration. The results show that the axial thorium blankets with no recycling option can be effectively used to increase TRU burning rate with a significant reduction of burnup reactivity swing in comparison with typical SFR burner cores having no blankets while the recycling of thorium blanket degrades TRU burning rate and burnup reactivity swing but it leads to a reduction of sodium void worth and more negative Doppler coefficient. (author)

  4. Preliminary neutronics design and analysis of helium cooled solid breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Highlights: • Neutronics design of a helium cooled solid breeder blanket for CFETR was presented. • The breeding zones parallel to FW and perpendicular to FW were optimized. • A series of neutronics analyses for the proposed blanket were shown. - Abstract: Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. Tritium self-sufficiency is one of the most important issues for CFETR and the tritium breeding ratio (TBR) is recommended not less than 1.2. As one of the candidates, a helium cooled solid breeder blanket for CFETR superconducting tokamak option was proposed. In the concept, radial arranged U-shaped breeding zones are adopted for higher TBR and simpler structure. In this work, three-dimensional neutronics design and analysis of the blanket were performed using the Monte Carlo N-Particle transport code MCNP with IAEA data library FENDL-2.1. Tritium breeding capability of the proposed blanket was assessed and the breeding zones parallel to first wall (FW) and perpendicular to FW were optimized. Meanwhile, the nuclear heating analysis and shielding performance were also presented for later thermal and structural analysis. The results showed that the blanket could well meet the tritium self-sufficiency target and the neutron shield could satisfy the design requirements

  5. Simplification of blanket system for SlimCS fusion DEMO reactor

    International Nuclear Information System (INIS)

    Simplification of blanket system is necessary for a fusion DEMO reactor. Although the conceptual design of a tritium-breeding blanket for SlimCS has been studied in the past several years, the structure of the previous blanket seems to be complex and difficult to manufacture from the viewpoint of engineering. In this paper, we proposed simplification of blanket structure without decreasing the net Tritium Breeding Ratio (TBR). In the proposed concept, the blanket is filled with the mixture of Li4SiO4 pebbles or Li2O pebbles for the tritium breeding and Be12Ti pebbles for the neutron multiplication. To confirm the effectiveness of this concept, an ANIHEAT code with the nuclear library FENDL-2.0 was used for calculations of the neutronic and thermal analyses. The result indicated that, under the constraint of the blanket thickness being less than 0.5 m, the mixture of Li2O pebbles and Be12Ti ones is the most effective and that the TBR is expected to be greater than 1.05.

  6. Activation Characteristics of Fuel Breeding Blanket Module in Fusion Driven Subcritical System

    Institute of Scientific and Technical Information of China (English)

    HUANG Qun-Ying; LI Jian-Gang; CHEN Yi-Xue

    2004-01-01

    @@ Shortage of energy resources and production of long-lived radioactivity wastes from fission reactors are among the main problems which will be faced in the world in the near future. The conceptual design of a fusion driven subcritical system (FDS) is underway in Institute of Plasma Physics, Chinese Academy of Sciences. There are alternative designs for multi-functional blanket modules of the FDS, such as fuel breeding blanket module (FBB)to produce fuels for fission reactors, tritium breeding blanket module to produce the fuel, i.e. tritium, for fusion reactor and waste transmutation blanket module to try to permanently dispose of long-lived radioactivity wastes from fission reactors, etc. Activation of the fuel breeding blanket of the fusion driven subcritical system (FDS-FBB) by D-T fusion neutrons from the plasma and fission neutrons from the hybrid blanket are calculated and analysed under the neutron wall loading 0.5 MW/m2 and neutron fluence 15 MW. yr/m2. The neutron spectrum is calculated with the worldwide-used transport code MCNP/4C and activation calculations are carried out with the well known European inventory code FISPACT/99 with the latest released IAEA Fusion Evaluated Nuclear Data Library FENDL-2.0 and the ENDF/B-V uranium evaluated data. Induced radioactivities, dose rates and afterheats, etc, for different components of the FDS-FBB are compared and analysed.

  7. 18 CFR 157.212 - Synthetic and liquefied natural gas facilities.

    Science.gov (United States)

    2010-04-01

    ... natural gas facilities. 157.212 Section 157.212 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY COMMISSION, DEPARTMENT OF ENERGY REGULATIONS UNDER NATURAL GAS ACT APPLICATIONS FOR CERTIFICATES... 7 OF THE NATURAL GAS ACT Interstate Pipeline Blanket Certificates and Authorization Under Section...

  8. 76 FR 8293 - Natural Gas Pipelines; Project Cost and Annual Limits

    Science.gov (United States)

    2011-02-14

    ... Energy Regulatory Commission 18 CFR Part 157 Natural Gas Pipelines; Project Cost and Annual Limits...) computes and publishes the project cost and annual limits for natural gas pipelines blanket construction... practice and procedure, Natural Gas, Reporting and recordkeeping requirements. Jeff C. Wright,...

  9. 75 FR 8245 - Natural Gas Pipelines; Project Cost and Annual Limits

    Science.gov (United States)

    2010-02-24

    ... Energy Regulatory Commission 18 CFR Part 157 Natural Gas Pipelines; Project Cost and Annual Limits...) computes and publishes the project cost and annual limits for natural gas pipelines blanket construction... Part 157 Administrative practice and procedure, Natural gas, Reporting and recordkeeping...

  10. 77 FR 8724 - Natural Gas Pipelines; Project Cost and Annual Limits

    Science.gov (United States)

    2012-02-15

    ... Energy Regulatory Commission 18 CFR Part 157 Natural Gas Pipelines; Project Cost and Annual Limits... (OEP) computes and publishes the project cost and annual limits for natural gas pipelines blanket..., Natural gas, Reporting and recordkeeping requirements. Jeff C. Wright, Director, Office of Energy...

  11. Dust removal experiments for ITER blanket remote handling system

    International Nuclear Information System (INIS)

    To reduce maintenance workers' dose rate caused by activated dust adhering to the ITER blanket remote handling system (BRHS), dust must be removed from BRHS surfaces. Dust that adheres to the top surface of the BRHS rail from cyclic loading of the vehicle manipulator is considered to be the most difficult dust to remove. Dust removal experiments were conducted to simulate the materials, conditions, and cyclic loading of actual BRHS operations. The tungsten powder used to simulate the dust was squashed, and the area of contact by cyclic load was increased, but the powder was not embedded into the matrix. The increase in the area of contact increased the total intermolecular force between a tungsten particle and the surface, which was considered the main force adhering dust to the test piece surface. A combination of dust removal methods, including vacuum cleaning and brushing, was applied to the simulated dust that adhered to the test pieces. The results showed that vacuum cleaning is effective in removing dust from the non-cyclic loaded surface. The combined methods were highly efficient in removing the dust that strongly adhered to the rail surface. (author)

  12. Physics aspects of metal fuelled fast reactors with thorium blanket

    International Nuclear Information System (INIS)

    Metal fuelled fast breeder reactors (MFBR) with high breeding ratio will play a major role in meeting the high nuclear power growth envisaged in India. In this regard several conceptual reactor designs with alloys of U–Pu–Zr fuel have been suggested for commercial operations. This study focusses on the physics design aspects of a sodium cooled U–Pu–6%Zr fuelled 1000 MWe fast breeder reactor, which can attain a breeding ratio of nearly 1.5. The calculation results on reactor kinetics and safety parameters of the 1000 MWe MFBR are presented. The changes in the breeding ratio by introduction of thorium in the blankets of the MFBR are also investigated. Burnup analyses are carried out to compare the core burnup effects in MOX and metal fuelled FBRs. Since the MOX fuelled 500 MWe prototype fast breeder is getting constructed at IGCAR, for burnup comparisons a MFBR of similar design is considered. The results of this study indicate that the loss of reactivity in the metal core with burnup is less than half that of a MOX core and its breeding ratio remains nearly constant. It is also found that the isotopic composition of plutonium (Pu-vector composition) remains more steady with burnup in a metal core

  13. Vacuum Permeator Analysis for Extraction of Tritium from DCLL Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Humrickhouse, Paul Weston [Idaho National Laboratory; Merrill, Brad Johnson [Idaho National Laboratory

    2014-11-01

    It is envisioned that tritium will be extracted from DCLL blankets using a vacuum permeator. We derive here an analytical solution for the extraction efficiency of a permeator tube, which is a function of only two dimensionless numbers: one that indicates whether radial transport is limited in the PbLi or in the solid membrane, and another that is the ratio of axial and radial transport times in the PbLi. The permeator efficiency is maximized by decreasing the velocity and tube diameter, and increasing the tube length. This is true regardless of the mass transport correlation used; we review several here and find that they differ little, and the choice of correlation is not a source of significant uncertainty here. The PbLi solubility, on the other hand, is a large source of uncertainty, and we identify upper and lower bounds from the literature data. Under the most optimistic assumptions, we find that a ferritic steel permeator operating at 550 °C will need to be at least an order of magnitude larger in volume than previous conceptual designs using niobium and operating at higher temperatures.

  14. Probabilistic Fracture Mechanics Analysis of APT Blanket Tubes

    International Nuclear Information System (INIS)

    A probabilistic fracture mechanics (PFM) model that is specific to the Accelerator Production of Tritium (APT) helium tubes was developed. The model performs Monte Carlo analyses of potential failure modes caused by cyclic stresses generated by beam trips and de-pressurizations from normal operation, coupled with material aging due to irradiation. Dominant failure probabilities are due to crack through-growth while brittle fracture and ductile tearing have lower probability. Failure mechanisms of global plastic collapse and buckling or crack initiation mechanisms of fatigue or local fracture (upon loss of ductility) have negligible probability. For the population of (7,311) tubes in the APT blanket, the worst-case, annual probability of one tube failing is 3 percent. The probability of 2 or more failures is substantially lower; therefore, unavailability impacts are driven by single failure. The average annual loss of production (unavailability) is below about 0.2 percent. Helium outflow and water inflow rates were characterized for the failures. (authors)

  15. A hybrid blanket for a reversed-field pinch reactor

    International Nuclear Information System (INIS)

    A hybrid blanket for a reversed-field pinch (RFP) reactor is presented for breeding 233U from thorium. This study focuses on the shellblanket design and assumes a plasma like that used by Culham Laboratory (CL) in its designs. The 233U bred helps the important neutron economy and allows the tritium breeding ratio to be 0.96 at beginning of life for a mean of 1.06 for a 9 MW . yrm2 burnup. A thick conducting shell is assumed for discharge stability and field reversal. This need for a good conductor requires that only pure copper or aluminum or alloys thereof be used. Two designs were investigated, one with a pure copper first wallshell, the other with an aluminum alloy. In these designs 3.4% of the thorium is converted to 233U. In both designs the low metallurgic temperature limit means the large amount of power deposited on and in the shell is not attractive thermodynamically. The resulting large temperature differences in the shell cause high mechanical stresses. The design as it stands is not feasible from the point of view of radiation damage to materials

  16. Safety approach in the EU test blanket systems design

    International Nuclear Information System (INIS)

    Highlights: ► European Union (EU) Test Blanket Systems (TBS) safety objectives. ► EU TBS strategy. ► EU TBS project environment and status, risks and mitigations. ► EU TBS safety approach and its implementation. - Abstract: This paper reveals the safety strategy and approach developed and followed in the design of the two EU TBS describing its objectives, components and implementation. Addressing the safety in the early stage of the conceptual design of nuclear facilities is a well recognized international practice and industrial project-level requirement for the successful completion of the licensing process within expected project cost and schedule. The impact of the early development of the safety approach, its implementation and monitoring in the design of nuclear device like the TBS is not limited to the safety assessment and licensing activities only. Safety approach plays indispensible role in reducing the overall project risk. It infiltrates the entire design process through the unavoidable interfaces between the design features and its safety level. In reality the entire process of the TBS development, design, technological demonstration and implementation is affected by the project team safety culture.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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 (Li4SiO4 and Li2TiO3) with different enrichment in 6Li 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 oC for the steel, 950 o C for the breeder and 650 oC 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 Li4SiO4 breeder. The performed analyses have to be considered preliminary and an academic exercise, nevertheless they could give useful

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-22

    Lithium is often the preferred choice as breeder and coolant in fusion blankets as it offers excellent heat transfer and corrosion properties, and most importantly, it has a very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and exacerbates plant safety concerns. For this reason, over the years numerous blanket concepts have been proposed with the scope of reducing concerns associated with lithium. The European helium cooled pebble bed breeding blanket (HCPB) physically confines lithium within ceramic pebbles. The pebbles reside within a low activation martensitic ferritic steel structure and are cooled by helium. The blanket is composed of the tritium breeding lithium ceramic pebbles and neutron multiplying beryllium pebbles. Other blanket designs utilize lead to lower chemical reactivity; LiPb alone can serve as a breeder, coolant, neutron multiplier, and tritium carrier. Blankets employing LiPb coolants alongside silicon carbide structural components can achieve high plant efficiency, low afterheat, and low operation pressures. This alloy can also be used alongside of helium such as in the dual-coolant lead-lithium concept (DCLL); helium is utilized to cool the first wall and structural components made up of low-activation ferritic steel, whereas lithium-lead (LiPb) acts as a self-cooled breeder in the inner channels of the blanket. The helium-cooled steel and lead-lithium alloy are separated by flow channel inserts (usually made out of silicon carbide) which thermally insulate the self-cooled breeder region from the helium cooled steel walls. This creates a LiPb breeder with a much higher exit temperature than the steel which increases the power cycle efficiency and also lowers the magnetohydrodynamic (MHD) pressure drop [6]. Molten salt blankets with a mixture of lithium, beryllium, and fluorides (FLiBe) offer good tritium breeding

  20. Requirements for helium cooled pebble bed blanket and R and D activities

    Energy Technology Data Exchange (ETDEWEB)

    Carloni, D., E-mail: dario.carloni@kit.edu; Boccaccini, L.V.; Franza, F.; Kecskes, S.

    2014-10-15

    This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R and D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM. The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R and D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R and D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.

  1. Design criteria and mitigation options for thermal fatigue effects in ATW blankets

    International Nuclear Information System (INIS)

    Thermal fatigue due to beam interruptions is an issue that must be addressed in the design of an ATW blanket. Two different approaches can be taken to address this issue. One approach is to analyze current ATW blanket designs in order to set interrupt frequency design limits for the accelerator. The other approach is to assume that accelerator reliability can not be guaranteed before design and construction of the blanket. In this approach the blanket must be designed so as to accommodate an accelerator with a beam interruption frequency significantly higher than current high power accelerators in order to provide a margin of error. Both approaches are considered in this paper. Both a sodium cooled blanket design and a lead-bismuth cooled blanket design are considered. Thermal hydraulic analysis of the blanket for beam interruption transients is carried out with the SASSYS-1 systems analysis code to obtain the time histories of the coolant temperatures in contact with structural components. These coolant temperatures are then used in a detailed structure temperature calculation to obtain structure surface and structure average temperatures. The difference between the average temperature and the surface temperature is used to obtain thermal strains. Low cycle fatigue curves from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code are used to determine the number of cycles that the structural components can endure, based on these strains. Calculations are made for base case designs and for a number of mitigation options. The mitigation options include using two separate accelerators to provide the beam, reducing the thickness of the above core load pads in the subassemblies, increasing the coolant flow rate or reducing power in order to reduce the core temperature rise, and reducing the superheat in the once-through steam generator. (author)

  2. Physical investigation for neutron consumption and multiplication in fusion–fission hybrid test blanket module

    International Nuclear Information System (INIS)

    Highlights: • Preliminary design of hybrid test blanket module was done for ITER. • Blanket performance was analyzed for five fuel types with Li and LiPb coolants. • Detailed neutronic analysis was performed by computing neutron production and loss factors. • Inelastic neutron source factor of LiPb caused major change in blanket performance. • TiC reflector improved performance in TRU transmutation and tritium breeding. - Abstract: Inelastic scattering of high energy fusion neutrons does affect the performance of fusion blanket based on the choice of different materials. It will also affect the behavior of source neutrons in a subcritical fusion fission hybrid blanket and consequently the transmutation and tritium breeding performance. A fusion fission hybrid test blanket module (HTBM) is designed which is presumed to be tested in a large sized tokamak and plasma neutron source is similar to ITER. In this preliminary design of HTBM the neutron source and loss factors are computed for the detailed neutronic performance analysis. The neutronic analysis of hybrid blanket module is performed for five different TRU fuel types: TRU-Zr, TRU-Mo, TRU-Oxide, TRU-Carbide and TRU-Nitride. In this module design, it is aimed to burn and transmute the TRU nuclides from high-level radioactive waste of PWR spent fuel. The effect of TiC reflector on transmutation and tritium breeding performance of HTBM is also quantified. MCNPX is used for neutronic computations. Neutron spectrum, capture to fission ratio and waste transmutation ratio of each fuel type are compared to evaluate their waste transmutation performance. Tritium breeding ratio is also compared for two coolant options: Li and LiPb eutectic

  3. Experimental lithium-lead module for neutronics studies of fusion blankets

    International Nuclear Information System (INIS)

    In a (D,T) type fusion machine, about 80% of the fusion energy is transported by neutrons outside the reactor's core and deposited in the blanket, an assembly of materials surrounding the machine. Tritium breeders, such as lithium and lithium-lead (LiPb) eutectic alloys, mainly dictate the design of fusion blankets. Neutronics studies, on blanket module assemblies, form an initial step towards real construction of one or another blanket. Within the framework of this dissertation, different blanket elements: first wall/structural material, tritium breeder etc., and leading fusion blanket concepts are briefly reviewed. Lithium-lead eutectic is of particular interest since the neutron multiplication takes place in the breeder, where tritium is produced. Therefore, one-dimensional optimization calculations were performed to study the use of Li17Pb83, with natural lithium abundance. Generally, this breeder is used with very high Li6 enrichment. It was found that it would be difficult to design compact blankets and to achieve reasonable tritium production, with Li17Pb83 eutectic having natural lithium isotopic composition. Either the breeder should be highly enriched in Li6 or another enriched lithium zone should follow. This study, however, led to the design and construction of the Experimental Lithium-Lead Module (EL2M). The EL2M was also used for International Comparison on Measuring Techniques of Tritium Production Rate (TPR) for Fusion Neutronics Experiments, a program initiated by JAERI (Japan Atomic Energy Research Institute) in which eight international research institutions joined. (author) figs., tabs., 124 refs

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

    International Nuclear Information System (INIS)

    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 Li2TiO3 and so on, fabrication technology developments and characterization of the Li2TiO3 and Li4SiO4 pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li2TiO3 and Li4SiO4 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.)

  5. Requirements for helium cooled pebble bed blanket and R and D activities

    International Nuclear Information System (INIS)

    This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R and D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM. The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R and D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R and D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine

  6. Estimation of Nusselt number and the first wall heat extraction capability in self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    Self-cooled liquid metal breeding blankets are in principle the simplest type of blankets that can be used in fusion reactors. Two critical issues in such blankets are strong magnetohydrodynamic (MHD) effects and the heat extraction capability from the first wall using liquid metals. In this work we estimate from first principles the MHD friction factors and Nusselt numbers applicable to convective cooling of the first wall using liquid metals. This is followed by a parametric, thermal-hydraulic study of the first wall cooling capability in self-cooled blankets with emphasis on coolant channel dimensions, dynamic pressure drop and heat removal rate to pumping power ratio. (author)

  7. Neutronic and thermomechanical analysis of the water-cooled lithium-lead blanket design for a DEMONET reactor

    International Nuclear Information System (INIS)

    Within the framework of the European DEMO blanket study programme, CEA and the JRC of Ispra are jointly developing a water-cooled lithium-lead blanket concept. The new DEMONET reactor configuration released in Spring 1990 and currently specified in the EC programme is the basis of neutronic and thermomechanical studies for the proposed box-shaped blanket concept. Considering the high blanket coverage, it is now possible to reach tritium self-sufficiency without making use of beryllium (neutronic calculations indicate a global tritium breeding ratio of the order of 1.16). (orig.)

  8. Thermal-hydraulic design and analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    International Nuclear Information System (INIS)

    To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and being designed aiming at 50-200 MW fusion power, 30-50% duty time factor, and tritium self-sustained. Three kinds of tritium breeding blanket concepts, including helium-cooled solid blanket, water-cooled solid blanket and liquid metal-cooled liquid blanket, have been considered for CFETR. Compared to ITER test blanket module, the blanket design for CFETR is facing much more challenges due to the compulsive requirements of tritium self-sufficiency, nuclear heat removal and the space limitation for blanket installation. In this paper, a kind of helium cooled solid tritium breeder blanket was designed for CFETR full superconducting tokamak. The thermal-hydraulic designs were carried out based on the blanket structure design and neutronics calculation. The performance evaluation was conducted using ANSYS, and three-dimensional fluid-solid coupled models were modeled for the accuracy results. The results showed that the FW and BU can satisfy the design requirements. (author)

  9. Further adaptation of the European ceramic-B.I.T. blanket conceptual design to updated Demo specifications

    International Nuclear Information System (INIS)

    This paper presents the recent development studies on the adaptation of the European Ceramic Solid Breeder Inside Tube (BIT) Blanket to updated DEMO specifications. The adaptation work is in progress, since 1990, when a detailed comparison between two existing designs lead to the selection of an unique concept. The main new developments concern the separation in two parts of the inboard blanket segments at the level of the lower divertor, the consequent improvement of the blanket coverage, the simplification of maintenance operations, and finally the increased compactness of the blanket because of the inclusion of the shielding into the breeder assembly

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Water Cooled Solid Breeder Test Blanket Module (WCSB TBM) is being designed by JAEA for the primary candidate TBM of Japan, and the safety evaluation of WCSB TBM has been performed. This reports presents summary of safety evaluation activities of the Japanese WCSB TBM, including nuclear analysis, source of RI, waste evaluation, occupational radiolysis exposure (ORE), failure mode effect analysis (FMEA) and postulated initiating event (PIE). For the purpose of basic evaluation of source terms on nuclear heating and radioactivity generation, two-dimensional nuclear analysis has been carried out. By the nuclear analysis, distributions of neutron flux, tritium breeding ratio (TBR), nuclear heat, decay heat and induced activity are calculated. Tritium production is calculated by the nuclear analysis by integrating distributions of TBR values, as about 0.2 g-T/FPD. With respect to the radioactive waste, the induced activity of the irradiated TBM is estimated. For the purpose of occupational radiolysis exposure (ORE), RI inventory is estimated. Tritium inventory in pebble bed of TBM is about 3 x 1012 Bq, and tritium in purge gas is about 3 x 1011 Bq. FMEA has been carried out to identify the PIEs that need safety evaluation. PIEs are summarized into three groups, i.e., heating, pressurization and release of RI. PIEs of local heating are converged without any special cares. With respect to heating of whole module, two PIEs are selected as the most severe events, i.e., loss of cooling of TBM during plasma operation and ingress of coolant into TBM during plasma operation. With respect to PIEs about pressurization, the PIEs of pressurization of the compartment nearby the pipes of cooling system are evaluated, because rupture of the pipes result pressurization of such compartments, i.e., box structure of TBM, purge gas loop, TRS, VV, port cell and TCWS vault. Box structure of TBM is designed to withstand the maximum pressure of the cooling system. At other compartments

  12. Production enhancement and quality degradation of Pu produced in FBR blankets

    International Nuclear Information System (INIS)

    Full text: For smooth deployment of FBR in near future, securing economy and non-proliferation are pivotal factors. This study has two main objectives related to these two key factors. One is to enhance Pu production efficiency in blanket region of fast breeder reactor core. The other is Pu composition degradation to improve the proliferation resistance of FBR fuel cycle. The former contributes to reduce amount of blanket fuels to be fabricated and reprocessed to gain the same quantity of Pu, consequently it is expected to improve the fuel cycle economy. The composition of Pu generated in FBR blanket generally has very high quality, namely, equal to weapon or super grade. The latter objective is to deteriorate the quality of generated Pu by adjusting neutron spectrum in the blanket region and the initial composition of blanket fuels. A conceptual core design of MOX fueled, sodium cooled FBR with 1500MWe rating is performed by using SLAROM and CIRATION code with nuclear data set prepared for fast reactors of JFS-3-J3.2. The initial fuels in the active core region contains 5% of minor actinides produced in UO2-fueled PWR with enrichment of 4.1% and burnup of 43GWd/t after 10 years cooling. A index FIR (fissile inventory ratio at EOC and BOC) is used to measure the net fissile balance that is suited for fuel cycle mass balance analysis in addition to the BR (breeding ratio) in the standard definition. The moderator material used to tailor the neutron spectrum in the blanket region is ZrH. The depleted uranium pins in the radial blanket are replaced by ZrH pins in the volume fraction range from 0% to 30%. By increasing ZrH pins, the Pu production per unit mass of UO2 in blanket increased from 0.08kg-Pu/kg-UO2 to more than 0.1kg-Pu/kg-UO2. The FIR also slightly improved by replacing small fraction of UO2 by ZrH pins however it turned for the worse in the higher range more than 5% of ZrH. Loading ZrH into blanket fuel assembly drastically affects the Pu composition at

  13. Neutronics R and D efforts in support of the European breeder blanket development programme

    International Nuclear Information System (INIS)

    The EU fusion technology programme considers two blanket development lines, the Helium-Cooled Pebble Bed (HCPB) blanket with Lithium ceramics pebbles as breeder material and beryllium pebbles as neutron multiplier, and the Helium-Cooled Lithium-Lead (HCLL) blanket with the Pb-Li eutectic alloy acting both as breeder and neutron multiplier. The long-term strategy aims at providing validated engineering designs of breeder blankets for a fusion power demonstration reactor (DEMO). As an important intermediate step, the breeder blankets need to be tested in a real fusion environment as provided by ITER. HCPB and HCLL Test Blanket Modules (TBM) have been accordingly designed for tests in dedicated ITER blanket ports. The nuclear design and performance of the breeder blanket modules rely on the results provided by neutronics design calculations. Validated computational tools and qualified nuclear data are required for high prediction accuracies including reliable uncertainty assessments. Complementary to the application of established standard tools and data for design analysis, a dedicated neutronics R and D effort is therefore conducted in the EU. This includes the development of dedicated computational tools, the generation of high quality nuclear data and their validation through integral experiments. The recent neutronic design efforts have been devoted to the European DEMO reactor study comprising (i) Monte Carlo based pre-analysis for the dimensioning of the shielding system, (ii) the generation of a generic CAD based Monte Carlo geometry model, and (iii) performance analysis for HCLL and HCPB based DEMO variants. The recent focus of the validation effort is on neutronics TBM mock-up experiments. The first experiment of this kind was performed on a TBM mock-up of the HCPB breeder blanket. The follow-up experiment on a neutronics HCLL TBM mock-up is currently under preparation. Computational pre-analysis were performed to optimise the design of the mock

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

    International Nuclear Information System (INIS)

    This paper presents results of conceptual design activities and associated R and 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 and 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

  15. Technical evaluation of major candidate blanket systems for fusion power reactor

    International Nuclear Information System (INIS)

    The key functions required for tritium breeding blankets for a fusion power reactor are: (1) self-sufficient tritium breeding, (2) in-situ tritium recovery and low tritium inventory, (3) high temperature cooling giving a high efficiency of electricity generation and (4) thermo-mechanical reliability and simplified remote maintenance to obtain high plant availability. Blanket performance is substantially governed by materials selection. Major options of structure/breeder/coolant/neutron multiplier materials considered for the present design study are PCA/Li2O/H2O/Be, Mo-alloy/Li2O/He/Be, Mo-alloy/LiAlO2/He/Be, V-alloy/Li/Li/none, and Mo-alloy/Li/He/none. In addition, remote maintenance of blankets, tritium recovery system, heat transport and energy conversion have been investigated. In this report, technological problems and critical R and D issues for power reactor blanket development are identified and a comparison of major candidate blanket concepts is discussed in terms of the present materials data base, economic performance, prospects for future improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies. (author)

  16. Thermal Analysis on Conceptual K-DEMO Breeding Blanket on parallel flow configuration

    International Nuclear Information System (INIS)

    The feasibility study consists the design guidelines and requirements for the K-DEMO(the Korean Fusion DEMOnstration reactor). It is possible to design flexible and realistic concepts of the demonstration fusion power plant. As a part of the NFRI research, Seoul National University (SNU) is conducting thermal design, evaluation and validation of the breeding blanket for the K-DEMO reactor. Recently, a new breeding blanket concept has been proposed for the K-DEMO reactor, and preliminary feasibility studies are actively ongoing. Design concept of parallel plate-type blanket of K-DEMO and thermal limits on components was identified, in this study. It was concluded that an acceptable thermal design was achieved in the proposed breeding blanket design. However, some design improvements in the geometry of the breeding blanket are ongoing. In aspect of pressure drop and outlet temperature, some analysis need to be conducted. Stage of study on the K-DEMO reactor is in the preliminary concept definition

  17. Nuclear analysis of DEMO water-cooled blanket based on sub-critical water condition

    International Nuclear Information System (INIS)

    Highlights: ► For sub-critical water condition, the size of cooling loop would be more longer, for example, 2 m. ► Local TBR is related to the material fraction of breeders and multipliers, the beryllium is the dominant. ► Front area of blanket is dominant for blanket design and it would contribute the most of TBR comparing to the backside zones. - Abstract: For the water-cooled solid blanket of DEMO, the nuclear analysis was performed based on present cooling piping system. Especially, distributions of neutron load and temperature were calculated with Pn is 5 MW/m2. Furthermore, the local TBR was optimized by changing the material proportion for each Pn level (1–5 MW/m2). It was confirmed that the size of cooling loop for sub-critical water could be used as about 2000 × 450 mm and the cooling pipe diameter of D is 12 mm, d is 9 mm at v is 5.36 m/s. The pipe pitches would vary with Pn level which is related to the blanket structure design. Nuclear heat distribution is the base to decide the distribution of cooling pipe positions. It was found that the local TBR of blanket would be dropped down along with the Pn level rising which was mainly depended on the thickness of beryllium variation. Finally, the layout of cooling pipes for each level was obtained.

  18. Influence of nuclear data uncertainties on thorium fusion-fission hybrid blanket nucleonic performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, E.T.; Mathews, D.R.

    1979-09-01

    The fusion-fission hybrid blanket proposed for the Tandem Mirror Hybrid Reactor employs thorium metal as the fertile material. Based on the ENDF/B-IV nuclear data, the /sup 233/U and tritium production rate and blanket energy multiplication averaged over the blanket lifetime of about 9 MW-yr/m/sup 2/ are 0.76 and 1.12 per D-T neutron and 4.8, respectively. At the time of the blanket discharge, the /sup 233/U enrichment in the thorium metal is about 3%. The thorium cross sections given by the ENDF/B-IV and V were reviewed, and the important partial cross sections such as (n,2n), (n,3n), and (n,..gamma..) were found to be known to +-10 to 20% in the respective energy range of interest. A sensitivity study showed that the /sup 233/U and tritium production rate and blanket energy multiplication are relatively sensitive to the thorium capture and fission cross section uncertainties. In order to predict the above parameters within +-1%, the Th(n,..gamma..) and Th(n,..nu..f) cross sections must be measured within about +-2% in the energy range 3 to 3000 keV and 13.5 to 15 MeV, respectively.

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

    International Nuclear Information System (INIS)

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

  20. Influence of nuclear data uncertainties on thorium fusion-fission hybrid blanket nucleonic performance

    International Nuclear Information System (INIS)

    The fusion-fission hybrid blanket proposed for the Tandem Mirror Hybrid Reactor employs thorium metal as the fertile material. Based on the ENDF/B-IV nuclear data, the 233U and tritium production rate and blanket energy multiplication averaged over the blanket lifetime of about 9 MW-yr/m2 are 0.76 and 1.12 per D-T neutron and 4.8, respectively. At the time of the blanket discharge, the 233U enrichment in the thorium metal is about 3%. The thorium cross sections given by the ENDF/B-IV and V were reviewed, and the important partial cross sections such as (n,2n), (n,3n), and (n,γ) were found to be known to +-10 to 20% in the respective energy range of interest. A sensitivity study showed that the 233U and tritium production rate and blanket energy multiplication are relatively sensitive to the thorium capture and fission cross section uncertainties. In order to predict the above parameters within +-1%, the Th(n,γ) and Th(n,νf) cross sections must be measured within about +-2% in the energy range 3 to 3000 keV and 13.5 to 15 MeV, respectively

  1. 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.

  2. Sodium-cooled Fast Reactor Core Designs for the TRU burning with Thorium blanket

    International Nuclear Information System (INIS)

    In this study, the SFR(Sodium-cooled Fast Reactor) burner cores are designed with thorium blanket to have smaller burnup reactivity swing but higher TRU burning capability than the typical SFR burner cores using the TRU-U-10Zr fuel. Furthermore, we expect the SFR burner cores using thorium blanket have smaller coolant void reactivity because of the fact that the η-value increases much less with energy for 233U than for 239Pu and 232Th is less fissile than 238U. From the results, it is found that use of the thorium blanket both in inner and outer cores gives several desirable features such as the reduction of sodium void worth, small burnup reactivity swing but less negative Doppler coefficient and reduced control rod worth and that the use of thorium blanket only in the inner core gives much smaller sodium void worth but larger burnup reactivity swing than the cores using thorium blanket both in the inner and outer cores

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

    International Nuclear Information System (INIS)

    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 Li2O 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)

  4. Improved structure and long-life blanket concepts for heliotron reactors

    International Nuclear Information System (INIS)

    New design approaches are proposed for the LHD-type heliotron D-T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius over 14 m is selected to permit a blanket-shield thickness of about 1 m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity COE. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armor tiles that soften the neutron energy spectrum incident on the self-cooled Flibe-RAF blanket. In this adaptation of the Spectral-shifter and Tritium breeder Blanket (STB) concept a local tritium breeding ratio TBR over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the super-conducting magnet coils is also significantly improved. Using the constant cross sections of helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armor tiles. The key R and D issues to develop the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated. (author)

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

    International Nuclear Information System (INIS)

    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 Li2TiO3 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

  6. Neutronics and thermal design analyses of US solid breeder blanket for ITER

    International Nuclear Information System (INIS)

    The US Solid Breeder Blanket is designed to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Safety, low tritium inventory, reliability, flexibility cost, and minimum R ampersand D requirements are the other design criteria. To satisfy these criteria, the produced tritium is recovered continuously during operation and the blanket coolant operates at low pressure. Beryllium multiplier material is used to control the solid-breeder temperature. Neutronics and thermal design analyses were performed in an integrated manner to define the blanket configuration. The reference parameters of ITER including the operating scenarios, the neutron wall loading distribution and the copper stabilizer are included in the design analyses. Several analyses were performed to study the impact of the reactor parameters, blanket dimensions, material characteristics, and heat transfer coefficient at the material interfaces on the blanket performance. The design analyses and the results from the different studies are summarized. 6 refs., 3 figs., 3 tabs

  7. The integrated-blanket-coil concept applied to the spherical torus

    International Nuclear Information System (INIS)

    The purpose of this study is to investigate the potential of reactor embodiments based on the combination of two novel concepts, the integrated-blanket-coil (IBC) concept and the spherical torus (ST) concept. The IBC concept involves the combination of blanket and coil functions into a single component. The ST concept is based on the operation of the tokamak magnetic configuration at low aspect ratio. Two applications of the IBC concept to the ST are presented: (1) the IBC as the outer blanket and TF coil return legs (Application 1); and (2) the IBC as the inner blanket and OH solenoid coil set (Application 2). The application 1 IBC/ST yields reactor embodiments operating with high mass power density (≅500 kWe/MT) while, at the same time, yielding moderate neutron wall loading (≅5 MW/m/sup 2/) and modest electrical output (400-500 MWe). It is anticipated that fusion reactors operating in the above parameter space will exhibit attractive economic potential. When compared with a conventional ST design (copper TF coils and separate blanket), the Aplication 1 IBC/ST design exhibits a factor of two improvement in mass power density. The Application 2 IBC/ST provides an inductive current drive option which yields burn times in the range 3-4 minutes with no penalties in inboard breeding and energy recovery

  8. Concept for a vertical maintenance remote handling system for multi module blanket segments in DEMO

    International Nuclear Information System (INIS)

    Highlights: •A conceptual architectural model for a vertical maintenance DEMO is presented. •Novel concepts for a set of DEMO remote handling equipment are put forward. •Remote maintenance of a multi module segment blanket is found to be feasible. •The criticality of space in the vertical port is highlighted. -- Abstract: The anticipated high neutron flux, and the consequent damage to plasma-facing components in DEMO, results in the need to regularly replace the tritium breeding and radiation shielding blanket. The current European multi module segment (MMS) blanket concept favours a less invasive small port entry maintenance system over large sector transport concepts, because of the reduced impact on other tokamak systems – particularly the magnetic coils. This paper presents a novel conceptual remote maintenance strategy for a Vertical Maintenance Scheme DEMO, incorporating substantiated designs for an in-vessel mover, to detach and attach the blanket segments, and cask-housed vertical maintenance devices to open and close access ports, cut and join service connections, and extract blanket segments from the vessel. In addition, a conceptual architectural model for DEMO was generated to capture functional and spatial interfaces between the remote maintenance equipment and other systems. Areas of further study are identified in order to comprehensively establish the feasibility of the proposed maintenance system

  9. Optimized mass flow rate distribution analysis for cooling the ITER Blanket System

    International Nuclear Information System (INIS)

    Highlights: • Optimized water distribution in ITER blanket modules is presented. • All key challenging constraints are included. • The methodology and the successful result are presented. - Abstract: This paper presents the rationale to the optimization of water distribution in ITER blanket modules, meeting both Blanket System requirements and interface compliance requirements. The key challenging constraints include to: be compatible with the overall water allocation (3140 kg/s for 440 wall mounted BMs); meet the critical heat flux margin of 1.4 in the plasma facing units; meet a maximum temperature increase of 70 °C at the outlet of each single BM; and ensure that water velocity is less than 7 m/s in all manifolds, and that the pressure drops of all BMs can be equilibrated. The methodology and the successful result are presented

  10. High temperature blankets for non-electrical/electrical applications of fusion reactors: Annual report, [1983

    International Nuclear Information System (INIS)

    During FY '83 the Li2O solid-breeder, helium-cooled canister blanket emerged as the LLNL-UW choice for driving the low-temperature (2, high-temperature outer zone for driving the GA hydrogen synfuel process. Providing 3-dimensional neutronics analysis of power deposition and tritium breeding in both blankets was an important part of the UW-Rowe and Assoc. work. In both the LLNL-UW and MARS studies, the fusion driver as the Axi-Cell, A-cell version of the tandem mirror reactor (TMR). Physics parameters consistent with the synfuel interface were determined as part of the work. Defining and analyzing the thermal-electric interfaces between the TMR and the synfuel process continues to be of prime importance. The analysis of thermal transport and energy conversion in the interface, as well as thermal hydraulics analysis of the blanket, were part of the UW-Rowe Assoc. work

  11. A Seed-Blanket Plutonium-Thorium Symbiosis For The Incineration Of WEAPONS Grade Plutonium

    International Nuclear Information System (INIS)

    A PWR fuel cycle is proposed, with the purpose of eliminating weapons grade plutonium and degrading such plutonium to a proliferation resistance level. This Radkowsky Thorium-fuel Pu Incinerator cycle is based on retrofitability of core and assemblies to a Westinghouse type Pt. However the assembly is a seed-blanket unit in which the seed, fueled with a Pu/Zr alloy in a high moderator/fuel volume ratio, is supercritical, driving the subcritical blanket, loaded mainly with ThO2, generating and burning in-situ U233. Blankets are loaded once every six years. Seed undergo the usual three-batch management, one third of all seeds replaced every year. The core generates about 1100 M We

  12. Electrically insulating coatings for V-Li self-cooled blanket in a fusion system

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Reed, C. B.; Uz, M.; Park, J. H.; Smith, D. L.

    2000-05-17

    The blanket system is one of the most important components in a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The liquid-metal blanket concept requires an electrically insulating coating on the first-wall structural material to minimize the magnetohydrodynamic pressure drop that occurs during the flow of liquid metal in a magnetic field. Based on the thermodynamics of interactions between the coating and the liquid lithium on one side and the structural V-base alloy on the other side, several coating candidates are being examined to perform the insulating function over a wide range of temperatures and lithium chemistries.

  13. Further neutronic analyses of the European ceramic B.I.T. blanket for Demo

    International Nuclear Information System (INIS)

    The present study concerns the most recent neutronic analyses of two design versions of the european ceramic B.I.T. blanket, jointly developed by ENEA and CEA since few years. The last year developments required a new 3-D geometry evaluations of the global TBR (Tritium Breeding Ratio). The results indicated that the ENEA version reaches a global TBR value of 1.13. The CEA version, in a 3-D model using a simplified description of the breeder module layout, reaches a TBR value of 1.12. Nuclear heat deposition density has been determined for all blanket components as a function of the poloidal co-ordinate. Shielding properties of this type of blanket have been analyzed

  14. Tritium processing for the European test blanket systems: current status of the design and development strategy

    International Nuclear Information System (INIS)

    Tritium processing technologies of the two European Test Blanket Systems (TBS), HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed), play an essential role in meeting the main objectives of the TBS experimental campaign in ITER. The compliancy with the ITER interface requirements, in terms of space availability, service fluids, limits on tritium release, constraints on maintenance, is driving the design of the TBS tritium processing systems. Other requirements come from the characteristics of the relevant test blanket module and the scientific programme that has to be developed and implemented. This paper identifies the main requirements for the design of the TBS tritium systems and equipment and, at the same time, provides an updated overview on the current design status, mainly focusing onto the tritium extractor from Pb-16Li and TBS tritium accountancy. Considerations are also given on the possible extrapolation to DEMO breeding blanket. (authors)

  15. Assessment of the integration of a He-cooled divertor system in the power conversion system for the dual-coolant blanket concept (TW2-TRP-PPCS12D8)

    Energy Technology Data Exchange (ETDEWEB)

    Norajitra, P.; Kruessmann, R.; Malang, S.; Reimann, G.

    2002-12-01

    Application of a helium-cooled divertor together with the dual-coolant blanket concept is considered favourable for achieving a high thermal efficiency of the power plant due to its relatively high coolant outlet temperature. A new FZK He-cooled modular divertor concept with integrated pin arrays (HEMP) is introduced. Its main features and function are described in detail. The result of the thermalhydraulic analysis shows that the HEMP divertor concept has the potential of resisting, a heat flow density of at least 10-15 MW/m{sup 2} at a reachable heat transfer coefficient of approx. 60 kW/m{sup 2}K and a reasonable pumping power. Integration of this divertor concept into the power conversion system using a closed Brayton gas turbine system with three-stage compression leads to a net efficiency of the blanket/divertor cycle of about 43%. (orig.)

  16. ITER blanket module 17 shield block design and analysis

    International Nuclear Information System (INIS)

    The shield block reference design of the typical ITER blanket module has a number of grave disadvantages, precarious with relation to nuclear safety of the reactor. The main problems may arise when innage of the parallel cooling passages both in the first wall and in the shield block. Vapor locking in a radial channel with flow insert driver is very probable. Another problem, as a result of the same reason, is draining and dehydration of the coolant system. Then the highly dense packing of the radial channels in the collector array brings an essential flow irregularity. Customary as a rule, the lack of coolant is observed in the last channels, nearest to the outside, most heated surface of the shield block. A local boiling is possible in these dead spaces of coolant system. In consequence of the radial flow irregularity the cooling in the upper box header, directly under the first wall, may be extremely poor. Among the other imperfections one should note the large frontal figured lids, which overburden at welding and give to rise of stresses and shrinkages, and as a result, the large share of irreparable spoilage. The paper represents an alternative design of the shield block coolant system with predominantly sequential flow circuit. The cooling channels are drilled from the frontal side as inclined transverse holes. The open drilling ends are combined in pairs with milled grooves and welded with small lids. This gain the following advantages: the lids may have smaller thickness (7 mm instead 20 mm), the cooling passengers are placed closer to the lateral and upper sides and make cooling better, the welding stress and shrinkages are reduced, there are no any dead spaces of coolant, and the water fillup and draining are substantially improved. The listed hydraulic and thermo mechanical problems have been analysed with help of 3D models in ANSYS CFX program. The models include both the cooling space filled by water and the solid part of shield block. Thus the

  17. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    International Nuclear Information System (INIS)

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  18. Modeling of tritium transport in a pin-type solid breeder blanket

    International Nuclear Information System (INIS)

    This section of the pin-type solid breeder blanket study is the first detailed attempt at modeling tritium inventory and release within a fusion reactor blanket based on actual tritium generation and thermal hydraulic profiles, rather, than a simple average unit cell extrapolation or some assumed exponential profiles. The DIFFUSE 83 code was found to give inventory results consistent with previous modeling efforts and with the general spherical grain model. The inventories for this blanket design calculated using DIFFUSE were found to be very satisfactory, less than 14 g at steady-state for a 129,000 kg LiAlO2 blanket. This result is reasonable compared with a BCSS LiALO2 blanket inventory calculated by GA Technologies. DIFFUSE was found to be very useful in approximating tritium inventories during transient startup/shutdown modes. The evaluations of transient inventories in this study appear to be the most detailed to date. The results suggest the need for controlling coolant flow during start-up to maintain high breeder temperatures and low tritium inventory, and the use of pre-heated coolant to bake out tritium inventory after shut-down. DIFFUSE modeling of breeder pins of 100% theoretical density indicates very limited tritium release from the LiAlO2 ceramic, suggesting batch processing of the pins for tritium extraction at the end of blanket lifetime. Preliminary analysis of other surface and radiative trapping effects shows DIFFUSE to be potentially a very useful tool in approximating and evaluating experimental results. Additional DIFFUSE analysis of these effects given the available experimental data is warranted

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

    International Nuclear Information System (INIS)

    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)

  20. TRICICLO/PB. A computational tool modelling dynamic tritium transfers at HCPB demo blankets systems

    International Nuclear Information System (INIS)

    The design of the cycle and the control of tritium of DEMO breeding blankets (Inner Breeding Tritium Cycle, IBTC) represent a complex and ambitious technological objective of Fusion Nuclear Technology. The IBTC design is today conceptually open to the selection and scale demonstration of tritium processing technologies and to the choice of system design operational modes and parameters. Numerical tools modelling dynamic tritium transfers between IBTC systems based on Flow Process Diagram lay-outs support IBTC developments in many aspects serving to: (1) establish hierarchy for the IBTC design constraints and criteria, (2) to quantify on-diagram system processing technologies, (3) to fix underlying physics needed to express dynamic flux and inventories between systems, and finally (4) to make global parametric tuning and optimization of cycle parameters possible. Among the available options, the Rankine cycle is the most conservative solution for the Power Conversion Cycle in terms of technological maturity and tritium control requirements. Optimization of Gas Cooled-High Temperature Reactor and design adaptation to DEMO primary coolant (PC) [300/500 C, 80bar] permit one to assess the two general diverse coolant chemistry options (HT oxidation or H2 isotopic swamping). Both options are discussed in terms of tritium control, and internal and external IBTC processing requirements for HCPB/DEMO. Permeation from the breeding ceramic into the He primary coolant and extraction of tritium by purge gas act as given inputs for the IBTC concept. Dynamic tritium transfer and radial breeding sources are inputs for actual assessments based on 2D moving-slab numerical techniques. Ultimate tritium processing technologies performance (CPS: Coolant Purification System, TES: Tritium Extraction System from purging lines) acts as boundary IBTC design constraints. Actual limits for transient modes are discussed. The IBTC design variables concern: (i) CPS system disposition in the IBTC

  1. Blanket concept of water-cooled lithium lead with beryllium for the SlimCS fusion DEMO reactor

    International Nuclear Information System (INIS)

    As an advanced option for SlimCS blanket, conceptual design study of water-cooled lithium lead (WCLL) blanket was performed. In SlimCS, the net tritium breeding ratio (TBR) supplied from WCLL blanket was not enough because the thickness of blanket in SlimCS was limited to about 0.5 m so as to allocate the conducting shell position near the plasma for high beta access and vertical stability of plasma. Therefore, the beryllium (Be) pebble bed was adopted as additional multiplier to reach a required TBR (≥ 1.05). Considering the operating temperature of blanket materials, a double pipe structure was adopted. The nuclear and thermal analysis were carried out by a nuclear-thermal-coupled code, ANIHEAT and DOHEAT so that blanket materials were appropriately arranged to satisfy the acceptable operation temperatures. The temperatures of materials were kept in appropriate range for the neutron wall load Pn = 5 MW/m2. It was found that the local TBR of WCLL with Be blanket was comparable with that of solid breeder blanket. (author)

  2. APT Blanket Detailed Bin Model Based on Initial Plate-Type Design -3D FLOWTRAN-TF Model

    International Nuclear Information System (INIS)

    This report provides background information for a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report for the APT. This report gives a brief description of the FLOWTRAN-TF code which was used for detailed blanket bin modeling

  3. Development of Thermal-hydraulic Analysis Methodology for Multi-module Breeding Blankets in K-DEMO

    International Nuclear Information System (INIS)

    In this paper, the purpose of the analyses is to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. Afterwards, the plan for the whole blanket system analysis using MARS-KS is introduced and the result of the multiple blanket module analysis is summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for the conceptual design of the K-DEMO breeding blanket thermal analysis. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering pressure drops arises in each module. For a feasibility test of the proposed methodology, 10 outboard blankets in a toroidal field sector were simulated, which are connected with each other through the inlet and outlet common headers. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation and thanks to the parallelization using MPI, almost linear speed-up could be obtained

  4. Development of Thermal-hydraulic Analysis Methodology for Multi-module Breeding Blankets in K-DEMO

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

    In this paper, the purpose of the analyses is to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. Afterwards, the plan for the whole blanket system analysis using MARS-KS is introduced and the result of the multiple blanket module analysis is summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for the conceptual design of the K-DEMO breeding blanket thermal analysis. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering pressure drops arises in each module. For a feasibility test of the proposed methodology, 10 outboard blankets in a toroidal field sector were simulated, which are connected with each other through the inlet and outlet common headers. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation and thanks to the parallelization using MPI, almost linear speed-up could be obtained.

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

    International Nuclear Information System (INIS)

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

  6. Investigation of heat treatment conditions of structural material for blanket fabrication process

    International Nuclear Information System (INIS)

    This paper presents recent results of thermal hysteresis effects on ceramic breeder blanket structural material. Reduced activation ferritic/martensitic (RAF) steel is the leading candidates for the first wall structural materials of breeding blankets. RAF steel demonstrates superior resistance to high dose neutron irradiation, because the steel has tempered martensite structure which contains the number of sink site for radiation defects. This microstructure obtained by two-step heat treatment, first is normalizing at temperature above 1200 K and the second is tempering at temperature below 1100 K. Recent study revealed the thermal hysteresis has significant impacts on the post-irradiation mechanical properties. The breeding blanket has complicated structure, which consists of tungsten armor and thin first wall with cooling pipe. The blanket fabrication requires some high temperature joining processes. Especially hot isostatic pressing (HIP) is examined as a near-net-shape fabrication process for this structure. The process consists of heating above 1300 K and isostatic pressing at the pressure above 150 MPa followed by tempering. Moreover ceramics pebbles are packed into blanket module and the module is to be seamed by welding followed by post weld heat treatment in the final assemble process. Therefore the final microstructural features of RAFs strongly depend on the blanket fabrication process. The objective of this work is to evaluate the effects of thermal hysteresis corresponding to blanket fabrication process on RAFs microstructure in order to establish appropriate blanket fabrication process. Japanese RAFs F82H (Fe-0.1C-8Cr-2W-0.2V-0.05Ta) was investigated by metallurgical method after isochronal heat treatment up to 1473 K simulating high temperature bonding process. Although F82H showed significant grain growth after conventional solid HIP conditions (1313 K x 2 hr.), this coarse grained microstructure was refined by the post HIP normalizing at

  7. Overview of the Last Progresses for the European Test Blanket Modules Projects

    International Nuclear Information System (INIS)

    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

  8. Concept for testing fusion first wall/blanket systems in existing nuclear facilities

    International Nuclear Information System (INIS)

    A novel concept to produce a reasonable simulation of a fusion first wall/blanket test environment (except the 14 MeV neutron component) employing an existing nuclear facility is presented. Preliminary results show that an asymmetric, nuclear test environment with surface and volumetric heating rates similar to those expected in a fusion first wall/blanket or divertor chamber surface appears feasible. The proposed concept takes advantage of nuclear reactions within the annulus of a test space (15 cm in diameter and approximately 100 cm high) to provide an energy flux to the surface of a test module

  9. Waste management of first wall and blanket structural materials for tokamak fusion reactors

    International Nuclear Information System (INIS)

    A comparison has been made of the induced radioactivities in the first wall and structural materials of the breeder blanket in the high-flux region for two different fusion-reactor types. One system is the STARFIRE, a tokamak reactor with PCA, a modified stainless steel, as a first wall and a LiAlO2 breeder blanket; the other is a reactor based on the STARFIRE design with a vanadium alloy as the first wall and structural material, and circulating molten lithium as the breeder/coolant. The recycling or disposal of these structural materials is evaluated

  10. Neutronic Analysis of Flux Dispersion in a Multi-Layered, (D-T) Driven Hybrid Blanket

    OpenAIRE

    İPEK, Osman

    2000-01-01

    The concept of `hybrid blanket' is based on the placement of the nuclear fuel layer, which is a fertile material and fissionable by the fusion neutrons, at the front or the rear sides of the tritium breeding zone so that, in addition to gaining fission energy, a fissile fuel is produced. The neutronic flux distribution (neutron spectrum) along the radial direction varies with the type of material and the geometry used in the blanket, and also according to whether it is multi-layered or s...

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

    International Nuclear Information System (INIS)

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

  12. Overview of the Last Progresses for the European Test Blanket Modules Projects

    International Nuclear Information System (INIS)

    The long-term objective of the EU Breeding Blankets programme is the development of DEMO breeding blankets, which shall assure tritium self-sufficiency, an economically attractive use of the heat produced inside the blankets for electricity generation and a sufficiently high shielding of the superconducting magnets for long time operation. In the short-term so-called DEMO relevant Test Blanket Modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. The Helium Cooled Lithium-Lead (HCLL) breeder blanket and the Helium Cooled Pebble Bed (HCPB) concepts are the two breeder blanket lines presently developed by the EU. The main objective of the EU test strategy related to TBMs in ITER is to provide the necessary information for the design and fabrication of breeding blankets for a future DEMO reactor. EU TBMs shall therefore use the same structural and functional materials, apply similar fabrication technologies, and test adequate processes and components. This paper gives an overview of the last progresses in terms of system design, calculations, test program, safety and R(and)D done these last two years in order to cope with the ambitious objective to introduce two EU TBM systems for day-1 of ITER operation. The engineering design of the two systems is mostly concluded and the priority is now on the development and qualification of the fabrication technologies. From calculations point of view, the last modelling efforts related to the thermal-hydraulic of the first wall, the tritium behaviour, and the box thermal and mechanical resistance in accidental conditions are presented. Last features of the TBM and cooling system designs and integration in ITER reactor are highlighted. In particular, this paper also describes the safety and licensing analyses performed for each concept to be able to include the TBM systems in the ITER preliminary safety report

  13. Two-dimensional heating analysis of fusion blankets for synfuel production

    International Nuclear Information System (INIS)

    Fusion reactors could be used to generate electric power and produce synthetic fuels with relatively high efficiencies (about 60%). A two temperature zone blanket coupled to a high temperature electrolysis system would be used. An important parameter in this system is the ratio of the fusion neutron kerma energy absorbed by the hot interior (the higher temperature zone) to the total energy/fusion. This parameter is calculated as approximately .5 for both a one and two-dimensional model of the blanket module, and is a reasonable value for efficient energy production

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

    International Nuclear Information System (INIS)

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

  15. Activation Analysis for a He/LiPb dual Coolant Blanket for DEMO Reactor

    OpenAIRE

    Catalán, J.P.; Ogando Serrano, Francisco; Sanz Gonzalo, Javier

    2010-01-01

    The objective of the Spanish national project TECNO_FUS is to generate a conceptual design of a DCLL (Dual-Coolant Lithium-Lead) blanket for the DEMO fusion reactor. The dually-cooled breeding zone is composed of He/Pb-15.7 6Li and SiC as liquid metal flow channel inserts. Structural materials are ferritic-martensitic steel (Eurofer-97) for the blanket and austenitic steel (316LN) for the Vacuum Vessel (VV). The goal of this work is to analyze the radioactive waste production by the neutron-i...

  16. Design window calculations for a constant q' lithium blanket comparing lithium and sodium coolants

    International Nuclear Information System (INIS)

    In previous work, a design window approach has been applied to a liquid metal cooled, stagnant lithium breeding blanket, where the cooling tubes are spaced such that they all have the same heat flux per unit length (constant q'). This report is partly supplemental in that is is a detailed clarification of the equations and assumptions used, including several refinements. However, it also includes documentation for a revised version of the WINDOW code used to generate the design windows, and (as an example of the usefulness of the design window approach) a comparison of lithium cooling to sodium cooling of this blanket. The results confirm the desirability of lithium as a coolant

  17. On the use of double-walled tubes as a means to improve safety and availability of the EU DEMO Water-Cooled Pb-17Li blanket

    International Nuclear Information System (INIS)

    The impact on the blanket reliability and availability of both double-walled tube and welded joint failures in the Water-Cooled Pb-17Li Demo Single-Box blanket reference design is examined. The pertinence of employing a leak detection system is analysed and its contribution to the blanket safety and availability is evaluated. The contribution of welded joints to the blanket safety and availability is evaluated in normal operation. (orig.)

  18. Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 3. Comprehending the blanket structure

    International Nuclear Information System (INIS)

    The functions and structure design of shield blanket of ITER and test blanket module (TBM) are stated. The design of shield blanket of ITER is finished and beginning its supply. ITER shield blanket consists of the first wall and shielding block made of SUS316LN-IG, of which design and joint method are explained in details. TBM is used for engineering test of nuclear fusion reactor. The fuel breeding function, power function and structure design of TBM are described. The cooling conditions of shield blanket, TBM and fusion reactor, structure of TBM, structure of the first wall, results of stress analysis of TBM first wall, heat and mechanical behavior measurement device of pebble packed layer, the effective thermal conductivity, and relation between stress and compressive strain of Li2TiO3 pebble packed layer, and distribution of Tresca stress on the first wall are illustrated. (S.Y.)

  19. Calculation of the possible tritium production in irradiation positions of the FRJ-2(DIDO) for fusion blanket experiments

    International Nuclear Information System (INIS)

    In the field of tritium and fusion blanket technology possibly an important and early contribution to the development of a fusion reactor blanket can be obtained by irradiation experiments at the research reactor FRJ-2 in Juelich, Federal Republic of Germany. However, the tritium production rate of 0.2 x 1013 to 2 x 1013 cm-3s-1 and the power per volume of 2 to 20 W cm-3 characteristic for a fusion reactor blanket have to be realized. The present report shows the reachable tritium values calculated for different irradiation positions in the FRJ-2 for natural lithium as a breeder material considering the actual existing neutron spectrum. Based on these results we come to the conclusion that the specified blanket data can actually be reached and adjusted. Therefore irradiation experiments at the FRJ-2 would be able to supply basical results for the fusion blanket development. (orig.)

  20. 78 FR 30918 - Natural Gas Pipeline Company of America LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-05-23

    ... FERC Online Support at FERCOnlineSupport@ferc.gov or call toll- free at (866) 206-3676, or, for TTY... improve their withdrawal capacity. The stimulation will be conducted during daytime hours using truck mounted equipment. The trucks will use existing roads to access the well sites. No new ground...

  1. 75 FR 19954 - Cheniere Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-04-16

    ..., Turkey, Brazil, Argentina, Chile, Mexico, the Dominican Republic, Japan, South Korea, India, China, and... * * * authorization * * * is completely consistent with, if not mandated by, the statutory directive.'' \\5\\ See 49 FR... trade is not prohibited by U.S. law or policy, over a two year period commencing on the date of...

  2. 78 FR 13663 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-02-28

    ... its Alden Storage Field in Rice County, Kansas, all as more fully set forth in the application which..., protests, and interventions via the Internet in lieu of paper. See 18 CFR 385.2001(a)(1)(iii) and...

  3. 75 FR 60095 - Sempra LNG Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-09-29

    ... LNG supplies. \\1\\ 15 U.S.C. 717b. \\2\\ See 49 FR 6684, February 22, 1984. Sempra states that in DOE/FE... LNG receipt and storage terminals in North America, including the Cameron Terminal in Cameron Parish... protest, motion to intervene or notice of intervention, but may do so using the following process...

  4. Options and methods for instrumentation of Test Blanket Systems for experiment control and scientific mission

    International Nuclear Information System (INIS)

    Highlights: • This work defined options and methods to instrument ITER TBSs based on functional categories: safety, interlock and control and scientific exploitation based on the ITER research program. • Presented the general architecture of the HCLL and HCPB Test Blanket System Instrumentation and Control. • Defined safety and interlock sensors count and technology selection based on preliminary safety analysis. • Discussed the development status of scientific instrumentation, with focus on integration with design and fulfillment of TBM research program. - Abstract: Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept which uses the eutectic Pb-16Li as both breeder and neutron multiplier; the Helium-Cooled Pebble-Bed (HCPB) concept which features lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Each TBM is associated with several sub-systems required for their operation; together they form the Test Blanket System (TBS). This paper presents the state of HCLL and HCPB TBS instrumentation design. The discussion is based on the systems functional analysis, from which three main categories of instrumentation are defined: those relevant to safety functions; those relevant to interlock functions; those designed for the control and scientific exploitation of the devices based on the TBM program objectives

  5. 75 FR 33803 - Sabine Pipe Line LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-06-15

    ... Energy Regulatory Commission Sabine Pipe Line LLC; Notice of Request Under Blanket Authorization June 8, 2010. Take notice that on June 1, 2010, Sabine Pipe Line LLC (Sabine), 4800 Fournace Place, Bellaire... L. Kirk, Regulatory Specialist, Chevron Pipe Line Company, 4800 Fournace Place, Bellaire,...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. Transient electromagnetic and dynamic structural analyses of a blanket structure with coupling effects

    International Nuclear Information System (INIS)

    Transient electromagnetic and dynamic structural analyses of a blanket structure in the fusion experimental reactor (FER) under a plasma disruption event and a vertical displacement event (VDE) have been performed to investigate the dynamic structural characteristics and the feasibility of the structure. Coupling effects between eddy currents and dynamic deflections have also been taken into account in these analyses. In this study, the inboard blanket was employed because of our computer memory limitation. A 1/192 segment model of a full torus was analyzed using the analytical code, EDDYCUFF. In the plasma disruption event, the maximum magnetic pressure caused by eddy currents and poloidal fields was 1.2MPa. The maximum stress intensity by this magnetic pressure was 114MPa. In the VDE, the maximum magnetic pressure was 2.4MPa and the maximum stress intensity was 253MPa. This stress was somewhat beyond the allowable stress limit. Therefore, the blanket structure and support design should be reviewed to reduce the stress to a suitable value. In summary, the dynamic structural characteristics and design issues of the blanket structure have been identified. (orig.)

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

    International Nuclear Information System (INIS)

    This report is the Proceedings of ''the Sixth International Workshop on Ceramic Breeder Blanket Interactions'' which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: 1) fabrication and characterization of ceramic breeders, 2) properties data for ceramic breeders, 3) tritium release characteristics, 4) modeling of tritium behavior, 5) irradiation effects on performance behavior, 6) blanket design and R and D requirements, 7) hydrogen behavior in materials, and 8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li2TiO3, tritium release behavior of Li2TiO3 and Li2ZrO3 including tritium diffusion, modeling of tritium release from Li2ZrO3 in ITER condition, helium release behavior from Li2O, results of tritium release irradiation tests of Li4SiO4 pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  10. Overview of Helium Cooled Ceramic Reflector Test Blanket Module development in Korea

    International Nuclear Information System (INIS)

    Korea plans to install and test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in the ITER, because the HCCR blanket concept is one of options of the DEMO blanket. Currently, many design and R and D activities have been performed to develop the Korean HCCR TBM. An integrated design tool for a fusion breeder blanket has been developed based on nuclear technologies including a safety analysis for obtaining a license for testing in the ITER. A half-scale sub-module mockup of the first wall with the manifold was fabricated, and the manufacturability and thermo-hydraulic performances were evaluated. High heat load and helium cooling test facilities have been constructed. Next, the recent status of TBM material development in Korea was introduced including Reduced Activation Ferritic Martensitic (RAFM) steel, lithium ceramic pebbles and silicon carbide (SiC) coated graphite pebbles. Several fabrication methods of RAFM steel, lithium ceramic pebbles, and silicon carbide coating on graphite pebbles were investigated. Recent design and R and D progress on these areas are introduced here

  11. 77 FR 53874 - The Dow Chemical Company; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-09-04

    ... application (Application), filed on July 13, 2012, by The Dow Chemical Company (Dow), requesting blanket... comments are invited. \\1\\ The Dow Chemical Company, DOE/FE Order No. 2859 (October 5, 2010) extends through... chemical and plastics manufacturing company with operations in a number of U.S. states. Dow owns...

  12. Overview of requirements and design integration for the ITER EU Test Blanket Systems instrumentation

    International Nuclear Information System (INIS)

    The ITER project aims at building a fusion device with the general goal of demonstrating the scientific and technical feasibility of fusion power. The testing of Tritium Breeder Blanket concepts is one of the ITER missions and has been recognized as an essential milestone in the development of a future fusion reactor ensuring tritium self-sufficiency, extraction of high grade heat and electricity production. Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The strategy for the development of the instrumentation of the HCLL and HCPB Test Blanket Systems, which include the TBMs and their Ancillary Systems, is briefly recalled in this paper, along with the overview of the requirements coming from the harsh operational environment and the main challenges related to the integration with the complex design of the TBS components. (authors)

  13. Tritium confinement requirements for the water-cooled Pb-17Li blanket for demo

    International Nuclear Information System (INIS)

    In the water-cooled liquid Pb-17Li blanket concept for DEMO the limitation of tritium permeation from the breeder material into the cooling water will be required. In order to find out on what conditions this tritium permeation remains within reasonable limits, a 1-d FEM code was developed which evaluates the tritium partial pressure in Pb-17Li, the tritium inventory in the blanket material, and the tritium permeation from the Pb-17Li into the cooling water as a function of the permeation reduction factor of a barrier and the efficiency of the tritium extraction from Pb-17Li. With a parametric study the conditions were identified which allow a permeation rate of as little as 1 g.d-1 without pushing the requirements for permeation barriers and extraction efficiencies excessively far. an example is a barrier with a permeation reduction factor of 75 together with an extractor efficiency of approximately 83%. In these conditions the expected tritium inventory is attributed to approximately one third to the martensitic blanket structure (some ten grams) while two thirds will be found in the Pb-17Li. These inventory values are two orders of magnitude lower than in solid breeder blankets and are thus not considered a critical issue. 6 refs., 5 figs., 2 tabs

  14. Conceptual study of a helium cooled ceramics/Beryllium blanket for a power reactor

    International Nuclear Information System (INIS)

    In the frame of recent CEA studies aiming at the evaluation and the comparison of various candidate blanket concepts in view of their possible extrapolation to anticipated power reactor operating conditions (p/subNgreater than or equal to2 MW/m2), the present work examines the performances of a design which combines the attractive thermal performances of helium cooling in the radial direction, which minimizes the breeder temperature gradient along a cooling channel, with the promising breeding capability of composite Beryllium/LiA102 (85/15 %) breeder elements. The optimization of the neutronic and thermomechanical performances converges on a canister blanket concept, featured by a breeding capability in excess of 1.45, a pumping power of only 1 % of the thermal power and a breeder temperature distribution quasi uniform throughout the blanket (500 +- 200C) and largely independent of the power level. This unique feature provides a natural adaptation to ceramic breeders assigned to very strict and narrow working conditions and provides a valuable margin for any change in the thermal and heat transfer characteristics over the blanket lifetime

  15. Status of the EU domestic agency electromagnetic analyses of ITER vacuum vessel and blanket modules

    International Nuclear Information System (INIS)

    Highlights: Eddy and halo currents and corresponding Lorentz forces on the ITER vacuum vessel and blanket modules have been computed. VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge have been simulated. The maximum vertical force in the VV (about 120 MN downwards) is experienced in VDE-DW-SLOW cat III. For the FW panel of blanket 18 the most demanding load case is the VDE downward cat III producing a radial torque of about 110 kNm. For the FW of blanket module 10 the most demanding load case is the VDE upward exp cat III producing a poloidal torque of about 130 kNm. -- Abstract: This paper presents the results of the electromagnetic analyses of the ITER vacuum vessel and blanket modules. A wide collection of electromagnetic transients has been simulated: VDEs and MDs belonging to cat III, II and I, and a magnet fast discharge. Eddy and halo currents and corresponding Lorentz forces have been computed using 3D solid FE models implemented in ANSYS and CARIDDI. The plasma equilibrium configurations (displacement and quench of the plasma current, toroidal flux variation due to the β drop and halo currents wetting the first wall) used as an input for the EM analyses have been supplied by the 2D axisymmetric code DINA. The paper describes in detail the methodology used for the analyses and the main results obtained

  16. Identifikasi Potensi Produksi Biogas dari Limbah Cair Tahu dengan Reaktor Upflow Anaerobic Sludge Blanket (UASB)

    OpenAIRE

    WAGIMAN

    2007-01-01

    Limbah cair tahu mengandung bahan organik sehingga dapat digunakansebagai bahan baku produksi biogas. Sebagai produk samping sistempengolahan limbah secara anaerobik, biogas masih belum banyak dikajinilai tambahnya. Penelitian ini bertujuan untuk mengevaluasi potensilimbah cair tahu sebagai bahan baku produksi biogas dengan menggunkanreaktor Upflow Anaerobic Sludge Blanket (UASB). Dari aspek teknik dilihat laju produksi biogas dan laju degadasi bahan organik (COD), sedangkan dari sisi ekonomi...

  17. Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Gao, Fangfang; Zhang, Xiaokang; Pu, Yong; Zhu, Qingjun; Liu, Songlin

    2016-08-01

    Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor (CFETR) operating on a Deuterium-Tritium (D-T) fuel cycle. It is necessary to study the tritium breeding ratio (TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder (WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket, the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code (MCNP) and the fusion activation code FISPACT-2007. The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation. In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015GB108002, and 2014GB119000), and by National Natural Science Foundation of China (No. 11175207)

  18. Neutron measurements in the core and blankets of the reactor Rapsodie

    International Nuclear Information System (INIS)

    Beside a brief general discussion, the report contains all the core and blanket neutronic measurements. It covers successively the methods, the measurements themselves and the results. The later concern: spectral indexes, axial and radial fission rates, activation foil measurements and neutronic power determination. (authors)

  19. 78 FR 44558 - Stingray Pipeline Company, L.L.C.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-07-24

    ... Energy Regulatory Commission Stingray Pipeline Company, L.L.C.; Notice of Request Under Blanket Authorization Take notice that on July 3, 2013, Stingray Pipeline Company, L.L.C. (Stingray), 1100 Louisiana... in the federal waters offshore Louisiana. Specifically, Stingray proposes to abandon, by sale, its...

  20. Salted lamb meat blanket of Petrolina-Pernambuco, Brazil: process and quality

    Directory of Open Access Journals (Sweden)

    Nely de Almeida Pedrosa

    2014-03-01

    Full Text Available Salted lamb meat blanket, originated from boning, salting, and drying of whole lamb carcass, was studied aiming at obtaining information that support the search for guarantees of origin for this typical regional product from the city of Petrolina-Pernambuco-Brazil. Data from three processing units were obtained, where it was observed the use of a traditional local technology that uses salting, an ancient preservation method; however, with a peculiar boning technique, resulting in a meat product with great potential for exploitation in the form of meat blanket. Based on the values of pH (6.22 ± 0.22, water activity (0.97 ± 0.02, and moisture (69.86 ± 2.26 lamb meat blanket is considered a perishable product, and consequently it requires the use of other preservation methods combined with salt, which along with the results of the microbiological analyses (absence of Salmonella sp, score <10 MPN/g of halophilic bacteria, total coliforms between 6.7 × 10³ and 5.2 × 10(6 FUC/g, and Staphylococcus from 8.1 × 10³ CFU/g at uncountable reinforce the need of hygienic practices to ensure product safety. These results, together with the product notoriety and the organization of the sector are important factors in achieving Geographical Indication of the Salted lamb Meat blanket of Petrolina.

  1. 32 CFR Appendix C to Part 327 - DeCA Blanket Routine Uses

    Science.gov (United States)

    2010-07-01

    ... Blanket Routine Uses (a) Routine Use—Law Enforcement. If a system of records maintained by a DoD Component, to carry out its functions, indicates a violation or potential violation of law, whether civil... decision concerning the hiring or retention of an employee, the issuance of a security clearance,...

  2. 32 CFR Appendix D to Part 505 - Exemptions; Exceptions; and DoD Blanket Routine Uses

    Science.gov (United States)

    2010-07-01

    ... 32 National Defense 3 2010-07-01 2010-07-01 true Exemptions; Exceptions; and DoD Blanket Routine Uses D Appendix D to Part 505 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY AID OF CIVIL AUTHORITIES AND PUBLIC RELATIONS ARMY PRIVACY ACT PROGRAM Pt. 505, App. D Appendix D...

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

    International Nuclear Information System (INIS)

    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

  4. A note on the problems of plasma-neutral gas interaction

    International Nuclear Information System (INIS)

    The presence of an immersed neutral gas affects the equilibrium and the instability and wave phenomena of magnetically confined plasmas. In this note attention is drawn to some areas for further research on plasma-neutral gas interaction, including the spatial distributions of plasma and neutral gas, relaxation times, constraints and limitations of gas blanket systems, interaction with radiation, oscillations and wave phenomena, as well as instabilities. (Auth.)

  5. Integral neutronics experiments in analytical mockups for blanket of a hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Rong, E-mail: liurongzy@163.com; Zhu, Tonghua; Lu, Xinxin; Wang, Xinhua; Yan, Xiaosong; Feng, Song; Yang, Yiwei; Wang, Mei; Jiang, Li

    2014-12-15

    Highlights: • For checking property of the hybrid blanket by integral experiments, three mockups are established. • In spherical mockup with depleted uranium and cubic mockup with natural uranium, the plutonium production rates and uranium fission rates are measured. • In spherical mockup with depleted uranium and LiPb, tritium production rates are measured. • The measured results are compared to the calculated ones with MCNP-4B code and ENDF/B-VI library data. - Abstract: The paper describes recent progress in integral neutronics experiments in the analytical mockups for the blanket in a fusion-fission hybrid energy reactor. A conceptual blanket of the hybrid reactor is mainly loaded with natural uranium and lithium material. In the fission fuel region, uranium material and light water are arranged alternately. The mockups of the conceptual blanket are designed and used for checking neutron property of the blanket by integral experiments. Based on materials available, the spherical fission mockup for fission research and plutonium production consists of three layers of depleted uranium shells and several layers of polyethylene and graphite shells. The spherical lithium mockup for tritium production consists of depleted uranium and LiPb alloy shells. The cubic mockup consists of natural uranium and polyethylene and its structure is basically consistent with one of the fuel region. In the mockups with the D-T neutron source, the plutonium production rates, uranium fission rates and tritium production rates are measured, separately. The measured results are compared to the calculated ones with MCNP-4B code and ENDF/B-VI library data.

  6. Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

    International Nuclear Information System (INIS)

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (about 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)

  7. Initial meetings of the re-established Test Blanket Working Group

    International Nuclear Information System (INIS)

    The ITER Test Blanket Working Group (TBWG) was first established in 1995. Its activities covered successively the final part of the ITER EDA and the extension period, the main results being a preliminary assessment of the breeding blanket testing capabilities of ITER and a proposal of a coherent test blanket programme, reported in 2001, that optimized the sharing of the three available testing ports between the three Parties present in 2001 (EU, JA and RF) taking into account the different coolant characteristics. The TBWG was re-established by the ITER Interim Project Leader in September 2003, with the support of the Participant Team Leaders. It is now comprised of four members from the ITER International Team and up to three members from each of the six ITER Participant Teams. The International Team delegation is led by Dr. V. Chuyanov, who has also been appointed as TBWG Co-Chair, while the six Participant Team delegations are led by Prof. M. Abdou (US), Dr. M. Akiba (JA), Dr. A. Cardella (EU), Dr. B.G. Hong (KO), Dr. C. Pan (CN) and Dr.Y. Strebkov (RF). The revised TBWG charter defines the four missions of the activities: i) provide the Design Description Document (DDD) of the Test Blanket Module (TBM) systems proposed by the participants, including the description of the interfaces with the main ITER machine, ii) promote cooperation among participants on the associated R and D programmes, iii) verify the integration of TBM testing in ITER site safety and environmental evaluations, and finally, iv) develop and propose coordinated TBM test programmes taking into account ITER operation planning. TBMs have to be representative of the breeding blanket for DEMO (the next reactor after ITER), capable of ensuring tritium-breeding self-sufficiency and of accommodating high-grade coolants for electricity production

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

    International Nuclear Information System (INIS)

    The structural design study of a reactor module for a swimming pool type reactor (SPTR) was conducted. Since pool water plays the role of radiation shielding in the SPTR, the module does not have a solid shield. It consists of tritium breeding blankets, divertor collector plates and a vacuum vessel. The object of this study is to show the reactor module design which has a simple structure and a sufficient tritium breeding ratio. A large coverage of the plasma chamber surface with tritium breeding blanket is essential in order to obtain a high tritium breeding ratio. A breeding blanket is also placed behind the divertor collector plate, i.e. in the upper and lower region, as well as in the outboard and inboard regions of the module. A concept in which the first wall is an integral part of the blanket is employed to minimize the thickness of structural and cooling material brazed in front of the breeding material (Li2O) and to enhance the tritium breeding capability. In order to simplify the module structure the vacuum vessel and breeding blanket is also integrated in the inboard region. One of the features inherent in the swimming pool type reactor is an additional external force on the vacuum vessel, namely hydraulic pressure. A detailed structural analysis of the vacuum vessel is performed. Divertor collector plates are assemblies of co-axial tubes. They minimize the electromagnetic force on the plate induced by the plasma disruption. A thermal and structural analysis and life time estimation of the first wall and divertor collector plates are performed. (author)

  9. Development of a dynamic model of a Upflow Anaerobic Sludge Blanket (UASB) reactor

    International Nuclear Information System (INIS)

    'Full text:' The Upflow Anaerobic Sludge Blanket Reactor has been modeled in this paper using a semi-empirical correlation. The main objective of the modeling exercise was to examine the dynamics (hydrodynamics and transformation) of critical variables that are important for the reactor process design, performance and operation. These factors include some microbiological variables such as soluble (as COD) and biodegradable organics (as BOD), suspended solids (SS), and biomass concentration (as VSS). The hydraulic variables such as the upflow velocity of the liquid and by extension, upflow velocity of the solids and the volumetric fractions of the liquid (voidage) are also given considerations. The dynamics of external mass transfer of organics and inorganics and the transformation reactions within the reactor are modeled by five non-linear partial differential equations (PDE's), which define dependent variables in one-dimensional temporal and spatial planes, one integrand equation for the gas rate of flow and six algebraic equations that define the reactor performance (conversion rates and removal efficiencies). The PDE's are solved using backward finite differencing method. The discretized equations are computed using a simulation program that was written in the Borland C++ programming language. The model defines the dynamics of the UASB in a simplistic way but which, nonetheless, can be useful for engineering designs where the interest is in qualitative results and not quantitative outcomes. The simulation was validated with data derived from the operation of a pilot model UASB. The pilot model was operated using raw domestic wastewater as substrate, which had an average total COD of 492 mg/L, soluble COD of 172 mg/L, BOD of 220 mg/L, suspended solid of 0.235 g/L and a high sulphate content of 108 mg/L. The average nutrient concentrations were, NH3-N 20 mg/L and PO4 15 mg/L. The UASB was operated at HRT's of 10.968, 8.326 and 5.51 hours The average removal

  10. Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

    2014-06-15

    Highlights: • Detailed burnup calculations are performed on pressurized water cooled blankets with pressure tube assemblies. • The blanket is fueled with simple fuel, namely spent nuclear fuel discharged from light water reactors or natural uranium oxide. • The refueling strategies are proposed, and the uranium resource utilization rate can reach 5–6%. - Abstract: A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.

  11. Tritium self-sufficiency time and inventory evolution for solid-type breeding blanket materials for DEMO

    Science.gov (United States)

    Packer, L. W.; Pampin, R.; Zheng, S.

    2011-10-01

    One of the primary functions of a fusion blanket is to generate enough tritium to make a fusion power plant (FPP) self-sufficient. To ensure that there is satisfactory tritium production in a real plant the tritium breeding ratio (TBR) in the blanket must be greater than 1 + M, where M is the breeding margin. For solid-type blanket designs, the initial TBR must be significantly higher than 1 + M, since the blanket TBR will be reduced over time as the lithium fuel is consumed. The rate of TBR reduction will impact on the overall blanket self-sufficiency time, the time in which the net tritium inventory of the system is positive. DEMO relevant blanket materials, Li 4SiO 4 and Li 2TiO 3, are investigated by computational simulation using radiation transport tools coupled with time-dependent inventory calculations. The results include tritium inventory assessments and depletion of breeding materials over time, which enable self-sufficiency times and maximum surplus tritium inventories to be evaluated, which are essential quantities to determine to allow one to design a credible FPP using solid-type breeding material concepts. The blanket concepts investigated show self-sufficiency times of several years in some cases and maximum surplus inventories of up to a few tens of kg.

  12. Integral experiment on effects of large opening in fusion reactor blanket on tritium breeding using annular geometry

    International Nuclear Information System (INIS)

    An experiment involving a simulated blanket with an opening has been performed using the line source and annular blanket system developed under the JAERI/USDOE collaborative programme in order to examine the effects of the opening on neutronics parameters such as the tritium-breeding ratio. The annular test assembly was rectangular in shape and consisted of a lithium oxide blanket covered with graphite and SS304 which simulated the graphite armour plate and first wall in a fusion device. A large opening (376mm x 425.5mm) was made in the middle of the test blanket. This opening simulated a neutral beam injector.Tritium production rates and reaction rates were measured inside the blanket. Neutron spectra and reaction rates were also measured on the surfaces of both sides and without the opening of the inner cavity. The opening decreased the number of low energy neutrons contained in the cavity and especially decreased 6Li tritium production by 10% inside the blanket at the opposite side of the opening. The Monte Carlo code GMVP using the JENDL-3 nuclear data library predicted the measured nuclear parameters in the test blankets, such as the tritium production rate, to within 10% accuracy. (orig.)

  13. Neutronic evaluation of fissile fuel breeding blankets for the fission-suppressed Tandem-Mirror Hybrid Reactor

    International Nuclear Information System (INIS)

    A computational study was performed on the blanket design of the Lawrence Livermore National Laboratory (LLNL) fission-suppressed Tandem Mirror Hybrid Reactor (TMHR) to qualify the methods and data bases available at Oak Ridge National Laboratory (ORNL) for use in analyzing the neutronic performance of fissile fuel breeding blankets. The eventual goal of the study was to establish the capability for analysis and optimization of advanced fissile fuel production blanket designs. Discrete ordinates radiation transport calculations were performed in one-dimensional cylindrical geometry to obtain the blanket spatial distribution and energy spectra of the neutron and gamma-ray fluxes resulting from the monoenergetic (14.1 MeV) fusion first wall source. Key macroscopic cross sections of the blanket materials were then folded with the flux spectra to obtain reaction rates critical to evaluating blanket feasibility. Finally, a time-dependent depletion analysis was performed to evaluate the blanket performance during equilibrium cycle conditions. The results of the study are presented both as graphs and tables

  14. Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Highlights: • Detailed burnup calculations are performed on pressurized water cooled blankets with pressure tube assemblies. • The blanket is fueled with simple fuel, namely spent nuclear fuel discharged from light water reactors or natural uranium oxide. • The refueling strategies are proposed, and the uranium resource utilization rate can reach 5–6%. - Abstract: A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%

  15. Thermal Protection with 5% Dextrose Solution Blanket During Radiofrequency Ablation

    OpenAIRE

    Chen, Enn Alexandria; Neeman, Ziv; Lee, Fred. T.; Kam, Anthony; Wood, Brad

    2006-01-01

    A serious complication for any thermal radiofrequency ablation is thermal injury to adjacent structures, particularly the bowel, which can result in additional major surgery or death. Several methods using air, gas, fluid, or thermometry to protect adjacent structures from thermal injury have been reported. In the cases presented in this report, 5% dextrose water (D5W) was instilled to prevent injury to the bowel and diaphragm during radiofrequency ablation. Creating an Insulating envelope or...

  16. Development of high temperature fusion blanket with LiPb-SiC and its socio-economic aspects

    International Nuclear Information System (INIS)

    This paper describes recent results of the development of SiC-LiPb blanket in Kyoto University with advanced SiC composite, and its implication for high temperature blanket preferable from socio-economic aspects. Blanket is the interface between fusion energy and outside world, i.e., industry, public and environment. Material and energy balances, such as fuel supply and waste discharge, or supply of product energy from fusion plants are characterized by the specific blanket concepts, and performance and characteristics of blankets are considered to dominate the feature of fusion energy that should respond to the requirements of the sponsors and public. Thus development strategy for blanket concepts are affected by the aspects of socio-economics such as; environmental release, assumed accidental scenario, rad-waste, and deployment to the market in each countries/area. Selections of breeders, coolants, and structural materials are based on such consideration, and ITER/TBM is expected to foster the evolving concepts. Combination of LiPb, helium and SiC is of particular interests for a demo blanket concept, because it is expected to be feasible in early stage of development such as TBM by Dual Coolant Lithium Lead concept as an intermediate step, and would eventually achieve high operating temperature above current fission reactors'. In the power plant design in Europe, US and Japan, Lithium lead will work as breeder and primary heat medium, and SiC composites are used for flow insert, enclosure and Intermediate Heat Exchanger to generate high temperature helium that is considered as a medium for power generation including possible hydrogen production. Staged development strategy can be planned; and experiments to verify compatibility, tritium permeability, MHD pressure drop and heat transfer characteristics are being evaluated with LiPb-He loop in Kyoto University. Hydrogen production process with this blanket is also experimentally studied. Based on the results

  17. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    International Nuclear Information System (INIS)

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits. (fusion engineering)

  18. A Thermal/Hydraulic Safety Assessment of the Blanket Conceptual Design for the Accelerator Production of Tritium Facility

    International Nuclear Information System (INIS)

    In support of the Accelerator Production of Tritium (APT) project, safety analyses for the blanket system have been performed based on the conceptual design for the Target/Blanket (T/B) Facility. During mitigated event sequences safety engineered features, such as the residual heat removal (RHR) and cavity flood systems, provide sufficient protection for maintaining the structural integrity of the blanket system and its components. During unmitigated (with beam shutdown only) event sequences, passive features such as natural circulation, thermal inertia, and boil-off provide significant time for corrective measures to be taken

  19. Improving proliferation resistance of high breeding gain generation 4 reactors using blankets composed of light water reactor waste

    Energy Technology Data Exchange (ETDEWEB)

    Hellesen, C.; Grape, S.; Haakanson, A.; Jacobson Svaerd, S.; Jansson, P. [Division of Applied Nuclear Physics, Uppsala University, Aangstroemlaboratoriet Laegerhyddsvaegen 1, 751 20 Uppsala (Sweden)

    2013-07-01

    Fertile blankets can be used in fast reactors to enhance the breeding gain as well as the passive safety characteristics. However, such blankets typically result in the production of weapons grade plutonium. For this reason they are often excluded from Generation IV reactor designs. In this paper we demonstrate that using blankets manufactured directly from spent light water (LWR) reactor fuel it is possible to produce a plutonium product with non-proliferation characteristics on a par with spent LWR fuel of 30-50 MWd/kg burnup. The beneficial breeding and safety characteristics are retained. (authors)

  20. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    Science.gov (United States)

    Ma, Xuebin; Liu, Songlin; Li, Jia; Pu, Yong; Chen, Xiangcun

    2014-04-01

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.

  1. First adaptation of the European ceramic B.I.T. blanket design to the updated DEMO specifications

    International Nuclear Information System (INIS)

    The DEMO specifications defined so as to ensure the consistency of the various blanket conceptual design studies performed within the framework of the European Test Blanket Programme have been recently updated. A very first attempt has been made to adapt the European Ceramic Breeder Inside-Tube DEMO blanket to these new specifications. Two solutions have been investigated. The first would ensure tritium self-sufficiency of the plant with a large safety margin. The other one, which fully preserves the design simplicity and reliability of the initial design, appears to be somewhat marginal from the tritium breeding capability point of view, but to offer good improvement prospects. (orig.)

  2. MHD issues related to the use of Lithium Lead eutectic as breeder material for blankets of fusion power plants

    OpenAIRE

    Aiello, G.; Mistrangelo, C.; Buehler, L; Mas de les Valls Ortiz, Elisabet; Aubert, J. -J.; Li-Puma, A.; Rapisarda, David; Del Nevo, A.

    2015-01-01

    The European Community is committed to the development of a DEMOnstration fusion power plant whose operation could start as soon as 2050. The blanket is one of the most critical components in a fusion reactor; three of the four blanket concepts currently under development are based on the use of the liquid eutectic alloy Pb-15.7Li. Since the blanket will operate under the strong magnetic eld used to con ne the plasma, electromagnetic forces will occur in the PbLi ow, giving rise to magnetohy...

  3. Semi-Technical Cryogenic Molecular Sieve Bed for the Tritium Extraction System of the Test Blanket Module for ITER

    International Nuclear Information System (INIS)

    The tritium extraction from the ITER Helium Cooled Pebble Bed (HCPB) Test Blanket Module purge gas is proposed to be performed in a two steps process: trapping water in a cryogenic Cold Trap, and adsorption of hydrogen isotopes (H2, HT, T2) as well as impurities (N2, O2) in a Cryogenic Molecular Sieve Bed (CMSB) at 77K. A CMSB in a semi-technical scale (one-sixth of the flow rate of the ITER-HCPB) was design and constructed at the Forschungszentrum Karlsruhe. The full capacity of CMSB filled with 20 kg of MS-5A was calculated based on adsorption isotherm data to be 9.4 mol of H2 at partial pressure 120 Pa. The breakthrough tests at flow rates up to 2 Nm3h-1 of He with 110 Pa of H2 conformed with good agreement the adsorption capacity of the CMSB. The mass-transfer zone was found to be relatively narrow (12.5 % of the MS Bed height) allowing to scale up the CMSB to ITER flow rates

  4. Thermal Protection with 5% Dextrose Solution Blanket During Radiofrequency Ablation

    International Nuclear Information System (INIS)

    A serious complication for any thermal radiofrequency ablation is thermal injury to adjacent structures, particularly the bowel, which can result in additional major surgery or death. Several methods using air, gas, fluid, or thermometry to protect adjacent structures from thermal injury have been reported. In the cases presented in this report, 5% dextrose water (D5W) was instilled to prevent injury to the bowel and diaphragm during radiofrequency ablation. Creating an Insulating envelope or moving organs with D5W might reduce risk for complications such as bowel perforation

  5. 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

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

    International Nuclear Information System (INIS)

    Neutronics calculations were performed to analysis the parameters of blanket energy multiplication factor (M) and tritium breeding ratio (TBR) in a fusion-fission hybrid reactor for energy production named FDS-EM (Energy Multiplier) blanket. The most significant and main goal of the water-cooled FDS-EM blanket is to achieve the energy gain of about 1 GW with self-sustaining tritium, which can operate for as long as possible without fuel unloading and reloading. The preliminarily designed neutronics parameters for FDS-EM were presented, which show that the blanket loaded with the Nuclear Waste (transuranic from 33 000 MWD/MTU PWR and depleted uranium) for energy multiplication (M≅90) with tritium self-sufficiency can operate for at least 10 years without fuel unloading and reloading. (authors)

  7. Status of R&D on Tritium Permeation Barrier Coatings for Tritium Breeding Blanket of Fusion Reactor

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The paper overviewed the recent progress in the application of several typical tritium permeation barrier (TPB) coatings and their corresponding fabrication technologies for tritium breeding blanket of fusion reactor. According to the design requirements of

  8. Development of a two-dimensional nuclear-thermal-coupled analysis code for conceptual blanket design of fusion reactors

    International Nuclear Information System (INIS)

    For conceptual blanket design of fusion reactors, a two-dimensional (2-D) nuclear-thermal-coupled analysis code, DOHEAT, has been developed. A remarkable feature of the code is a user-friendly graphical interface which allows to prepare an appropriate analysis model of a variety of blanket concepts with different materials in a short time. Furthermore, the code has a strong point that it provides the cross-sectional distribution of temperature, neutron/gamma fluxes, nuclear heating and tritium breeding ratio (TBR) inside the blanket after the completion of a single run. Throughout the DEMO design work in JAEA, it has proven that DOHEAT is extremely valuable in guiding the design study toward a wide variety of blanket concepts.

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

    International Nuclear Information System (INIS)

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

  10. Microstructure and hardness of HIP-bonded regions in F82H blanket structures

    International Nuclear Information System (INIS)

    Metallurgical examinations and hardness measurements were performed at hot isostatic pressing (HIP)-bonded regions in blanket structures made from F82H alloy in order to investigate the HIP-bondability and the influence on the microstructure due to the HIP and heat treatments which would correspond to the fabrication of an actual blanket. The metallurgical examination showed that the HIP-bonded interfaces were sufficiently diffusion-bonded without significant defects, i.e. voids and/or exfoliations, although grain coarsening was observed at a part of the HIP interfaces. Hardness was nearly equal in the coarsening region and a region without coarsening, but about a 10 Hv increase was found in a boundary in between the regions with and without coarsening. Microcrystallized grains were observed in a region about ∼6 μm from HIP interfaces, and the hardness increased by about 0.2 GPa in the region

  11. Efficient COD removal and nitrification in an upflow microaerobic sludge blanket reactor for domestic wastewater.

    Science.gov (United States)

    Zheng, Shaokui; Cui, Cancan

    2012-03-01

    The treatment performance of an upflow microaerobic sludge blanket reactor (UMSB) for synthetic domestic wastewater was investigated at two dissolved oxygen (DO) levels, 0.3-0.5 and 0.7-0.9 mg l(-1), focusing on nitrification performance. The higher DO level induced complete nitrification of ammonia nitrogen (NH(3)-N), achieving chemical oxygen demand and NH(3)-N removals of 97 and 92%, respectively. There were consistently significantly higher nitrate nitrogen (NO(3)-N) and nitrite nitrogen (NO(2)-N) levels in the effluent, with ~66% of newly-produced oxidised nitrogen as NO(2)-N. Despite the high nitrification efficiency, only about 23% of the removed NH(3)-N amount from the influent was ultimately transformed into oxidised nitrogen due to the simultaneous nitrification-denitrification. Sludge blanket development and granulation occurred simultaneously in the UMSB. PMID:22105554

  12. An electro-hydraulic servo control system research for CFETR blanket RH

    International Nuclear Information System (INIS)

    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

  13. Integral approach for neutronics analyses of the European test blanket modules in ITER

    International Nuclear Information System (INIS)

    An advanced integral approach has been implemented for neutronic analyses of the European test blanket modules (TBMs) in ITER. The central element of this approach is the use of the geometry conversion tool McCad for the generation of Monte Carlo analysis models from CAD geometry data. Following this approach, an MCNP model of the test blanket port plug with HCPB and HCLL TBM assemblies, elaborated by the European TBM Consortium of Associates (CA), was generated and integrated into the Alite MCNP model of ITER. Neutronic performance and shielding analyses were conducted on the basis of MCNP-5 calculations for the HCPB and HCLL TBMs and the entire shield system. The results indicated the need for a further optimization of the shield system complemented by a rigorous shutdown dose rate analysis.

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

    International Nuclear Information System (INIS)

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

  15. Applicability of tungsten/EUROFER blanket module for the DEMO first wall

    International Nuclear Information System (INIS)

    In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ∼14 MW/m2. It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface

  16. Tauro: a ceramic composite structural material self-cooled Pb-17Li breeder blanket concept

    International Nuclear Information System (INIS)

    The use of a low-activation (LA) ceramic composite (CC) as structural material appears essential to demonstrate the potential of fusion power reactors for being inherently or, at least, passively safe. Tauro is a self-cooled Pb-17Li breeder blanket with a SiC/SiC composite as structure. This study determines the required improvements for existing industrial LA composites (mainly SiC/SiC) in order to render them acceptable for blanket operating conditions. 3D SiC/SiC CC, recently launched on the market, is a promising candidate. A preliminary evaluation of a possible joining technique for SiC/SiC is also described. (orig.)

  17. Benchmark experiment on the model of fusion reactor blanket with uranium neutron multiplier

    International Nuclear Information System (INIS)

    Benchmark experiment on the model of thermonuclear reactor blanket with 14 MeV neutron source is described. The model design corresponds to the known concept of the fast hybrid blanket with 238U neutron multiplier and main tritium production on 6Li. Detailed measurements of the following process velocities were carried out: tritium production on lithium isotopes; reactions modelling tritium production; (n, γ) and (n, 2n) processes for 238U; fission reactions for 235,238U, 239Pu, 237Np. Neutron flux integral measurements were performed by a set of threshold detectors on the basic of the 115In(n, n'), 204Pb(n, n'), 64Zn(n, p), 27Al(n, p), 56Fe(n, p), 107Ag(n, 2n), 63Cu(n, 2n) and 64(n, 2n) reactions

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-01-01

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

  20. 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.