WorldWideScience

Sample records for patient blanket preheaters

  1. Gravity mediated preheating

    International Nuclear Information System (INIS)

    Maity, Debaprasad

    2015-01-01

    In this work we propose a mechanism of natural preheating of our universe induced by the inflation field dependent effective mass term for the gravitational wave. For any single field inflationary model, the inflation must go through the oscillatory phase after the end of inflation. As has recently been shown, if the gravitational fluctuation has inflation dependent mass term, there will be a resonant amplification of the amplitude of the gravitational wave during the oscillatory phase of inflation though parametric resonance. Because of this large enhancement of the amplitude of the gravitational wave, we show that universe can be naturally pre-heated through a minimally coupled matter field with gravity. Therefore, during the pre-heating phase, there is no need to introduce any arbitrary coupling between the matter field and the inflation. (author)

  2. Inflation After Preheating

    CERN Document Server

    Felder, G; Linde, Andrei D; Tkachev, Igor I; Felder, Gary; Kofman, Lev; Linde, Andrei; Tkachev, Igor

    2000-01-01

    Preheating after inflation may lead to nonthermal phase transitions with symmetry restoration. These phase transitions may occur even if the total energy density of fluctuations produced during reheating is relatively small as compared with the vacuum energy in the state with restored symmetry. As a result, in some inflationary models one encounters a secondary, nonthermal stage of inflation due to symmetry restoration after preheating. We review the theory of nonthermal phase transitions and make a prediction about the expansion factor during the secondary inflationary stage. We then present the results of lattice simulations which verify these predictions, and discuss possible implications of our results for the theory of formation of topological defects during nonthermal phase transitions.

  3. Nuclear fuel preheating system

    International Nuclear Information System (INIS)

    Andrea, C.

    1975-01-01

    A nuclear reactor new fuel handling system which conveys new fuel from a fuel preparation room into the reactor containment boundary is described. The handling system is provided with a fuel preheating station which is adaptd to heat the new fuel to reactor refueling temperatures in such a way that the fuel is heated from the top down so that fuel element cladding failure due to thermal expansions is avoided. (U.S.)

  4. Preheating in new inflation

    International Nuclear Information System (INIS)

    Desroche, Mariel; Felder, Gary N.; Kratochvil, Jan M.; Linde, Andrei

    2005-01-01

    During the last ten years a detailed investigation of preheating was performed for chaotic inflation and for hybrid inflation. However, nonperturbative effects during reheating in the new inflation scenario remained practically unexplored. We investigate preheating in new inflation, using a combination of analytical and numerical methods. We find that the decay of the homogeneous component of the inflaton field and the resulting process of spontaneous symmetry breaking in the simplest models of new inflation usually occurs almost instantly: for the new inflation on the GUT scale it takes only about 5 oscillations of the field distribution. The decay of the homogeneous inflaton field is so efficient because of a combined effect of tachyonic preheating and parametric resonance. At that stage, the homogeneous oscillating inflaton field decays into a collection of waves of the inflaton field, with a typical wavelength of the order of the inverse inflaton mass. This stage usually is followed by a long stage of decay of the inflaton field into other particles, which can be described by the perturbative approach to reheating after inflation. The resulting reheating temperature typically is rather low

  5. Preheating with extra dimensions

    International Nuclear Information System (INIS)

    Tsujikawa, S.

    2000-01-01

    We investigate preheating in a higher-dimensional generalized Kaluza-Klein theory with a quadratic inflaton potential V(/φ) = /frac12 m 2 /φ 2 including metric perturbations explicitly. The system we consider is the multi-field model where there exists a dilaton field /σ which corresponds to the scale of compactifications and another scalar field /χ coupled to inflaton with the interaction frac12 g 2 /φ 2 /χ 2 +/g-tilde 2 /φ 3 /χ. In the case of g-tilde=0, we find that the perturbation of dilaton does not undergo parametric amplification while the χ field fluctuation can be enhanced in the usual manner by parametric resonance. In the presence of the /g-tilde 2 /φ 3 /χ coupling, the dilaton fluctuation in sub-Hubble scales is modestly amplified by the growth of metric perturbations for the large coupling g-tilde. In super-Hubble scales, the enhancement of the dilaton fluctuation as well as metric perturbations is weak, taking into account the backreaction effect of created /χ particles. We argue that not only is it possible to predict the ordinary inflationary spectrum in large scales but extra dimensions can be held static during preheating in our scenario. (author)

  6. Preheating curvaton perturbations

    International Nuclear Information System (INIS)

    Bastero-Gil, M.; Di Clemente, V.; King, S.F.

    2005-01-01

    We discuss the potentially important role played by preheating in certain variants of the curvaton mechanism in which isocurvature perturbations of a D-flat (and F-flat) direction become converted to curvature perturbations during reheating. We discover that parametric resonance of the isocurvature components amplifies the superhorizon fluctuations by a significant amount. As an example of these effects we develop a particle physics motivated model which involves hybrid inflation with the waterfall field N being responsible for generating the μ term, the right-handed neutrino mass scale, and the Peccei-Quinn symmetry breaking scale. The role of the curvaton field can be played either by usual Higgs field, or the lightest right-handed sneutrino. Our new results show that it is possible to achieve the correct curvature perturbations for initial values of the curvaton fields of order the weak scale. In this model we show that the prediction for the spectral index of the final curvature perturbation only depends on the mass of the curvaton during inflation, where consistency with current observational data requires the ratio of this mass to the Hubble constant to be 0.3

  7. Gravitational-wave mediated preheating

    Energy Technology Data Exchange (ETDEWEB)

    Alexander, Stephon [Center for Cosmic Origins and Department of Physics and Astronomy, Dartmouth College, Hanover, NH 03755 (United States); Cormack, Sam, E-mail: samuel.c.cormack.gr@dartmouth.edu [Center for Cosmic Origins and Department of Physics and Astronomy, Dartmouth College, Hanover, NH 03755 (United States); Marcianò, Antonino [Center for Field Theory and Particle Physics & Department of Physics, Fudan University, 200433 Shanghai (China); Yunes, Nicolás [Department of Physics, Montana State University, Bozeman, MT 59717 (United States); Kavli Institute for Theoretical Physics, University of California, Santa Barbara, CA 93106 (United States)

    2015-04-09

    We propose a new preheating mechanism through the coupling of the gravitational field to both the inflaton and matter fields, without direct inflaton–matter couplings. The inflaton transfers power to the matter fields through interactions with gravitational waves, which are exponentially enhanced due to an inflation–graviton coupling. One such coupling is the product of the inflaton to the Pontryagin density, as in dynamical Chern–Simons gravity. The energy scales involved are constrained by requiring that preheating happens fast during matter domination.

  8. Efficacy of Prewarming With a Self-Warming Blanket for the Prevention of Unintended Perioperative Hypothermia in Patients Undergoing Hip or Knee Arthroplasty

    DEFF Research Database (Denmark)

    Rosenkilde, Charlotte; Vamosi, Marianne; Lauridsen, Jorgen T.

    2017-01-01

    PURPOSE: Unintended perioperative hypothermia (UPH) is a common and serious complication for patients undergoing anesthesia. The purpose of this study was to identify the incidence of UPH and evaluate the efficacy of a self-warming blanket on the drop in core temperature and risk of UPH in patients...

  9. Hybrid preheat/recirculating steam generator

    International Nuclear Information System (INIS)

    Lilly, G.P.

    1985-01-01

    The patent describes a hybrid preheat/recirculating steam generator for nuclear power plants. The steam generator utilizes recirculated liquid to preheat incoming liquid. In addition, the steam generator incorporates a divider so as to limit the amount of recirculating water mixed with the feedwater. (U.K.)

  10. Breeding blanket for Demo

    International Nuclear Information System (INIS)

    Proust, E.; Giancarli, L.

    1992-01-01

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

  11. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  12. Preheat operating experiences at the FFTF

    International Nuclear Information System (INIS)

    Tucker, W.R.

    1978-01-01

    The rather extensive test program performed on the FFTF preheat control system resulted in successful sodium fill of one secondary heat transport loop on July 2, 1978. The data obtained during testing and the attendant operating experience gained resulted in some design changes and provided the information necessary to fully characterize system performance. Temperature excursions and deviations from preset limits of only a minor nature were encountered during preheat for sodium fill. The addition of the rate alarm feature was beneficial to operation of the preheat system and allowed early detection and correction of impending excursions

  13. Materials for breeding blankets

    International Nuclear Information System (INIS)

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

    1995-09-01

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

  14. Materials for breeding blankets

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  15. Products cooked in preheated versus non-preheated ovens. Baking times, calculated energy consumption, and product quality compared.

    Science.gov (United States)

    Odland, D; Davis, C

    1982-08-01

    Plain muffins, yellow cake, baked custard, apple pie, tuna casserole, frozen tuna casserole, cheese soufflé, and meat loaf were baked in preheated and non-preheated standard gas, continuous-clean gas, standard electric, and self-cleaning electric ovens. Products generally required 5 min. or less extra baking time when cooked in non-preheated rather than in preheated ovens. The variability in baking times often was less between preheated and non-preheated ovens than among oven types. Calculated energy consumption values showed that usually less energy was required to bake products in non-preheated than in preheated ovens; savings averaged about 10 percent. Few significant differences were found in physical measurements or eating quality either between preheated and non-preheated ovens or among oven types. Overall, for the products tested, findings confirmed that preheating the oven is not essential for good product quality and, therefore, is an unnecessary use of energy.

  16. Dynamics of Symmetry Breaking and Tachyonic Preheating

    CERN Document Server

    Felder, G; Greene, P B; Kofman, L A; Linde, Andrei D; Tkachev, Igor I; Felder, Gary; Garcia-Bellido, Juan; Greene, Patrick B.; Kofman, Lev; Linde, Andrei; Tkachev, Igor

    2001-01-01

    We reconsider the old problem of the dynamics of spontaneous symmetry breaking using 3d lattice simulations, and develop a theory of tachyonic preheating, which occurs due to the spinodal instability of the scalar field. Tachyonic preheating is so efficient that symmetry breaking typically completes within a single oscillation of the field distribution as it rolls towards the minimum of its effective potential. As an application of this theory we consider preheating in the hybrid inflation scenario, including SUSY-motivated F-term and D-term inflationary models. We show that preheating in hybrid inflation is typically tachyonic and the stage of oscillations of a homogeneous component of the scalar fields driving inflation ends after a single oscillation. Our results may also be relevant for the theory of the formation of disoriented chiral condensates in heavy ion collisions.

  17. Preheating Mechanism in F-term SUSY Hybrid Inflation

    International Nuclear Information System (INIS)

    Mazumdar, Arindam

    2012-01-01

    Supersymmetric F-term hybrid inflation is one of the most popular models of inflation. Preheating process occurs in this model via two different mechanism. Firstly the standard parametric resonance and secondly, the tachyonic preheating. Generally tachyonic preheating dominates the parametric resonance for this type of models. For different values of the parameters of the theory dominance of tachyonic preheating can vary.

  18. Thermal response of a pin-type fusion reactor blanket during steady and transient reactor operation

    International Nuclear Information System (INIS)

    Grotz, S.; Ghoniem, N.M.

    1986-02-01

    The thermal analysis of the blanket examines both the steady-state and transient reactor operations. The steady-state analysis covers full power and fractional power operation whereas the transient analysis examines the effects of power ramps and blanket preheat. The blanket configuration chosen for this study is a helium cooled solid breeder design. We first discuss the full power, steady-state temperature fields in the first wall, beryllium rods, and breeder rods. Next we examine the effects of fractional power on coolant flow and temperature field distributions. This includes power plateaus of 10%, 20%, 50%, 80%, and 100% of full power. Also examined are the restrictions on the rates of power ramping between plateaus. Finally we discuss the power and time requirements for pre-heating the primary from cold iron conditions up to startup temperature (250 0 C)

  19. Elevated temperature forming method and preheater apparatus

    Science.gov (United States)

    Krajewski, Paul E; Hammar, Richard Harry; Singh, Jugraj; Cedar, Dennis; Friedman, Peter A; Luo, Yingbing

    2013-06-11

    An elevated temperature forming system in which a sheet metal workpiece is provided in a first stage position of a multi-stage pre-heater, is heated to a first stage temperature lower than a desired pre-heat temperature, is moved to a final stage position where it is heated to a desired final stage temperature, is transferred to a forming press, and is formed by the forming press. The preheater includes upper and lower platens that transfer heat into workpieces disposed between the platens. A shim spaces the upper platen from the lower platen by a distance greater than a thickness of the workpieces to be heated by the platens and less than a distance at which the upper platen would require an undesirably high input of energy to effectively heat the workpiece without being pressed into contact with the workpiece.

  20. Gravitational radiation from preheating with many fields

    International Nuclear Information System (INIS)

    Jr, John T. Giblin; Price, Larry R.; Siemens, Xavier

    2010-01-01

    Parametric resonances provide a mechanism by which particles can be created just after inflation. Thus far, attention has focused on a single or many inflaton fields coupled to a single scalar field. However, generically we expect the inflaton to couple to many other relativistic degrees of freedom present in the early universe. Using simulations in an expanding Friedmann-Lemaître-Robertson-Walker spacetime, in this paper we show how preheating is affected by the addition of multiple fields coupled to the inflaton. We focus our attention on gravitational wave production — an important potential observational signature of the preheating stage. We find that preheating and its gravitational wave signature is robust to the coupling of the inflaton to more matter fields

  1. Gravitational radiation from preheating with many fields

    Energy Technology Data Exchange (ETDEWEB)

    Jr, John T. Giblin [Department of Physics, Kenyon College, 201 North College Road, Gambier, OH 43022 (United States); Price, Larry R.; Siemens, Xavier, E-mail: giblinj@kenyon.edu, E-mail: larry@gravity.phys.uwm.edu, E-mail: siemens@gravity.phys.uwm.edu [Center for Gravitation and Cosmology, Department of Physics, University of Wisconsin — Milwaukee, P.O. Box 413, Milwaukee, WI 53201 (United States)

    2010-08-01

    Parametric resonances provide a mechanism by which particles can be created just after inflation. Thus far, attention has focused on a single or many inflaton fields coupled to a single scalar field. However, generically we expect the inflaton to couple to many other relativistic degrees of freedom present in the early universe. Using simulations in an expanding Friedmann-Lemaître-Robertson-Walker spacetime, in this paper we show how preheating is affected by the addition of multiple fields coupled to the inflaton. We focus our attention on gravitational wave production — an important potential observational signature of the preheating stage. We find that preheating and its gravitational wave signature is robust to the coupling of the inflaton to more matter fields.

  2. Mirror reactor blankets

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  3. Effects of dissipation and fluctuation in preheating

    International Nuclear Information System (INIS)

    Vartuli, Rodrigo; Ramos, Rudnei de O.

    2006-01-01

    In this paper, we study the effects of dissipation and fluctuation in preheating after inflation. The effective equation of motion for a scalar field χ interacting with lighter fields is derived using the field theoretical method of closed time path due to Schwinger, winch is suitable to study nonequilibrium and time dependent process. In this derivation the emergent equation is intrinsically dissipative and stochastic in nature. The resulting dynamics is then studied both analytically and numerically. The results obtained are then discussed for then relevance for the reheating epoch right after an inflationary phase(preheating) for the case of the evolution of the scalar field χ and its decay into fermion. (author)

  4. Fusion fuel blanket technology

    International Nuclear Information System (INIS)

    Hastings, I.J.; Gierszewski, P.

    1987-05-01

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

  5. Blanket testing in NET

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  6. Simple air collectors for preheating fresh air

    NARCIS (Netherlands)

    Hensen, J.L.M.; Wit, de M.H.; Ouden, den C.

    1984-01-01

    In dwellings with mechanical ventilation systems the fresh air can easily be preheated by means of simple solar air systems. These can be an integral part of the building facade or roof and the costs are expected to be low. By means of computer experiments a large number of systems were evaluated.

  7. Limitations on blanket performance

    International Nuclear Information System (INIS)

    Malang, S.

    1999-01-01

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

  8. 7 CFR 58.919 - Pre-heat, pasteurization.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 3 2010-01-01 2010-01-01 false Pre-heat, pasteurization. 58.919 Section 58.919... Procedures § 58.919 Pre-heat, pasteurization. When pasteurization is intended or required by either the vat... requirements outlined in § 58.128. Pre-heat temperatures prior to ultra pasteurization will be those that have...

  9. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  10. Tritium breeding blanket

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  11. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

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

  12. Blankets for thermonuclear device

    International Nuclear Information System (INIS)

    Maki, Koichi; Fukumoto, Hideshi.

    1986-01-01

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

  13. ITER blanket designs

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  14. Metric preheating and limitations of linearized gravity

    International Nuclear Information System (INIS)

    Bassett, Bruce A.; Tamburini, Fabrizio; Kaiser, David I.; Maartens, Roy

    1999-01-01

    During the preheating era after inflation, resonant amplification of quantum field fluctuations takes place. Recently it has become clear that this must be accompanied by resonant amplification of scalar metric fluctuations, since the two are united by Einstein's equations. Furthermore, this 'metric preheating' enhances particle production, and leads to gravitational rescattering effects even at linear order. In multi-field models with strong preheating (q>>1), metric perturbations are driven non-linear, with the strongest amplification typically on super-Hubble scales (k→0). This amplification is causal, being due to the super-Hubble coherence of the inflaton condensate, and is accompanied by resonant growth of entropy perturbations. The amplification invalidates the use of the linearized Einstein field equations, irrespective of the amount of fine-tuning of the initial conditions. This has serious implications on all scales - from large-angle cosmic microwave background (CMB) anisotropies to primordial black holes. We investigate the (q,k) parameter space in a two-field model, and introduce the time to non-linearity, t nl , as the timescale for the breakdown of the linearized Einstein equations. t nl is a robust indicator of resonance behavior, showing the fine structure in q and k that one expects from a quasi-Floquet system, and we argue that t nl is a suitable generalization of the static Floquet index in an expanding universe. Backreaction effects are expected to shut down the linear resonances, but cannot remove the existing amplification, which threatens the viability of strong preheating when confronted with the CMB. Mode-mode coupling and turbulence tend to re-establish scale invariance, but this process is limited by causality and for small k the primordial scale invariance of the spectrum may be destroyed. We discuss ways to escape the above conclusions, including secondary phases of inflation and preheating solely to fermions. The exclusion principle

  15. Are black holes overproduced during preheating?

    International Nuclear Information System (INIS)

    Suyama, Teruaki; Tanaka, Takahiro; Bassett, Bruce; Kudoh, Hideaki

    2005-01-01

    We provide a simple but robust argument that primordial black hole production generically does not exceed astrophysical bounds during the resonant preheating phase after inflation. This conclusion is supported by fully nonlinear lattice simulations of various models in two and three dimensions which include rescattering but neglect metric perturbations. We examine the degree to which preheating amplifies density perturbations at the Hubble scale and show that, at the end of the parametric resonance, power spectra are universal, with no memory of the power spectrum at the end of inflation. In addition, we show how the probability distribution of density perturbations changes from exponential on very small scales to Gaussian when smoothed over the Hubble scale - the crucial length for studies of primordial black hole formation - hence justifying the standard assumption of Gaussianity

  16. Influence of preheating on grindability of coal

    Science.gov (United States)

    Lytle, J.; Choi, N.; Prisbrey, K.

    1992-01-01

    Enormous quantities of coal must be ground as feed to power generation facilities. The energy cost of grinding is significant at 5 to 15 kWh/ton. If grindability could be increased by preheating the coal with waste heat, energy costs could be reduced. The objective of this work was to determine how grindability was affected by preheating. The method was to use population balance grinding models to interpret results of grinding coal before and after a heat treatment. Simulation of locked cycle tests gave a 40% increase in grindability. Approximately 40% grinding energy saving can be expected. By using waste heat for coal treatment, the targeted energy savings would be maintained. ?? 1992.

  17. Nuclear reactor insulation and preheat system

    International Nuclear Information System (INIS)

    Wampole, N.C.

    1978-01-01

    An insulation and preheat system is disclosed for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the ocmpartment. An external surface of the compartment of enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair

  18. Preheating of tap water with solar collectors

    Energy Technology Data Exchange (ETDEWEB)

    Granum, H; Raaen, H

    1992-05-05

    In 1991 SINTEF Architecture and Building Technology won the second prize in 'The Nordic Competition for Low Energy Buildings' with a project proposal named 'LOWe'. The paper gives a description of the energy-saving features of this project, particularly the use of a solar collector for preheating of tap water. Compared with the economic profitability of other saving efforts in the project, such as good thermal insulation and efficient heat recovering system, the system for solar preheating of tap water does not seem very attractive for the time being. Loose estimates indicate a cost of close of NOK 1.00 per kWh for the produced energy in the solar collector, while the present price for electricity in Norway is about NOK 0.50 per kWh. Compared with a heat pump solution however the energy cost is not unreasonable.

  19. Pre-heating mitigates composite degradation.

    Science.gov (United States)

    Silva, Jessika Calixto da; Rogério Vieira, Reges; Rege, Inara Carneiro Costa; Cruz, Carlos Alberto dos Santos; Vaz, Luís Geraldo; Estrela, Carlos; Castro, Fabrício Luscino Alves de

    2015-01-01

    Dental composites cured at high temperatures show improved properties and higher degrees of conversion; however, there is no information available about the effect of pre-heating on material degradation. Objectives This study evaluated the effect of pre-heating on the degradation of composites, based on the analysis of radiopacity and silver penetration using scanning electron microscopy/energy-dispersive X-ray spectroscopy (SEM/EDS). Material and Methods Thirty specimens were fabricated using a metallic matrix (2x8 mm) and the composites Durafill VS (Heraeus Kulzer), Z-250 (3M/ESPE), and Z-350 (3M/ESPE), cured at 25°C (no pre-heating) or 60°C (pre-heating). Specimens were stored sequentially in the following solutions: 1) water for 7 days (60°C), plus 0.1 N sodium hydroxide (NaOH) for 14 days (60°C); 2) 50% silver nitrate (AgNO3) for 10 days (60°C). Specimens were radiographed at baseline and after each storage time, and the images were evaluated in gray scale. After the storage protocol, samples were analyzed using SEM/EDS to check the depth of silver penetration. Radiopacity and silver penetration data were analyzed using ANOVA and Tukey's tests (α=5%). Results Radiopacity levels were as follows: Durafill VSZ-350>Z-250 (pheated specimens presented higher radiopacity values than non-pre-heated specimens (pheated specimens (pheating at 60°C mitigated the degradation of composites based on analysis of radiopacity and silver penetration depth.

  20. Spectroscopic Measurements of Target Preheating on OMEGA

    International Nuclear Information System (INIS)

    Elton, R.C.; Griem, H.R.; Iglesias, E.J.

    2000-01-01

    The preheating of laser-heated microballoon targets has been measured by time-resolved x-ray and extreme ultraviolet (euv) spectroscopy on the 30 kJ, 351 nm, 60-beam laser-fusion system at the University of Rochester Laboratory for Laser Energetics. Thin coatings of aluminum overcoated with magnesium served as indicators. both the sequence of the x-ray line emission and the intensity of euv radiation were used to determine a preheating peaking at ∼ 10 ns prior to onset of the main laser pulse, with a power density ≅1% of the main pulse. The measurements are supported by numerical modeling. Further information is provided by absorption spectra from the aluminum coating, backlighted by continuum from the heated surface. The exact source of the preheating energy remains unknown at present, but most likely arrives from early laser leakage through the system. The present target diagnostic is particularly useful when all beams cannot be monitored directly at all laser wavelengths

  1. Novel blanket design for ICTR's

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  2. Sintering uranium oxide using a preheating step

    International Nuclear Information System (INIS)

    Jensen, N.J.; Nivas, Y.; Packard, D.R.

    1977-01-01

    Compacted pellets of uranium oxide or uranium oxide with one or more additives are heated in a kiln in a process having a preheating step, a sintering step, a reduction step, and a cooling step in a controlled atmosphere. The process is practiced to give a range of temperature and atmosphere conditions for obtaining optimum fluoride removal from the compacted pellets along with optimum sintering in a single process. The preheating step of this process is conducted in a temperature range of about 600 0 to about 900 0 C and the pellets are held for at least twenty min, and preferably about 60 min, in an atmosphere having a composition in the range of about 10 to about 75 vol % hydrogen with the balance being carbon dioxide. The sintering step is conducted at a temperature in the range of about 900 0 C to 1500 0 C in the presence of an atmosphere having a composition in the range of about 0.5 to about 90 vol % hydrogen with the balance being carbon dioxide. The reduction step reduces the oxygen to metal ratio of the pellets to a range of about 1.98 to 2.10:1 and this is accomplished by gradually cooling the pellets for about 30 to about 120 min from the temperature of the sintering step to about 1100 0 C in an atmosphere of about 10 to 90 vol % hydrogen with the balance being carbon dioxide. Thereafter the pellets are cooled to about 100 0 C under a protective atmosphere, and in one preferred practice the same atmosphere used in the reduction step is used in the cooling step. The preheating, sintering and reduction steps may also be conducted with their respective atmospheres having an initial additional component of water vapor and the water vapor can comprise up to about 20 vol %

  3. From (p)reheating to nucleosynthesis

    International Nuclear Information System (INIS)

    Jedamzik, Karsten

    2002-01-01

    This paper gives a brief qualitative description of the possible evolution of the early universe between the end of an inflationary epoch and the end of big-bang nucleosynthesis. After a general introduction, establishing the minimum requirements cosmologists impose on this cosmic evolutionary phase, namely, successful baryogenesis, the production of cosmic dark matter and successful light-element nucleosynthesis, a more detailed discussion on some recent developments follows. This latter includes the physics of preheating, the putative production of (alternative) dark matter and the current status of big bang nucleosynthesis

  4. When can preheating affect the CMB?

    Science.gov (United States)

    Tsujikawa, Shinji; Bassett, Bruce A.

    2002-05-01

    We discuss the principles governing the selection of inflationary models for which preheating can affect the CMB. This is a (fairly small) subset of those models which have nonnegligible entropy/isocurvature perturbations on large scales during inflation. We study new models which belong to this class-two-field inflation with negative nonminimal coupling and hybrid/double/supernatural inflation models where the tachyonic growth of entropy perturbations can lead to the variation of the curvature perturbation, /R, on super-Hubble scales. Finally, we present evidence against recent claims for the variation of /R in the absence of substantial super-Hubble entropy perturbations.

  5. Magnetoconvection in HCLL blankets

    International Nuclear Information System (INIS)

    Mistrangelo, C.; Buehler, L.

    2014-01-01

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

  6. Preheating the universe in hybrid inflation

    CERN Document Server

    García-Bellido, J

    1998-01-01

    One of the fundamental problems of modern cosmology is to explain the origin of all the matter and radiation in the Universe today. The inflationary model predicts that the oscillations of the scalar field at the end of inflation will convert the coherent energy density of the inflaton into a large number of particles, responsible for the present entropy of the Universe. The transition from the inflationary era to the radiation era was originally called reheating, and we now understand that it may consist of three different stages: preheating, in which the homogeneous inflaton field decays coherently into bosonic waves (scalars and/or vectors) with large occupation numbers; backreaction and rescattering, in which different energy bands get mixed; and finally decoherence and thermalization, in which those waves break up into particles that thermalize and acquire a black body spectrum at a certain temperature. These three stages are non-perturbative, non-linear and out of equilibrium, and we are just beginning ...

  7. Blanket for thermonuclear device

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  8. Gravitational wave production from preheating: parameter dependence

    Energy Technology Data Exchange (ETDEWEB)

    Figueroa, Daniel G. [Theory Division, CERN, 1211 Geneva (Switzerland); Torrentí, Francisco, E-mail: daniel.figueroa@cern.ch, E-mail: f.torrenti@csic.es [Instituto de Física Teórica IFT-UAM/CSIC, Universidad Autónoma de Madrid, Cantoblanco 28049 Madrid, Spain. (Spain)

    2017-10-01

    Parametric resonance is among the most efficient phenomena generating gravitational waves (GWs) in the early Universe. The dynamics of parametric resonance, and hence of the GWs, depend exclusively on the resonance parameter q . The latter is determined by the properties of each scenario: the initial amplitude and potential curvature of the oscillating field, and its coupling to other species. Previous works have only studied the GW production for fixed value(s) of q . We present an analytical derivation of the GW amplitude dependence on q , valid for any scenario, which we confront against numerical results. By running lattice simulations in an expanding grid, we study for a wide range of q values, the production of GWs in post-inflationary preheating scenarios driven by parametric resonance. We present simple fits for the final amplitude and position of the local maxima in the GW spectrum. Our parametrization allows to predict the location and amplitude of the GW background today, for an arbitrary q . The GW signal can be rather large, as h {sup 2Ω}{sub GW}( f {sub p} ) ∼< 10{sup −11}, but it is always peaked at high frequencies f {sub p} ∼> 10{sup 7} Hz. We also discuss the case of spectator-field scenarios, where the oscillatory field can be e.g. a curvaton, or the Standard Model Higgs.

  9. A New Laser Preheat Protocol For Maglif

    Science.gov (United States)

    Weis, M. R.; Harvey-Thompson, A. J.; Geissel, M.; Jennings, C. A.; Peterson, K. J.; Glinsky, M. E.; Awe, T. J.; Bliss, D. E.; Gomez, M. R.; Harding, E. C.; Hansen, S. B.; Kimmel, M. W.; Knapp, P. F.; Lewis, S. M.; Porter, J. L.; Rochau, G. A.; Schollmeier, M.; Schwarz, J.; Shores, J. E.; Slutz, S. A.; Sinars, D. B.; Smith, I. C.; Speas, C. S.

    2017-10-01

    Previous Magnetized Liner Inertial Fusion experiments at Sandia National Labs have preheated the fuel with the unsmoothed 2 ω Z-Beamlet Laser. A new low intensity laser configuration, using phase plate smoothing and a low-power pulse shape, improved laser propagation and reduced stimulated Brillouin scattering in offline laser experiments. This allows for more efficient use of laser energy and better spot reproducibility. The new laser protocol is estimated to couple at least 650 J to the fuel, sufficient to produce comparable neutron yields with the previous unsmoothed configuration. Mid-Z dopants were also fielded on the underside of the window. Observation of these dopants provided evidence of window material mixing into the fuel with both the unsmoothed and smoothed beam, consistent with MHD simulation predictions. Sandia National Laboratories is a multi-mission laboratory managed and operated by NTESS, LLC, a wholly owned subsidiary of Honeywell International, Inc., for the U.S. DOE's NNSA under contract DE-NA0003525.

  10. Prediction of flame formation in highly preheated air combustion

    International Nuclear Information System (INIS)

    Yang, Jang Sik; Choi, Gyung Min; Kim, Duck Jool; Katsuki, Masashi

    2008-01-01

    Fundamental information about the ignition position and shape of a flame in highly preheated air combustion was obtained, and the suitability of the suggested reduced kinetic mechanism that reflects the characteristics of the highly preheated air combustion was demonstrated. Flame lift height and flame length with variations of premixed air temperature and oxygen concentration were measured by CH chemiluminescence intensity, and were computed with a reduced kinetic mechanism. Flame attached near a fuel nozzle started to lift when preheated air temperature became close to auto-ignition temperature and/or oxygen concentration reduced. The flame lift height increased but the flame length decreased with decreasing preheated air temperature and flame length reversed after a minimum value. Calculated results showed good agreement with those of experiment within tolerable error. Flame shape shifted from diffusion flame shape to partial premixed flame shape with increasing lift height and this tendency was also observed in the computation results

  11. Prediction of flame formation in highly preheated air combustion

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Jang Sik; Choi, Gyung Min; Kim, Duck Jool [Pusan National University, Busan (Korea, Republic of); Katsuki, Masashi [Osaka University, Osaka (Japan)

    2008-11-15

    Fundamental information about the ignition position and shape of a flame in highly preheated air combustion was obtained, and the suitability of the suggested reduced kinetic mechanism that reflects the characteristics of the highly preheated air combustion was demonstrated. Flame lift height and flame length with variations of premixed air temperature and oxygen concentration were measured by CH chemiluminescence intensity, and were computed with a reduced kinetic mechanism. Flame attached near a fuel nozzle started to lift when preheated air temperature became close to auto-ignition temperature and/or oxygen concentration reduced. The flame lift height increased but the flame length decreased with decreasing preheated air temperature and flame length reversed after a minimum value. Calculated results showed good agreement with those of experiment within tolerable error. Flame shape shifted from diffusion flame shape to partial premixed flame shape with increasing lift height and this tendency was also observed in the computation results

  12. Numerical Simulation of Anisotropic Preheating Ablative Rayleigh–Taylor Instability

    International Nuclear Information System (INIS)

    Li-Feng, Wang; Wen-Hua, Ye; Ying-Jun, Li

    2010-01-01

    The linear growth rate of the anisotropic preheating ablative Rayleigh–Taylor instability (ARTI) is studied by numerical simulations. The preheating model κ(T) = κ SH [1 + f(T)] is applied, where f(T) is the preheating function interpreting the preheating tongue effect in the cold plasma ahead of the ablative front. An arbitrary coefficient D is introduced in the energy equation to study the influence of transverse thermal conductivity on the growth of the ARTI. We find that enhancing diffusion in a plane transverse to the mean longitudinal flow can strongly reduce the growth of the instability. Numerical simulations exhibit a significant stabilization of the ablation front by improving the transverse thermal conduction. Our results are in general agreement with the theory analysis and numerical simulations by Masse [Phys. Rev. Lett. 98 (2007) 245001]. (physics of gases, plasmas, and electric discharges)

  13. Numerical simulation of anisotropic preheating ablative Rayleigh-Taylor instability

    International Nuclear Information System (INIS)

    Wang Lifeng; Ye Wenhua; Li Yingjun

    2010-01-01

    The linear growth rate of the anisotropic preheating ablative Rayleigh-Taylor instability (ARTI) is studied by numerical simulations. The preheating model κ(T)=κ SH [1+f(T)] is applied, where f(T) is the preheating function interpreting the preheating tongue effect in the cold plasma ahead of the ablative front. An arbitrary coefficient D is introduced in the energy equation to study the influence of transverse thermal conductivity on the growth of the ARTI. We find that enhancing diffusion in a plane transverse to the mean longitudinal flow can strongly reduce the growth of the instability. Numerical simulations exhibit a significant stabilization of the ablation front by improving the transverse thermal conduction. Our results are in general agreement with the theory analysis and numerical simulations by Masse. (authors)

  14. Effect of inflation on parametric resonance during preheating

    International Nuclear Information System (INIS)

    Hirai, Shiro

    2002-01-01

    The effect of inflation on parametric resonance during preheating is investigated. The behaviour of the preheating scalar field during inflation is investigated and is found to become squeezed in cases ranging from small-scale cases to large-scale cases. However, the positive-frequency solution is usually adopted in the initial condition of the scalar field at preheating. Although large squeezing occurs during inflation, the difference in the comoving occupation number of particles n k between two initial conditions is shown to be not so large. Rather, the ratio n k varies from 0.2 to 5.0, depending on k. In order to clarify this situation, we introduce the squeeze formulation. The squeeze parameters r and φ are calculated not only in preheating, but also in inflation. Since the squeeze parameters are calculated from inflation to preheating, we can clarify the behaviour of the parametric resonance. In preheating, the behaviour of r is shown to remain relatively unchanged with respect to k; however, the squeeze angle φ displays different behaviour for large-scale cases and small-scale cases

  15. Behavior of the turbine - regenerating preheaters functional assembly

    International Nuclear Information System (INIS)

    Bigu, Melania; Nita, Iulian Pavel; Tenescu, Mircea

    2004-01-01

    In the classical calculation of pressure distribution in the turbine-regenerating heaters' assembly a uniform distribution of feedwater enthalpy rise at each regenerating preheating step is usually assumed. This is accurately enough as a basis of designing of the preheating installation operating at rated power regime. But at partial regimes this is not totally valid since the preheaters are already shaped and the quasi-equal distribution does not satisfy the equation system describing the heat transfer correlations in these installations. A more detailed analysis shows that pressure in the feeding line preheaters and the bleeding steam flow rates at the turbine outlets are described physically by solving simultaneously the equations of hydrodynamic flow through the turbine and the equations of the heat transfer in the preheaters of the feedwater preheating line. This work approaches this more accurate solving method at least from a theoretical standing point; two cases are illustrated in the annexes of the work: a case of a secondary circuit with a single regenerating inlet and a case with two regenerating inlets. A classical - Panzer method of transformation of a many regenerative stages scheme may lead to one or another of the above cases. (authors)

  16. Blanket comparison and selection study. Volume II

    International Nuclear Information System (INIS)

    1983-10-01

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

  17. Susceptibility of CANDU steam generator preheater to cavitation erosion

    International Nuclear Information System (INIS)

    Laroche, S.L.; Sun, L.; Pietralik, J.M.

    2012-01-01

    In 2009, Darlington Steam Generator (SG) tube inspections revealed some tubes had degraded in the preheater. The tube degradation occurred at the clearance gap between the tube and the preheater baffle and reached up to 50% through-wall depth at the baffles in the middle portion of the preheater. The general pattern of the damage and the elemental composition analysis suggested that the degradation was the result of a hydrodynamic process, such as cavitation erosion. Cavitation erosion occurs when vapour bubbles exist or form in the flowing liquid and then these bubbles collapse violently in the vicinity of the wall. These bubbles collapse when steam bubbles contact water that is sufficiently subcooled, below the saturation temperature. In the gap between the tube and the preheater baffle, low flow will exist due to the pressure difference across the baffle plate. In addition, heat transfer occurs from the primary-side fluid to the secondary-side fluid within this clearance gap that is driven by the primary-to-secondary temperature difference. Factors, such as the tube position in the baffle hole and fouling, influence the local conditions and can cause subcooled boiling that result in cavitation. This paper presents a study of flow and heat transfer phenomena to determine the factors contributing to cavitation erosion in SG preheaters. The analysis used the THIRST1 code for a 3-dimensional thermalhydraulic simulation of the steam generators and the ANSYS FLUENT®2 code for detailed calculations of flow and heat transfer in the clearance gaps. This study identifies that tubes in the preheater region are susceptible to cavitation erosion and indicates that this area should be part of the station inspection program because, regardless of preheater design, some tubes may experience the thermalhydraulic conditions and undergo degradations similar to those observed for the tubes in Darlington SGs. (author)

  18. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

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

  19. An analysis of electron beam welds in a dual coolant liquid metal breeder blanket

    International Nuclear Information System (INIS)

    Cizelj, L.; Riesch-Oppermann, H.; Kernforschungszentrum Karlsruhe GmbH

    1994-10-01

    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.) [de

  20. Fusion blanket design and optimization techniques

    International Nuclear Information System (INIS)

    Gohar, Y.

    2005-01-01

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

  1. The effect of preheating on the IRSL signal from feldspar

    DEFF Research Database (Denmark)

    Murray, A.S.; Buylaert, J.P.; Thomsen, Kristina Jørkov

    2009-01-01

    between the loss of blue IRSL and TL signals with preheating, and the effect of prior IRSL on the TL signal. Using IRSL measured at 50 °C and a SAR protocol, we then examine the dependence on preheat temperature of equivalent dose (De), laboratory fading rate (g), and the resulting luminescence age, from...... is consistent with a kinetic analysis of sensitivity-corrected IRSL data. The corollary to our observations is that shallow (unstable) traps do not give rise to a significant IRSL signal....

  2. Effects of Preheat on Weldments of NICOP Steel.

    Science.gov (United States)

    1983-09-01

    percent Nital solution (nitric acid (HNO3) ,* and ethanol (C2HsOH) which revealed the weld area, heat affected zone and base metal. A section 25.2mm (1 inch...electrolyte, consisting of 10% per- cloric acid (HC104 ) and 90% methanal (CH30H) was maintained at a temperature of -450C (-49 0 F). The Polipower was set...Preheated Weidment. N on Non-Preheated Weidment. Figre3. Loaton o McrhadnssTrvese I17 ~.4. .9 G° s s E 43 C 0 CL 44’ 00 Hda *SBUPJQH Figure 4. Comparison

  3. Feed water pre-heater with two steam spaces

    International Nuclear Information System (INIS)

    Tratz, H.; Kelp, F.; Netsch, E.

    1976-01-01

    A feed water pre-heater for the two stage heating of feed water by condensing steam, having a low installed height is described, which can be installed in the steam ducts of turbines of large output, as in LWRs in nuclear power stations. The inner steam space is closed on one side by the water vessel, while the tubes of the inner steam space go straight from the water vessel, and the tubes of the outer steam space are bent into a U shape and open out into the water vessel. The two-stage preheater is thus surrounded by feedwater in two ways. (UWI) [de

  4. Gravity waves from tachyonic preheating after hybrid inflation

    Energy Technology Data Exchange (ETDEWEB)

    Dufaux, Jean-Francois [Instituto de Fisica Teorica UAM/CSIC, Universidad Autonoma de Madrid, Cantoblanco, 28049 Madrid (Spain); Felder, Gary [Department of Physics, Clark Science Center, Smith College, Northampton, MA 01063 (United States); Kofman, Lev [CITA, University of Toronto, 60 St. George Street, Toronto, ON M5S 3H8 (Canada); Navros, Olga, E-mail: jeff.dufaux@uam.es, E-mail: gfelder@email.smith.edu, E-mail: kofman@cita.utoronto.ca, E-mail: navros@email.unc.edu [Department of Mathematics, University of North Carolina Chapel Hill, CB3250 Philips Hall, Chapel Hill, NC 27599 (United States)

    2009-03-15

    We study the stochastic background of gravitational waves produced from preheating in hybrid inflation models. We investigate different dynamical regimes of preheating in these models and we compute the resulting gravity wave spectra using analytical estimates and numerical simulations. We discuss the dependence of the gravity wave frequencies and amplitudes on the various potential parameters. We find that large regions of the parameter space leads to gravity waves that may be observable in upcoming interferometric experiments, including Advanced LIGO, but this generally requires very small coupling constants.

  5. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

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

    1986-02-01

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

  6. Fusion blanket inherent safety assessment

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  7. The blanket interface to TSTA

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Werner, R.W.

    1978-01-01

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

  9. Blanket safety by GEMSAFE methodology

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Saito, Masaki

    2001-01-01

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

  10. Method for pre-heating lmfbr type reactors

    International Nuclear Information System (INIS)

    Yokozawa, Atsushi; Kataoka, Hajime.

    1978-01-01

    Purpose: To enable pre-heating for the inside of the reactor container and the inside of the coolant recycling system with no additional facilities. Method: The coolant recycling system is composed of a heat exchanger, a mechanical pump, a check valve, a flow meter or the like and it is connected in series by way of a pipe line to a reactor container. The mechanical pump is used as a gas recycling device upon pre-heating and it is designed so that a blower such as a fan can be replaced for the impeller of the pump. The inside of the reactor container and the inside of the coolant recycling system is at first filled with an inert gas such as for use with cover gas. Then, nuclear fuels are loaded to attain criticality. Simultaneously, the blower is started and the control rods are operated while cooling the nuclear fuel with the inert gas thus to obtain heat required for pre-heating the pipe line or the like from the nuclear fuels. After the completion of the pre-heating, the liquid metal is charged. (Ikeda, J.)

  11. Constraints on variations in inflaton decay rate from modulated preheating

    Energy Technology Data Exchange (ETDEWEB)

    Mazumdar, Arindam [Theory Division, Saha Institute of Nuclear Physics, 1/AF Bidhannagar, Kolkata-64 (India); Modak, Kamakshya Prasad, E-mail: arindam.mazumdar@saha.ac.in, E-mail: kamakshya.modak@saha.ac.in [Astroparticle Physics and Cosmology Division, Saha Institute of Nuclear Physics, 1/AF Bidhannagar, Kolkata-64 (India)

    2016-06-01

    Modulated (p)reheating is thought to be an alternative mechanism for producing super-horizon curvature perturbations in CMB. But large non-gaussianity and iso-curvature perturbations produced by this mechanism rule out its acceptability as the sole process responsible for generating CMB perturbations. We explore the situation where CMB perturbations are mostly generated by usual quantum fluctuations of inflaton during inflation, but a modulated coupling constant between inflaton and a secondary scalar affects the preheating process and produces some extra curvature perturbations. If the modulating scalar field is considered to be a dark matter candidate, coupling constant between the fields has to be unnaturally fine tuned in order to keep the local-form non-gaussianity and the amplitude of iso-curvature perturbations within observational limit; otherwise parameters of the models have to be tightly constrained. Those constraints imply that the curvature perturbations generated by modulated preheating should be less than 15% of the total observed CMB perturbations. On the other hand if the modulating scalar field is not a dark matter candidate, parameters of the models could not be constrained, but the constraints on the maximum amount of the curvature perturbations coming from modulated preheating remain valid.

  12. Constraints on variations in inflaton decay rate from modulated preheating

    International Nuclear Information System (INIS)

    Mazumdar, Arindam; Modak, Kamakshya Prasad

    2016-01-01

    Modulated (p)reheating is thought to be an alternative mechanism for producing super-horizon curvature perturbations in CMB. But large non-gaussianity and iso-curvature perturbations produced by this mechanism rule out its acceptability as the sole process responsible for generating CMB perturbations. We explore the situation where CMB perturbations are mostly generated by usual quantum fluctuations of inflaton during inflation, but a modulated coupling constant between inflaton and a secondary scalar affects the preheating process and produces some extra curvature perturbations. If the modulating scalar field is considered to be a dark matter candidate, coupling constant between the fields has to be unnaturally fine tuned in order to keep the local-form non-gaussianity and the amplitude of iso-curvature perturbations within observational limit; otherwise parameters of the models have to be tightly constrained. Those constraints imply that the curvature perturbations generated by modulated preheating should be less than 15% of the total observed CMB perturbations. On the other hand if the modulating scalar field is not a dark matter candidate, parameters of the models could not be constrained, but the constraints on the maximum amount of the curvature perturbations coming from modulated preheating remain valid.

  13. Gauge-preheating and the end of axion inflation

    Energy Technology Data Exchange (ETDEWEB)

    Adshead, Peter; Sfakianakis, Evangelos I. [Department of Physics, University of Illinois at Urbana-Champaign, 1110 West Green Street, Urbana, Illinois 61801 (United States); Giblin, John T. Jr.; Scully, Timothy R., E-mail: adshead@illinois.edu, E-mail: giblinj@kenyon.edu, E-mail: tscully2@illinois.edu, E-mail: esfaki@illinois.edu [Department of Physics, Kenyon College, 201 North College Rd, Gambier, Ohio 43022 (United States)

    2015-12-01

    We study the onset of the reheating epoch at the end of axion-driven inflation where the axion is coupled to an Abelian, U(1), gauge field via a Chern-Simons interaction term. We focus primarily on m{sup 2φ2} inflation and explore the possibility that preheating can occur for a range of coupling values consistent with recent observations and bounds on the overproduction of primordial black holes. We find that for a wide range of parameters preheating is efficient. In certain cases the inflaton transfers all of its energy to the gauge fields within a few oscillations. In most cases, we find that the gauge fields on sub-horizon scales end preheating in an unpolarized state due to the existence of strong rescattering between the inflaton and gauge-field modes. We also present a preliminary study of an axion monodromy model coupled to U(1) gauge fields, seeing a similarly efficient preheating behavior as well as indications that the coupling strength has an effect on the creation of oscillons.

  14. Gauge-preheating and the end of axion inflation

    International Nuclear Information System (INIS)

    Adshead, Peter; Sfakianakis, Evangelos I.; Giblin, John T. Jr.; Scully, Timothy R.

    2015-01-01

    We study the onset of the reheating epoch at the end of axion-driven inflation where the axion is coupled to an Abelian, U(1), gauge field via a Chern-Simons interaction term. We focus primarily on m 2φ2 inflation and explore the possibility that preheating can occur for a range of coupling values consistent with recent observations and bounds on the overproduction of primordial black holes. We find that for a wide range of parameters preheating is efficient. In certain cases the inflaton transfers all of its energy to the gauge fields within a few oscillations. In most cases, we find that the gauge fields on sub-horizon scales end preheating in an unpolarized state due to the existence of strong rescattering between the inflaton and gauge-field modes. We also present a preliminary study of an axion monodromy model coupled to U(1) gauge fields, seeing a similarly efficient preheating behavior as well as indications that the coupling strength has an effect on the creation of oscillons

  15. Theory and numerics of gravitational waves from preheating after inflation

    International Nuclear Information System (INIS)

    Dufaux, Jean-Francois; Kofman, Lev; Bergman, Amanda; Felder, Gary; Uzan, Jean-Philippe

    2007-01-01

    Preheating after inflation involves large, time-dependent field inhomogeneities, which act as a classical source of gravitational radiation. The resulting spectrum might be probed by direct detection experiments if inflation occurs at a low enough energy scale. In this paper, we develop a theory and algorithm to calculate, analytically and numerically, the spectrum of energy density in gravitational waves produced from an inhomogeneous background of stochastic scalar fields in an expanding universe. We derive some generic analytical results for the emission of gravity waves by stochastic media of random fields, which can test the validity/accuracy of numerical calculations. We contrast our method with other numerical methods in the literature, and then we apply it to preheating after chaotic inflation. In this case, we are able to check analytically our numerical results, which differ significantly from previous works. We discuss how the gravity-wave spectrum builds up with time and find that the amplitude and the frequency of its peak depend in a relatively simple way on the characteristic spatial scale amplified during preheating. We then estimate the peak frequency and amplitude of the spectrum produced in two models of preheating after hybrid inflation, which for some parameters may be relevant for gravity-wave interferometric experiments

  16. Development of blanket remote maintenance system

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  17. Development of blanket remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

  18. Concepts for fusion fuel production blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1986-06-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  20. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

    Rocaboy, Alain.

    1982-06-01

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

  1. Disruption problematics in segmented blanket concepts

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  2. Fusion reactor blanket-main design aspects

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  3. On the generation of a non-gaussian curvature perturbation during preheating

    Energy Technology Data Exchange (ETDEWEB)

    Kohri, Kazunori; Lyth, David H. [Department of Physics, Lancaster University, Lancaster LA1 4YB (United Kingdom); Valenzuela-Toledo, Cesar A., E-mail: k.kohri@lancaster.ac.uk, E-mail: d.lyth@lancaster.ac.uk, E-mail: cavalto@ciencias.uis.edu.co [Escuela de Física, Universidad Industrial de Santander, Ciudad Universitaria, Bucaramanga (Colombia)

    2010-02-01

    The perturbation of a light field might affect preheating and hence generate a contribution to the spectrum and non-gaussianity of the curvature perturbation ζ. The field might appear directly in the preheating model (curvaton-type preheating) or indirectly through its effect on a mass or coupling (modulated preheating). We give general expressions for ζ based on the δN formula, and apply them to the cases of quadratic and quartic chaotic inflation. For the quadratic case, curvaton-type preheating is ineffective in contributing to ζ, but modulated preheating can be effective. For quartic inflation, curvaton-type preheating may be effective but the usual δN formalism has to be modified. We see under what circumstances the recent numerical simulation of Bond et al. [0903.3407] may be enough to provide a rough estimate for this case.

  4. On the generation of a non-gaussian curvature perturbation during preheating

    International Nuclear Information System (INIS)

    Kohri, Kazunori; Lyth, David H.; Valenzuela-Toledo, Cesar A.

    2010-01-01

    The perturbation of a light field might affect preheating and hence generate a contribution to the spectrum and non-gaussianity of the curvature perturbation ζ. The field might appear directly in the preheating model (curvaton-type preheating) or indirectly through its effect on a mass or coupling (modulated preheating). We give general expressions for ζ based on the δN formula, and apply them to the cases of quadratic and quartic chaotic inflation. For the quadratic case, curvaton-type preheating is ineffective in contributing to ζ, but modulated preheating can be effective. For quartic inflation, curvaton-type preheating may be effective but the usual δN formalism has to be modified. We see under what circumstances the recent numerical simulation of Bond et al. [0903.3407] may be enough to provide a rough estimate for this case

  5. Production of gravitational waves during preheating with nonminimal coupling

    Science.gov (United States)

    Fu, Chengjie; Wu, Puxun; Yu, Hongwei

    2018-04-01

    We study the preheating and the in-process production of gravitational waves (GWs) after inflation in which the inflaton is nonminimally coupled to the curvature in a self-interacting quartic potential with the method of lattice simulation. We find that the nonminimal coupling enhances the amplitude of the density spectrum of inflaton quanta, and as a result, the peak value of the GW spectrum generated during preheating is enhanced as well and might reach the limit of detection in future GW experiments. The peaks of the GW spectrum not only exhibit distinctive characteristics as compared to those of minimally coupled inflaton potentials but also imprint information on the nonminimal coupling and the parametric resonance, and thus the detection of these peaks in the future will provide us a new avenue to reveal the physics of the early universe.

  6. Thermographic study of the preheating plugs in diesel engines

    OpenAIRE

    Royo Pastor, Rafael; Albertos Arranz, M.A.; CÁRCEL CUBAS, JUAN ANTONIO; Payá Herrero, Jorge

    2012-01-01

    The use of direct injection diesel engines has been widely applied during the past ten years. In such engines, the preheating plugs are a key element which has a significant contribution in the pollutant emissions. In this paper, two different plug designs from Renault are analyzed. The new plug reduces substantially the required electrical consumption. Nevertheless, the pollutant emissions are higher (fundamentally CO and HCs) and hereby a thorough analysis is required to underst...

  7. DEFROST: a new code for simulating preheating after inflation

    International Nuclear Information System (INIS)

    Frolov, Andrei V

    2008-01-01

    At the end of inflation, dynamical instability can rapidly deposit the energy of homogeneous cold inflaton into excitations of other fields. This process, known as preheating, is rather violent, inhomogeneous and non-linear, and has to be studied numerically. This paper presents a new code for simulating scalar field dynamics in an expanding universe written for that purpose. Compared to available alternatives, it significantly improves both the speed and the accuracy of calculations, and is fully instrumented for 3D visualization. We reproduce previously published results on preheating in simple chaotic inflation models, and further investigate non-linear dynamics of the inflaton decay. Surprisingly, we find that the fields do not 'want' to thermalize in quite the way that one would think. Instead of directly reaching equilibrium, the evolution appears to be stuck in a rather simple but quite inhomogeneous state. In particular, a one-point distribution function of total energy density appears to be universal among various two-field preheating models, and is exceedingly well described by a log-normal distribution. It is tempting to attribute this state to scalar field turbulence

  8. CFD modeling of fouling in crude oil pre-heaters

    International Nuclear Information System (INIS)

    Bayat, Mahmoud; Aminian, Javad; Bazmi, Mansour; Shahhosseini, Shahrokh; Sharifi, Khashayar

    2012-01-01

    Highlights: ► A conceptual CFD-based model to predict fouling in industrial crude oil pre-heaters. ► Tracing fouling formation in the induction and developing continuation periods. ► Effect of chemical components, shell-side HTC and turbulent flow on the fouling rate. - Abstract: In this study, a conceptual procedure based on the computational fluid dynamic (CFD) technique has been developed to predict fouling rate in an industrial crude oil pre-heater. According to the developed CFD concept crude oil was assumed to be composed of three pseudo-components comprising of petroleum, asphaltene and salt. The binary diffusion coefficients were appropriately categorized into five different groups. The species transport model was applied to simulate the mixing and transport of chemical species. The possibility of adherence of reaction products to the wall was taken into account by applying a high viscosity for the products in competition with the shear stress on the wall. Results showed a reasonable agreement between the model predictions and the plant data. The CFD model could be applied to new operating conditions to investigate the details of the crude oil fouling in the industrial pre-heaters.

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

    International Nuclear Information System (INIS)

    1979-11-01

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

  10. Blanket comparison and selection study. Volume I

    International Nuclear Information System (INIS)

    1983-10-01

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

  11. Design requirement on HYPER blanket fuel assembly

    International Nuclear Information System (INIS)

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

    2000-07-01

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

  12. Liquid metal cooled blanket concept for NET

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  13. Fusion blankets for high efficiency power cycles

    International Nuclear Information System (INIS)

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

    1980-04-01

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

  14. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

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

    1987-10-01

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

  15. Regenerative heat exchanger for cowper combustion air preheating

    Energy Technology Data Exchange (ETDEWEB)

    Molenaar, R.; Otterbach, G.

    1986-01-13

    The waste gas leaving cowper units at a temperature of 200/sup 0/C to 300/sup 0/C was previously discharged unused into the atmosphere. By providing a suitable heat exchanger, the heat content of the waste gas can be used to preheat the combustion agents of cowpers to an extent allowing both to increase the efficiency of cowpers and to decrease the amount of rich gas required. The operating results confirm to a large extent the theoretical assumptions and calculations. One may therefore expect the entire investment to have been fully redeemed in a little more than two years. (orig.).

  16. Bruce NGS A Unit 4 preheater divider plate failure

    International Nuclear Information System (INIS)

    Landridge, M.; McInnes, D.

    1995-01-01

    On May 19, 1995, without any prior operational indications, Bruce A discovered preheater divider plate damage in Unit 4 that had the potential to have a major impact on the continued safe operation of the station. Further investigations indicated that Unit 4 may have been operating with this damage for as long as ten years. In the two months following the discovery, Bruce A has procured and replaced the 4 divider plates, located most of the missing pieces, retrieved pieces from the PHT system, investigated historical operational information, performed detailed analytical investigations, investigated root cause, performed in-situ and mock-up testing, updated operational procedures and installed DP monitoring equipment

  17. Fresh fuel pre-heating device in reactor facility

    International Nuclear Information System (INIS)

    Samejima, Asakuni.

    1988-01-01

    Purpose: To simplify the structure of a fresh nuclear fuel pre-heating device and improve the reliability to gas supply. Constitution: Fresh fuels taken out from a fresh fuel stredge rack and contained in a fuel strage pipe of a fuel transportation cask are pre-heated at the pre-stage of transfer by sending heating gases from the outside. Gas outlet pipes of the device are led out from the lower portion of the strage pipe, disposed side by side at the top of the strage pipe and opened upwardly. Further, gas supply pipes are connected to the inside of a movable guiding cylinder on the side of the floor surface and the opening end of return pipes are opposed to the exit opening end of the strage pipe. In such a constitution, a gas recycling loop can be formed between the strage pipe and the gas heating device by way of the movable guiding cylinder only by the operation of combining the fuel strage pipe of the transportation cask and the movable guiding pipe disposed on the side of the floor surface. Thus, the coupling structure is facilitated, the connection operation can surely be conducted to improve the reliability as compared with the conventional case. (Horiuchi, T.)

  18. (D,T) Driven thorium hybrid blankets

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  19. Design and analysis of ITER shield blanket

    Energy Technology Data Exchange (ETDEWEB)

    Ohmori, Junji; Hatano, Toshihisa; Ezato, Kouichiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-12-01

    This report includes electromagnetic analyses for ITER shielding blanket modules, fabrication methods for the blanket modules and the back plate, the design and the fabrication methods for port limiter have been investigated. Studies on the runaway electron impact for Be armor have been also performed. (J.P.N.)

  20. Methods to enhance blanket power density

    International Nuclear Information System (INIS)

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

    1982-06-01

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

  1. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Sato, Satoshi; Hatano, Toshihisa; Tokami, Ikuhide; Kitamura, Kazunori; Miura, Hidenori; Ito, Yutaka; Kuroda, Toshimasa; Takatsu, Hideyuki

    1997-05-01

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

  2. NET test blanket design and remote maintenance

    International Nuclear Information System (INIS)

    Holloway, C.; Hubert, P.

    1991-01-01

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

  3. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

    2009-01-01

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

  4. Blanket materials for DT fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1981-01-01

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

  5. Effects of substrate preheating during direct energy deposition on microstructure, hardness, tensile strength, and notch toughness

    Science.gov (United States)

    Baek, Gyeong Yun; Lee, Ki Yong; Park, Sang Hu; Shim, Do Sik

    2017-11-01

    This study examined the effects of substrate preheating for the hardfacing of cold-press dies using the high-speed tool steel AISI M4. The preheating of the substrate is a widely used technique for reducing the degree of thermal deformation and preventing crack formation. We investigated the changes in the metallurgical and mechanical properties of the high-speed tool steel M4 deposited on an AISI D2 substrate with changes in the substrate preheating temperature. Five preheating temperatures (100-500 °C; interval of 100 °C) were selected, and the changes in the temperature of the substrate during deposition were observed. As the preheating temperature of the substrate was increased, the temperature gradient between the melting layer and the substrate decreased; this prevented the formation of internal cracks, owing to thermal stress relief. Field-emission scanning electron microscopy showed that a dendritic structure was formed at the interface between the deposited layer and the substrate while a cellular microstructure was formed in the deposited layer. As the preheating temperature was increased, the sizes of the cells and precipitated carbides also increased. Furthermore, the hardness increased slightly while the strength and toughness decreased. Moreover, the tensile and impact properties deteriorated rapidly at excessively high preheating temperatures (greater than 500 °C). The results of this study can be used as preheating criteria for achieving the desired mechanical properties during the hardfacing of dies and molds.

  6. Convertible shielding to ceramic breeding blanket

    International Nuclear Information System (INIS)

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

    1995-05-01

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

  7. ARIES-IV Nested Shell Blanket Design

    International Nuclear Information System (INIS)

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

    1993-11-01

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

  8. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

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

    1999-07-01

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

  9. Tritium transport analysis for CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-15

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

  10. ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  11. Reheating the D-brane universe via instant preheating

    International Nuclear Information System (INIS)

    Panda, Sudhakar; Sami, M.; Thongkool, I.

    2010-01-01

    We investigate a possibility of reheating in a scenario of D-brane inflation in a warped deformed conifold background which includes perturbative corrections to throat geometry sourced by a chiral operator of dimension 3/2 in the conformal field theory. The effective D-brane potential, in this case, belongs to the class of nonoscillatory models of inflation for which the conventional reheating mechanism does not work. We find that gravitational particle production is inefficient and leads to reheating temperature of the order of 10 8 GeV. We show that instant preheating is quite suitable to the present scenario and can easily reheat the universe to a temperature which is higher by about 3 orders of magnitude than its counterpart associated with gravitational particle production. The reheating temperature is shown to be insensitive to a particular choice of inflationary parameters suitable to observations.

  12. 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)

  13. Beryllium R and D for blanket application

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  14. Beryllium R and D for blanket application

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  15. Beryllium R&D for blanket application

    Science.gov (United States)

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

    1998-10-01

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

  16. Mechanical and thermal design of hybrid blankets

    International Nuclear Information System (INIS)

    Schultz, K.R.

    1978-01-01

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

  17. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  18. Benchmark calculations for fusion blanket development

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  19. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  20. Review: BNL Tokamak graphite blanket design concepts

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  1. Blanket design for imploding liner systems

    International Nuclear Information System (INIS)

    Schaffer, M. J.

    1980-01-01

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

  2. APT target-blanket fabrication development

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, D.L.

    1997-06-13

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

  3. Stress analysis of the tokamak engineering test breeder blanket

    International Nuclear Information System (INIS)

    Huang Zhongqi

    1992-01-01

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

  4. The fusion blanket program at Chalk River

    International Nuclear Information System (INIS)

    Hastings, I.J.

    1986-03-01

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

  5. Thermal comfort and safety of cotton blankets warmed at 130°F and 200°F.

    Science.gov (United States)

    Kelly, Patricia A; Cooper, Susan K; Krogh, Mary L; Morse, Elizabeth C; Crandall, Craig G; Winslow, Elizabeth H; Balluck, Julie P

    2013-12-01

    In 2009, the ECRI Institute recommended warming cotton blankets in cabinets set at 130°F or less. However, there is limited research to support the use of this cabinet temperature. To measure skin temperatures and thermal comfort in healthy volunteers before and after application of blankets warmed in cabinets set at 130 and 200°F, respectively, and to determine the time-dependent cooling of cotton blankets after removal from warming cabinets set at the two temperatures. Prospective, comparative, descriptive. Participants (n = 20) received one or two blankets warmed in 130 or 200°F cabinets. First, skin temperatures were measured, and thermal comfort reports were obtained at fixed timed intervals. Second, blanket temperatures (n = 10) were measured at fixed intervals after removal from the cabinets. No skin temperatures approached levels reported in the literature that cause epidermal damage. Thermal comfort reports supported using blankets from the 200°F cabinet, and blankets lost heat quickly over time. We recommend warming cotton blankets in cabinets set at 200°F or less to improve thermal comfort without compromising patient safety. Copyright © 2013 American Society of PeriAnesthesia Nurses. Published by Elsevier Inc. All rights reserved.

  6. Advanced high performance solid wall blanket concepts

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  7. Workshop on cold-blanket research

    International Nuclear Information System (INIS)

    1977-05-01

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

  8. Powertrain preheating system of tracked hybrid electric vehicle in cold weather

    International Nuclear Information System (INIS)

    Wang, Rui; Wang, Yichun; Feng, Chaoqing; Zhang, Xilong

    2015-01-01

    In order to make sure that the heavy duty tracked vehicle can work in various conditions, especially severe cold weather, preheating system of powertrain should be adopted, and a novel preheating system is presented for the tracked hybrid electric vehicle (HEV) in which heat is generated by the low-speed drive motor. The new preheating system can meet the need of cold start without adding any additional device. The characteristic of heat generation by motor is tested when the rotor of motor is rotated in very low speed. The heat loss from power cabin to external environment has been simulated, and the relevant test has been done to verify the simulation results. Combining the characteristic of heat generation and heat loss situation about preheating system, the heat transfer model of preheating system was implemented by MATLAB. The total energy required for preheating in different ambient temperature was calculated by this model. The results showed that: the minimum heating power was 70 kW and energy required was about 180 MJ when the HEV worked in −46 °C. If lithium ferrous phosphate (LFP) battery was used in power system, the minimum battery capacity is about 290 A h. - Highlights: • A novel preheating method was proposed for heavy duty tracked HEV. • Thermal energy in preheating system is produced by the PMSM in driving system. • This method can achieve preheating target by its own components without any adding. • Analyzing low temperature performance of power battery and select its capacity.

  9. Liquid lithium blanket processing studies

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  10. Measurement of preheat in aluminium target in indirect drive using the SGIII prototype facilities

    International Nuclear Information System (INIS)

    Zhang, C; Zheng, J; Wang, Z B; Liu, H; Peng, X S; Wang, F; Ding, Y K

    2016-01-01

    The velocity interferometer system for any reflector (VISAR) is used to demonstrate preheat effect in aluminium in indirect drive. The rear surface motion prior to shock front was observed and compared with a multi-group calculation. By properly adjusting the hard x-ray portion of the radiation source, the calculated rear surface motion fits well with the experimental results, which gives us confidence to predict the preheated temperature of the sample by hard x-rays. Further, the effect of hohlraum geometry is compared and discussed experimentally. The result suggests gas-filled hohlraum or hohlraum with low Z substrates should be considered to further reduce preheating. (paper)

  11. Analysis of pre-heated fuel combustion and heat-emission dynamics in a diesel engine

    Science.gov (United States)

    Plotnikov, S. A.; Kartashevich, A. N.; Buzikov, S. V.

    2018-01-01

    The article explores the feasibility of diesel fuel pre-heating. The research goal was to obtain and analyze the performance diagrams of a diesel engine fed with pre-heated fuel. The engine was tested in two modes: at rated RPMs and at maximum torque. To process the diagrams the authors used technique developed by the Central Diesel Research Institute (CDRI). The diesel engine’s heat emission curves were obtained. The authors concluded that fuel pre-heating shortened the initial phase of the combustion process and moderated the loads, thus making it possible to boost a diesel engine’s mean effective pressure.

  12. Effect of pre-heat treatment on a Fischer-Tropsch iron catalyst

    International Nuclear Information System (INIS)

    Rao, K.R.P.M.; Huggins, F.E.; Ganguly, B.; Mahajan, V.; Huffman, G.P.; Davis, B.; O'Brien, R.J.; Xu Liguang; Rao, V.U.S.

    1994-01-01

    Moessbauer spectroscopy was used to investigate the effect of heating the Fischer-Tropsch catalyst 100 Fe/5 Cu/4.2 K/24 SiO 2 in two different atmospheres while ramping the temperature of the catalyst from room temperature to 280 C in 5.5 h prior to pretreatment of the catalyst. Preheating in H 2 /CO = 0.7 gave rise to an iron (Fe 2+ ) silicate, while preheating in helium resulted in the formation of ε'-carbide Fe 2.2 C. Iron oxides and χ-carbide Fe 5 C 2 were also formed in both preheat treatments. (orig.)

  13. Calculation and design of natural gas preheater equipments. Berechnung und Auslegung von Erdgas-Vorwaermeanlagen

    Energy Technology Data Exchange (ETDEWEB)

    Fasold, H G [Ruhrgas AG, Essen (Germany); Wahle, H N [Ruhrgas AG, Essen (Germany)

    1994-04-01

    A greatly simplified model of a regulating station - consisting of the station components ''preheater'' and ''control unit'' - is used for the calculation and design of natural gas preheating plants. It is hereby possible to calculate the Joule-Thomson effect which occurs on the expansion of natural gas in the controller, the resulting drop in temperature and the thermal output required to compensate this which is to be supplied to the gas flow by the preheating plant. The calculation method and procedure are explained using a programming flowchart. The computational model presented was converted into a personal computer program, whose functioning is elucidated using a numerical example. (orig.)

  14. arXiv Gravitational wave production from preheating -- parameter dependence

    CERN Document Server

    Figueroa, Daniel G.

    2017-10-31

    Parametric resonance is among the most efficient phenomena generating gravitational waves (GWs) in the early Universe. The dynamics of parametric resonance, and hence of the GWs, depend exclusively on the resonance parameter q. The latter is determined by the properties of each scenario: the initial amplitude and potential curvature of the oscillating field, and its coupling to other species. Previous works have only studied the GW production for fixed value(s) of q. We present an analytical derivation of the GW amplitude dependence on q, valid for any scenario, which we confront against numerical results. By running lattice simulations in an expanding grid, we study for a wide range of q values, the production of GWs in post-inflationary preheating scenarios driven by parametric resonance. We present simple fits for the final amplitude and position of the local maxima in the GW spectrum. Our parametrization allows to predict the location and amplitude of the GW background today, for an arbitrary q. The GW si...

  15. Experimental Investigation of Flow Resistance in a Coal Mine Ventilation Air Methane Preheated Catalytic Oxidation Reactor

    Directory of Open Access Journals (Sweden)

    Bin Zheng

    2015-01-01

    Full Text Available This paper reports the results of experimental investigation of flow resistance in a coal mine ventilation air methane preheated catalytic oxidation reactor. The experimental system was installed at the Energy Research Institute of Shandong University of Technology. The system has been used to investigate the effects of flow rate (200 Nm3/h to 1000 Nm3/h and catalytic oxidation bed average temperature (20°C to 560°C within the preheated catalytic oxidation reactor. The pressure drop and resistance proportion of catalytic oxidation bed, the heat exchanger preheating section, and the heat exchanger flue gas section were measured. In addition, based on a large number of experimental data, the empirical equations of flow resistance are obtained by the least square method. It can also be used in deriving much needed data for preheated catalytic oxidation designs when employed in industry.

  16. Preheat-induced signal enhancement in the infrared stimulated luminescence of young and bleached sediment samples

    International Nuclear Information System (INIS)

    Richardson, C.A.

    2000-01-01

    Natural and laboratory bleached surface and young samples of potassium feldspar sand separates and polymineral silt had their infrared stimulated luminescence (IRSL) signal measured before and after preheating at 220 deg. C for 10 min or 160 deg. C for 16 h. For both preheats, the laboratory bleached sand samples underwent a signal enhancement which was stable with laboratory storage. The youngest samples also showed natural signal enhancement. The silt sample showed no recuperation of bleached signal on preheating, but some in the natural signal. A range of filtered bleaches was applied to one surface sand sample. Signal levels before and after preheating were reduced by filtering out the UV from the bleaching spectrum. The unfiltered bleach, however, most closely reproduced the behaviour of the natural sample

  17. Waste heat recovery at the glass industry with the intervention of batch and cullet preheating

    OpenAIRE

    Dolianitis Ioannis; Giannakopoulos Dionysios; Hatzilau Christina-Stavrula; Karellas Sotirios; Kakaras Emmanuil; Nikolova Evelina; Skarpetis Georgios; Christodoulou Nikolaos; Giannoulas Nikolaos; Zitounis Theodoros

    2016-01-01

    A promising option to reduce the specific energy consumption and CO2 emissions at a conventional natural gas fired container glass furnace deals with the advanced utilization of the exhaust gases downstream the air regenerators by means of batch and cullet preheating. A 3-dimensional computational model that simulates this process using mass and heat transfer equations inside a preheater has been developed. A case study for an efficient small-sized containe...

  18. Test Blanket Working Group's recent activities

    International Nuclear Information System (INIS)

    Vetter, J.E.

    2001-01-01

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

  19. Microwave pre-heating of natural rubber using a rectangular wave guide (MODE: TE10

    Directory of Open Access Journals (Sweden)

    Doo-ngam, N.

    2007-11-01

    Full Text Available This paper presents an application of microwave radiation for pre-heating of natural rubbercompounding with various sulphur contents. The natural rubber-compounding was pre-heated by microwave radiation using a rectangular wave guide system (MODE: TE10 operating at frequency of 2.45 GHz in which the power can vary from 0 to 1500 W. In the present work, the influence of power input, sample thickness, and sulphur content were examined after applying microwave radiation to the rubber samples. Results are discussed regarding the thermal properties, 3-D network, dielectric properties and chemical structures. From the result, firstly, it was found that microwave radiation can be applied to pre-heating natural rubber-compounding before the vulcanization process. Secondly, microwave radiation was very useful for pre-heating natural rubber-compounding that has a thickness greater than 5mm. Thirdly, crosslinking in natural rubber-compounding may occurs after pre-heating by microwave radiation though Fourier Transform Infrared Spectroscopy(FTIR. Finally, there a little effect of sulphur content on temperature profiles after applying microwave radiation to the natural rubber-compounding. Moreover, natural rubber-compounding without carbon black showed a lower heat absorption compared with natural rubbercompounding filled carbon black. This is due to the difference in dielectric loss factor. This preliminary result will be useful information in terms of microwave radiation for pre-heating natural rubber-compounding and rubber processing in industry.

  20. Optimal Substrate Preheating Model for Thermal Spray Deposition of Thermosets onto Polymer Matrix Composites

    Science.gov (United States)

    Ivosevic, M.; Knight, R.; Kalidindi, S. R.; Palmese, G. R.; Tsurikov, A.; Sutter, J. K.

    2003-01-01

    High velocity oxy-fuel (HVOF) sprayed, functionally graded polyimide/WC-Co composite coatings on polymer matrix composites (PMC's) are being investigated for applications in turbine engine technologies. This requires that the polyimide, used as the matrix material, be fully crosslinked during deposition in order to maximize its engineering properties. The rapid heating and cooling nature of the HVOF spray process and the high heat flux through the coating into the substrate typically do not allow sufficient time at temperature for curing of the thermoset. It was hypothesized that external substrate preheating might enhance the deposition behavior and curing reaction during the thermal spraying of polyimide thermosets. A simple analytical process model for the deposition of thermosetting polyimide onto polymer matrix composites by HVOF thermal spray technology has been developed. The model incorporates various heat transfer mechanisms and enables surface temperature profiles of the coating to be simulated, primarily as a function of substrate preheating temperature. Four cases were modeled: (i) no substrate preheating; (ii) substrates electrically preheated from the rear; (iii) substrates preheated by hot air from the front face; and (iv) substrates electrically preheated from the rear and by hot air from the front.

  1. First Wall, Blanket, Shield Engineering Technology Program

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1982-01-01

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

  2. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1994-11-01

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

  3. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

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

    1987-11-01

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

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

    International Nuclear Information System (INIS)

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

    1988-03-01

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

  5. LMFBR blanket physics project progress report No. 4

    International Nuclear Information System (INIS)

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

    1973-01-01

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

  6. Epoxy blanket protects milled part during explosive forming

    Science.gov (United States)

    1966-01-01

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

  7. Some new ideas for Tandem Mirror blankets

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.

    1981-01-01

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

  8. Fusion blanket high-temperature heat transfer

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1983-01-01

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

  9. Tritium behaviour in ceramic breeder blankets

    International Nuclear Information System (INIS)

    Miller, J.M.

    1989-01-01

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

  10. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1988-01-01

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

  12. Tritium inventory and permeation in liquid breeder blankets

    International Nuclear Information System (INIS)

    Reiter, F.

    1990-01-01

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

  13. Corrosion on air preheaters and economisers; Korrosion hos luftfoervaermare och ekonomisrar

    Energy Technology Data Exchange (ETDEWEB)

    Nordling, Magnus

    2012-05-15

    Combustion plants in Sweden are exposed to considerable stress regarding low temperature corrosion, and failures due to low temperature corrosion occur regularly. Particularly common is corrosion problems connected to air preheaters and economisers. The number of combustion plants having air preheaters and economisers is however large, and the result of a collection of experiences regarding corrosion on air preheaters and economisers therefore has the potential to give a broad knowledge base. The summary of collection of experiences that has been done here, complemented with a literature survey, is expected to give plant owners and plant constructors a valuable tool to prevent corrosion on the flue gas side of air preheaters and economisers. The choice of plants for the inquiry was made using a list from the Swedish Naturvaardsverket (Environmental Protection Agency) indicating the emissions of NO{sub x}gases from Swedish combustion plants. From that list mainly the plants with the largest emissions were chosen, resulting in a number of 30 plants. Depending on that most of the plants have several boilers, and that the connected tubes often have several economisers and air preheaters, the number of economisers and air preheaters in this experience collection is at least 85. The study was however not limited to economisers and air preheaters, but also experiences connected to corrosion of other units were collected when mentioned, and the most interesting information here is also included in the report. Also a number of the plants were visited to improve the basis of the report, e.g. by photographing the most interesting parts. As the insight of the extension of the problem increased, renewed interview rounds were made, and the last one was made in August 2011.

  14. INTOR first wall/blanket/shield activity

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  15. Optimization of beryllium for fusion blanket applications

    International Nuclear Information System (INIS)

    Billone, M.C.

    1993-01-01

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

  16. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  17. European blanket development for a demo reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-09-01

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

  19. Flexural Strength of Preheated Resin Composites and Bonding Properties to Glass-Ceramic and Dentin

    Directory of Open Access Journals (Sweden)

    Matthias Richard Kramer

    2016-01-01

    Full Text Available To test the impact of preheating (25, 37, 54, or 68 °C of TetricEvoCeram (TEC, FiltekSupremeXT (FSXT, and Venus (V on flexural strength (FS, shear bond strength (SBS and interfacial tension (IFT. FS was tested with TEC and FSXT. For SBS, glass-ceramic and human dentin substrate were fabricated and luted with the preheated resin composite (RC. SBSs of 1500 thermal cycled specimens were measured. For IFT, glass slides covered with the non-polymerized RC were prepared and contact angles were measured. Data were analyzed using 2/1-way ANOVA with Scheffé-test, and t-test (p < 0.05. Preheated TEC (37–68 °C showed higher FS compared to the control-group (25 °C (p < 0.001. FSXT presented higher FS than TEC (p < 0.001. For SBS to dentin higher values for FSXT than TEC were found. The preheating temperature showed no impact on SBS to dentin. SBS to glass-ceramic revealed a positive influence of temperature for TEC 25–68 °C (p = 0.015. TEC showed higher values than V and FSXT (p < 0.001. IFT values increased with the preheating temperature. A significant difference could be observed in every RC group between 25 and 68 °C (p < 0.001.

  20. Effect of pre-heating on the viscosity and microhardness of a resin composite.

    LENUS (Irish Health Repository)

    Lucey, S

    2010-04-01

    The effect of pre-heating resin composite on pre-cured viscosity and post-cured surface hardness was evaluated. Groups of uncured specimens were heated to 60 degrees C and compared with control groups (24 degrees C) with respect to viscosity and surface hardness. Mean (SD) viscosities of the pre-heated specimens (n = 15) were in the range of 285 (13)-377 (11) (Pa) compared with 642 (35)-800 (23) (Pa) at ambient temperature. There was a statistically significant difference between the two groups (P < 0.001). Mean (SD) Vickers microhardness (VHN) of the pre-heated group (n = 15) was 68.6 (2.3) for the top surface and 68.7 (1.8) for the bottom surface measured at 24 h post curing (specimen thickness = 1.5 mm). The corresponding values for the room temperature group were 60.6 (1.4) and 59.0 (3.5). There was a statistically significant difference between corresponding measurements taken at the top and bottom for the pre-heated and room temperature groups (P < 0.001). There was no significant difference between top and bottom measurements within each group. Pre-heating resin composite reduces its pre-cured viscosity and enhances its subsequent surface hardness. These effects may translate as easier placement together with an increased degree of polymerization and depth-of-cure.

  1. Efficiency of the pre-heater against flow rate on primary the beta test loop

    International Nuclear Information System (INIS)

    Edy Sumarno; Kiswanta; Bambang Heru; Ainur R; Joko P

    2013-01-01

    Calculation of efficiency of the pre-heater has been carried out against the flow rate on primary the BETA Test Loop. BETA test loop (UUB) is a facilities of experiments to study the thermal hydraulic phenomenon, especially for thermal hydraulic post-LOCA (Lost of Coolant Accident). Sequences removal on the BETA Test Loop contained a pre-heater that serves as a getter heat from the primary side to the secondary side, determination of efficiency is to compare the incoming heat energy with the energy taken out by a secondary fluid. Characterization is intended to determine the performance of a pre-heater, then used as tool for analysis, and as a reference design experiments. Calculation of efficiency methods performed by operating the pre-heater with fluid flow rate variation on the primary side. Calculation of efficiency on the results obtained that the efficiency change with every change of flow rate, the flow rate is 71.26% on 163.50 ml/s and 60.65% on 850.90 ml/s. Efficiency value can be even greater if the pre-heater tank is wrapped with thermal insulation so there is no heat leakage. (author)

  2. Exergy analysis on the irreversibility of rotary air preheater in thermal power plant

    International Nuclear Information System (INIS)

    Wang Hongyue; Zhao Lingling; Zhou Qiangtai; Xu Zhigao; Kim, Hyung Taek

    2008-01-01

    Energy recovery devices can have a substantial impact on process efficiency and their relevance to the problem of conservation of energy resources is generally recognized to be beyond dispute. One type of such a device, which is commonly used in thermal power plants and air conditioning systems, is the rotary air preheater. A major disadvantage of the rotary air preheater is that there is an unavoidable leakage due to carry over and pressure difference. There are gas streams involved in the heat transfer and mixing processes. There are also irreversibilities, or exergy destruction, due to mixing, pressure losses and temperature gradients. Therefore, the purpose of this research paper is based from the second law of thermodynamics, which is to build up the relationship between the efficiency of the thermal power plant and the total process of irreversibility in the rotary air preheater using exergy analysis. For this, the effects of the variation of the principal design parameters on the rotary air preheater efficiency, the exergy efficiency, and the efficiency of the thermal power plant are examined by changing a number of parameters of rotary air preheater. Furthermore, some conclusions are reached and recommendations are made so as to give insight on designing some optimal parameters

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

    International Nuclear Information System (INIS)

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

    1980-09-01

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

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

    International Nuclear Information System (INIS)

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

    1979-09-01

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

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

    International Nuclear Information System (INIS)

    Chen Hongli; Wu Yican; Bai Yunqing

    2006-01-01

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

  6. Waste heat recovery at the glass industry with the intervention of batch and cullet preheating

    Directory of Open Access Journals (Sweden)

    Dolianitis Ioannis

    2016-01-01

    Full Text Available A promising option to reduce the specific energy consumption and CO2 emissions at a conventional natural gas fired container glass furnace deals with the advanced utilization of the exhaust gases downstream the air regenerators by means of batch and cullet preheating. A 3-dimensional computational model that simulates this process using mass and heat transfer equations inside a preheater has been developed. A case study for an efficient small-sized container glass furnace is presented dealing with the investigation of the impact of different operating and design configurations on specific energy consumption, CO2 emissions, flue gas energy recovery, batch temperature and preheater efficiency. In specific, the effect of various parameters is studied, including the preheater’s dimensions, flue gas temperature, batch moisture content, glass pull, combustion air excess and cullet fraction. Expected energy savings margin is estimated to 12-15%.

  7. Effects of preheated combustion air on laminar coflow diffusion flames under normal and microgravity conditions

    Science.gov (United States)

    Ghaderi Yeganeh, Mohammad

    Global energy consumption has been increasing around the world, owing to the rapid growth of industrialization and improvements in the standard of living. As a result, more carbon dioxide and nitrogen oxide are being released into the environment. Therefore, techniques for achieving combustion at reduced carbon dioxide and nitric oxide emission levels have drawn increased attention. Combustion with a highly preheated air and low-oxygen concentration has been shown to provide significant energy savings, reduce pollution and equipment size, and uniform thermal characteristics within the combustion chamber. However, the fundamental understanding of this technique is limited. The motivation of the present study is to identify the effects of preheated combustion air on laminar coflow diffusion flames. Combustion characteristics of laminar coflow diffusion flames are evaluated for the effects of preheated combustion air temperature under normal and low-gravity conditions. Experimental measurements are conducted using direct flame photography, particle image velocimetry (PIV) and optical emission spectroscopy diagnostics. Laminar coflow diffusion flames are examined under four experimental conditions: normal-temperature/normal-gravity (case I), preheated-temperature/normal gravity (case II), normal-temperature/low-gravity (case III), and preheated-temperature/low-gravity (case IV). Comparisons between these four cases yield significant insights. In our studies, increasing the combustion air temperature by 400 K (from 300 K to 700 K), causes a 37.1% reduction in the flame length and about a 25% increase in peak flame temperature. The results also show that a 400 K increase in the preheated air temperature increases CH concentration of the flame by about 83.3% (CH is a marker for the rate of chemical reaction), and also increases the C2 concentration by about 60% (C2 is a marker for the soot precursor). It can therefore be concluded that preheating the combustion air

  8. Experimental Investigation of Flow Resistance in a Coal Mine Ventilation Air Methane Preheated Catalytic Oxidation Reactor

    OpenAIRE

    Zheng, Bin; Liu, Yongqi; Liu, Ruixiang; Meng, Jian; Mao, Mingming

    2015-01-01

    This paper reports the results of experimental investigation of flow resistance in a coal mine ventilation air methane preheated catalytic oxidation reactor. The experimental system was installed at the Energy Research Institute of Shandong University of Technology. The system has been used to investigate the effects of flow rate (200 Nm3/h to 1000 Nm3/h) and catalytic oxidation bed average temperature (20°C to 560°C) within the preheated catalytic oxidation reactor. The pressure drop and res...

  9. Solar pre-heating of water for steam generation in the friendship textile mill

    International Nuclear Information System (INIS)

    Sid -Ahmed, M.O.; Hussien, T.

    1994-01-01

    The technology of solar water heating is simple and can be used for pre-heating of water entering a boiler. In this paper the economics of solar pre-heating of water was calculated. The calculations were based on the performance and cost of a locally-made flat plate collector, and the performance and fuel consumption of a boiler in a textile mill. The results showed that a collector area of about 800 meter square with initial cost of about LS 5,000,000, could save annually about 130 tons of furnace oil. ( Author )

  10. EFFECT OF PRE-HEAT TREATMENT ON MECHANICAL PROPERTIES OF Ti-6Al-4V WELDS

    Directory of Open Access Journals (Sweden)

    Gnofam Jacques TCHEIN

    2016-11-01

    Full Text Available The work presented here is related to the optimization of the Friction Stir Welding (FSW process. The objective is to study the influence of some parameters used in the production of welded joints by FSW. The most important parameters are the welding speed and the rotational speed of the tool. The effect of pre-heat treatment on the plates to be welded is also studied by the design of experimental methods. These pre-heat treatments result not only in a change of mechanical properties of plates to be welded, but also of their microstructure. The experiments were performed following a 16 lines fractional Taguchi table.

  11. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  12. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    Carvalho, S.H. de.

    1980-03-01

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

  13. Conceptual design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Sato, Satoshi; Takatsu, Hideyuki; Kurasawa, Toshimasa

    1995-03-01

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

  14. Flow balancing in liquid metal blankets

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  15. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-01-01

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

  16. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  17. Conceptual design of Blanket Remote Handling System for CFETR

    International Nuclear Information System (INIS)

    Wei, Jianghua; Song, Yuntao; Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong

    2015-01-01

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  18. Conceptual design of Blanket Remote Handling System for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Jianghua, E-mail: weijh@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2015-11-15

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  19. Economic evaluation of the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Waganer, L.M.

    1985-01-01

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

  20. Influence of Powder Bed Preheating on Microstructure and Mechanical Properties of H13 Tool Steel SLM Parts

    Science.gov (United States)

    Mertens, R.; Vrancken, B.; Holmstock, N.; Kinds, Y.; Kruth, J.-P.; Van Humbeeck, J.

    Powder bed preheating is a promising development in selective laser melting (SLM), mainly applied to avoid large thermal stresses in the material. This study analyses the effect of in-process preheating on microstructure, mechanical properties and residual stresses during SLM of H13 tool steel. Sample parts are produced without any preheating and are compared to the corresponding parts made with preheating at 100°, 200°, 300°, and 400°C. Interestingly, internal stresses at the top surface of the parts evolve from compressive (-324MPa) without preheating to tensile stresses (371MPa) with preheating at 400°C. Nevertheless, application of powder bed preheating results in a more homogeneous microstructure with better mechanical properties compared to H13 SLM parts produced without preheating. The fine bainitic microstructure leads to hardness values of 650-700Hv and ultimate tensile strength of 1965MPa, which are comparable to or even better than those of conventionally made and heat treated H13 tool steel.

  1. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    1980-08-01

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

  2. Nuclear Analysis of an ITER Blanket Module

    Science.gov (United States)

    Chiovaro, P.; Di Maio, P. A.; Parrinello, V.

    2013-08-01

    ITER blanket system is the reactor's plasma-facing component, it is mainly devoted to provide the thermal and nuclear shielding of the Vacuum Vessel and external ITER components, being intended also to act as plasma limiter. It consists of 440 individual modules which are located in the inboard, upper and outboard regions of the reactor. In this paper attention has been focused on to a single outboard blanket module located in the equatorial zone, whose nuclear response under irradiation has been investigated following a numerical approach based on the Monte Carlo method and adopting the MCNP5 code. The main features of this blanket module nuclear behaviour have been determined, paying particular attention to energy and spatial distribution of the neutron flux and deposited nuclear power together with the spatial distribution of its volumetric density. Moreover, the neutronic damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rates. Finally, an activation analysis has been performed with FISPACT inventory code using, as input, the evaluated neutron spectrum to assess the module specific activity and contact dose rate after irradiation under a specific operating scenario.

  3. Tokamak blanket design study, final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

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

  4. Area 3, SRC-II coal slurry preheater studies report for the technical data analysis program

    Energy Technology Data Exchange (ETDEWEB)

    1984-08-01

    This report reviews the raw data gathered from the Preheater B test runs at Ft. Lewis, and also the Preheater B results presented in the Solvent Refined Coal (SRC) Process Final Report, Volumes 1 and 2 of Slurry Preheater Design, SRC-II Process and the Ft. Lewis Slurry Preheater Data Analysis, 1 1/2 Inch Coil by Gulf Science and Technology Corporation of Pittsburgh, Pennsylvania. attempts were made to correlate several variables not previously considered with slurry viscosity and thermal conductivity. Only partial success was realized. However, in the process of attempting to correlate these variables an understanding of why some variables could not be correlated was achieved. An attempt was also made, using multiple linear regression, to correlate coal slurry viscosity and thermal conductivity with several independent variables among which were temperature, coal concentration, total solids, coal type, slurry residence time, shear rate, and unit size. The final correlations included some, but not all, of these independent variables. This report is not a stand alone document and should be considered a supplement to work already done. It should be read in conjunction with the reports referenced above.

  5. Experimental Investigation of the Effects of Concrete Alkalinity on Tensile Properties of Preheated Structural GFRP Rebar

    Directory of Open Access Journals (Sweden)

    Hwasung Roh

    2017-01-01

    Full Text Available The combined effects of preexposure to high temperature and alkalinity on the tensile performance of structural GFRP reinforcing bars are experimentally investigated. A total of 105 GFRP bar specimens are preexposed to high temperature between 120°C and 200°C and then immersed into pH of 12.6 alkaline solution for 100, 300, and 660 days. From the test results, the elastic modulus obtained at 300 immersion days is almost the same as those of 660 immersion days. For all alkali immersion days considered in the test, the preheated specimens provide slightly lower elastic modulus than the unpreheated specimens, showing only 8% maximum difference. The tensile strength decreases for all testing cases as the increase of the alkaline immersing time, regardless of the prehearing levels. The tensile strength of the preheated specimens is about 90% of the unpreheated specimen for 300 alkali immersion days. However, after 300 alkali immersion days the tensile strengths are almost identical to each other. Such results indicate that the tensile strength and elastic modulus of the structural GFRP reinforcing bars are closely related to alkali immersion days, not much related to the preheating levels. The specimens show a typical tensile failure around the preheated location.

  6. Preheating to around 100°C under endcap blocks before welding at KHI.

    CERN Multimedia

    Loveless, D

    2000-01-01

    The 600mm thick sector blocks of the CMS endcaps are made from three layers of 200mm plates welded together. During the manufacture at KHI, the blocks are preheated to around 100°C to prevent cracks in the welds.

  7. Enhanced preheating after multi-field inflation: on the importance of being special

    International Nuclear Information System (INIS)

    Battefeld, Thorsten; Eggemeier, Alexander; Giblin, John T. Jr.

    2012-01-01

    We discuss preheating after multi-field inflation in the presence of several preheat matter fields that become light in the vicinity of (but not at) the inflatons' VEV, at distinct extra-species-points (ESP); this setup is motivated by inflationary models that include particle production during inflation, e.g. trapped inflation, grazing ESP encounters or modulated trapping, among others. While de-phasing of inflatons tends to suppress parametric resonance, we find two new effects leading to efficient preheating: particle production during the first in-fall (efficient if many preheat matter fields are present) and a subsequent (narrow) resonance phase (efficient if an ESP happens to be at one of several distinct distances from the inflatons' VEV). Particles produced during the first in-fall are comprised of many species with low occupation number, while the latter are made up of a few species with high occupation number. We provide analytic descriptions of both phases in the absence of back-reaction, which we test numerically. We further perform lattice simulations to investigate the effects of back-reaction. We find resonances to be robust and the most likely cause of inflaton decay in multi-field trapped inflation if ESP distributions are dense

  8. Effect of air preheat temperature on the MILD combustion of syngas

    International Nuclear Information System (INIS)

    Huang, Mingming; Zhang, Zhedian; Shao, Weiwei; Xiong, Yan; Liu, Yan; Lei, Fulin; Xiao, Yunhan

    2014-01-01

    Highlights: • MILD combustion is achieved with reaction zone covering the entire combustion chamber. • Critical equivalence ratio for the occurrence of MILD combustion is identified. • MILD regime can be established for syngas fuel under air preheating conditions. - Abstract: The effect of air preheat temperature on MILD (Moderate or Intense Low-oxygen Dilution) combustion of coal-derived syngas was examined in parallel jet forward flow combustor. The results were presented on flow field using numerical simulations and on global flame signatures, OH ∗ radicals distribution and exhaust emissions using experiments. The discrete and high speed air/fuel injections into the combustor is necessary for the establishment of MILD conditions, because they cause strong gas recirculation and form large mixing region between the air and fuel jets. The critical equivalence ratio above which MILD combustion occurred was identified. The MILD regime was established for syngas fuel under air preheating conditions with lean operational limit and suppressed NO x and CO emissions. In the MILD combustion regime, the air preheating resulted in higher NO x but lower CO emissions, while the increase of equivalence ratio led to the increase of NO x and the decrease of CO emissions

  9. Diagnostics of electron-heated solar flare models. III - Effects of tapered loop geometry and preheating

    Science.gov (United States)

    Emslie, A. G.; Li, Peng; Mariska, John T.

    1992-01-01

    A series of hydrodynamic numerical simulations of nonthermal electron-heated solar flare atmospheres and their corresponding soft X-ray Ca XIX emission-line profiles, under the conditions of tapered flare loop geometry and/or a preheated atmosphere, is presented. The degree of tapering is parameterized by the magnetic mirror ratio, while the preheated atmosphere is parameterized by the initial upper chromospheric pressure. In a tapered flare loop, it is found that the upward motion of evaporated material is faster compared with the case where the flare loop is uniform. This is due to the diverging nozzle seen by the upflowing material. In the case where the flare atmosphere is preheated and the flare geometry is uniform, the response of the atmosphere to the electron collisional heating is slow. The upward velocity of the hydrodynamic gas is reduced due not only to the large coronal column depth, but also to the increased inertia of the overlying material. It is concluded that the only possible electron-heated scenario in which the predicted Ca XIX line profiles agree with the BCS observations is when the impulsive flare starts in a preheated dense corona.

  10. Effect of pre-heating on the thermal decomposition kinetics of cotton

    Science.gov (United States)

    The effect of pre-heating at low temperatures (160-280°C) on the thermal decomposition kinetics of scoured cotton fabrics was investigated by thermogravimetric analysis under nonisothermal conditions. Isoconversional methods were used to calculate the activation energies for the pyrolysis after one-...

  11. Enhanced preheating after multi-field inflation: on the importance of being special

    Energy Technology Data Exchange (ETDEWEB)

    Battefeld, Thorsten; Eggemeier, Alexander [Institute for Astrophysics, University of Goettingen, Friedrich Hund Platz 1, D-37077 Goettingen (Germany); Giblin, John T. Jr., E-mail: tbattefe@astro.physik.uni-goettingen.de, E-mail: a.eggemeier@stud.uni-goettingen.de, E-mail: giblinj@kenyon.edu [Department of Physics, Kenyon College, Gambier, OH 43022 (United States)

    2012-11-01

    We discuss preheating after multi-field inflation in the presence of several preheat matter fields that become light in the vicinity of (but not at) the inflatons' VEV, at distinct extra-species-points (ESP); this setup is motivated by inflationary models that include particle production during inflation, e.g. trapped inflation, grazing ESP encounters or modulated trapping, among others. While de-phasing of inflatons tends to suppress parametric resonance, we find two new effects leading to efficient preheating: particle production during the first in-fall (efficient if many preheat matter fields are present) and a subsequent (narrow) resonance phase (efficient if an ESP happens to be at one of several distinct distances from the inflatons' VEV). Particles produced during the first in-fall are comprised of many species with low occupation number, while the latter are made up of a few species with high occupation number. We provide analytic descriptions of both phases in the absence of back-reaction, which we test numerically. We further perform lattice simulations to investigate the effects of back-reaction. We find resonances to be robust and the most likely cause of inflaton decay in multi-field trapped inflation if ESP distributions are dense.

  12. Symbiotic potential: the integration of preheating and dry cooling in cokemaking

    Energy Technology Data Exchange (ETDEWEB)

    Barker, J E

    1978-06-01

    In the USSR and Japan, heat recovered from the dry cooling of coke is used to raise steam for power generation or process use. This heat could be used to dry and preheat coal to improve both coke quality and oven productivity.

  13. Nuclear characteristics of D-D fusion reactor blankets

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao

    1978-01-01

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

  14. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.A.

    1980-01-01

    Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture

  15. Processing and waste disposal representative for fusion breeder blanket systems

    International Nuclear Information System (INIS)

    Finn, P.A.; Vogler, S.

    1987-01-01

    This study is an evaluation of the waste handling concepts applicable to fusion breeder systems. Its goal is to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under US Nuclear Regulatory regulations. The radionuclides expected in the materials used in fusion reactor blankets are described, as are plans for reprocessing and disposal of the components of different breeder blankets. An estimate of the operating costs involved in waste disposal is made

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  17. Electromagnetic analysis of ITER shield blanket under VDE

    International Nuclear Information System (INIS)

    Kang Weishan; Chen Jiming; Wu Jihong; Wang Mingxu

    2010-01-01

    Electromagnetic force and torque of ITER shield blanket system and their surrounding major component under vertical displacement event (VDE) were calculated with finite element method. ANSYS APDL was used to simulate the shape and magnitude of plasmas current dynamically in the VDE course, and external magnetic field was imposed, then the induced current distribution inside the all conductor including the blanket was obtained from the calculation. The force and torque for every blanket module was obtained to assess the safety of blanket system under VDE. (authors)

  18. Minimum thickness blanket-shield for fusion reactors

    International Nuclear Information System (INIS)

    Karni, Y.; Greenspan, E.

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Finn, P.A.

    1985-01-01

    This paper describes the methodology developed for estimating the relative capital costs of blanket processing systems. The capital costs of the nine blanket concepts selected in the Blanket Comparison and Selection Study are presented and compared

  20. Beryllium research on FFHR molten salt blanket

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  1. Structural materials for fusion reactor blanket systems

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  2. About possible technologies of creation nanostructures blankets

    International Nuclear Information System (INIS)

    Blednova, Zh.M.; Chaevskij, M.I.; Rusinov, P.O.

    2008-01-01

    Possible technologies of formation nanostructures blankets are considered: a method of thermal carrying over of weights in the conditions of a high gradient of temperatures; the combined method including cathode-plasma nitriding in the conditions of low pressure and drawing of nitride of the titan in a uniform work cycle; the combined method including high-frequency ionic nitriding and drawing of carbide of chrome by pyrolysis chrome and organic of connections in plasma of the decaying category. Possibility of formation layered nanostructures layers is shown.

  3. ITER solid breeder blanket materials database

    International Nuclear Information System (INIS)

    Billone, M.C.; Dienst, W.; Noda, K.; Roux, N.

    1993-11-01

    The databases for solid breeder ceramics (Li 2 ,O, Li 4 SiO 4 , Li 2 ZrO 3 and LiAlO 2 ) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized

  4. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-06-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-1 through 4 and PULSAR 1 and 2. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. Also, the requirements of engineering and physics systems for a pulsed reactor were evaluated by the PULSAR design studies. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies

  5. Reproducibility of LiF:Mg, Cup thermoluminescent dosimeter on kilo voltage and megavoltage photon beam using different preheat rate:A glow curve study

    International Nuclear Information System (INIS)

    Mohd Fahmi Mohd Yusof; Robert, T.S.B.; Puteri Norkhatijah Abdul Hamid; Nor Shazleen Abdul Shukor; Mohd Sazarman Mohd Salleh

    2013-01-01

    Full-text: Post-irradiation annealing or preheat of the LiF based TLD prior readout is commonly practiced for routine dosimetry to eliminate low temperature glow peaks. The aim of this study is to determine the effect of different preheating rate technique prior readout on the reproducibility and glow curve structure of LiF:Mg, Cu, P or TLD-1OOH exposed to low (109kVp) energy and high energy (6MV) photon beam. TLD chips were read after 24 hours of irradiation with three different preheat techniques; no preheat, low preheat rate (100 degree Celsius/ 10 minutes) and high preheat rate (135 degree Celsius/ 10 seconds) and reproducibility of TL signals were assessed in term of Standard Deviation (SD) and glow curve peaks. The high preheat rate technique was the most reproducible method for low energy photon with 1.05 % of mean reproducibility followed by low preheat rate (1.16 %) and no-preheat (1.33 %) techniques. The high preheat rate techniques was also the most reproducible method for high energy photon with 0.767 % of mean reproducibility as compared to low preheat rate (1.281 %). However the high preheat technique record highest TL signal lost with 10.35 % and 6.04 % for 24 and 72 hours of delayed TLD readout with respectively compared to 9.27 % and 4.51 % for 24 and 72 hours by low preheat rate. The low preheat was found to be optimal to eliminate low peaks (peak 1 and 2) but enable to remove peak 3 as it was shifted up word to combine with the main peak 4 of TL glow peak. It can be concluded that the reproducibility and structure of glow curve was strongly influenced by preheat technique prior readout. (author)

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  7. Overview of the TFTB lithium blanket module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

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

  8. Remote handling demonstration of ITER blanket module replacement

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  9. Accelerator driven heavy water blanket on circulating fuel

    International Nuclear Information System (INIS)

    Kazaritsky, V.D.; Blagovolin, P.P.; Mladov, V.R.; Okhlopkov, M.L.; Batyaev, V.F.; Stepanov, N.V.; Seliverstov, V.V.

    1997-01-01

    A conceptual design of a heavy water blanket with circulating fuel for an accelerator driven transmutation system is described. The hybrid system consists of a high-current linear accelerator of protons and 4 targets, each placed inside a subcritical blanket

  10. Achievements of element technology development for breeding blanket

    International Nuclear Information System (INIS)

    Enoeda, Mikio

    2005-03-01

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

  11. An assessment of the base blanket for ITER

    International Nuclear Information System (INIS)

    Raffray, A.R.; Abdou, M.A.; Ying, A.

    1991-01-01

    Ideally, the ITER base blanket would provide the necessary tritium for the reactor to be self-sufficient during operation, while having minimal impact on the overall reactor cost, reliability and safety. A solid breeder blanket has been developed in CDA phase in an attempt to achieve such objectives. The reference solid breeder base blanket configurations at the end of the CDA phase has many attractive features such as a tritium breeding ratio (TBR) of 0.8--0.9 and a reasonably low tritium inventory. However, some concerns regarding the risk, cost and benefit of the base blanket have been raised. These include uncertainties associated with the solid breeder thermal control and the potentially high cost of the amount of Be used to achieve high TBR and to provide the necessary thermal barrier between the high temperature solid breeder and low temperature coolant. This work addresses these concerns. The basis for the selection of a breeding blanket is first discussed in light of the incremental risk, cost and benefits relative to a non-breeding blanket. Key issues associated with the CDA breeding blanket configurations are then analyzed. Finally, alternative schemes that could enhance the attractiveness and flexibility of a breeding blanket are explored

  12. Objectives and status of EUROfusion DEMO blanket studies

    Energy Technology Data Exchange (ETDEWEB)

    Boccaccini, L.V., E-mail: lorenzo.boccaccini@kit.edu [Karlsruhe Institute of Technology (KIT) (Germany); Aiello, G.; Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Bachmann, C. [EUROfusion, PPPT, Garching (Germany); Barrett, T. [CCFE, Abingdon OX14 3DB (United Kingdom); Del Nevo, A. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Demange, D. [Karlsruhe Institute of Technology (KIT) (Germany); Forest, L. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Hernandez, F.; Norajitra, P. [Karlsruhe Institute of Technology (KIT) (Germany); Porempovic, G. [Fuziotech Engineering Ltd (Hungary); Rapisarda, D. [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Sardain, P. [CEA/IRFM, 13115 Saint-Paul-lès-Durance (France); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Vala, L. [Centrum výzkumu Řež, 250 68 Husinec-Řež (Czech Republic)

    2016-11-01

    Highlights: • Short description of the new Breeding Blanket Project in the EUROfusion consortium for the design of the EU PPPT DEMO: objectives. • Presentation of the design approach used in the development of the Breeding Blanket design: requirements. • Breeding Blanket design; in particular the four blanket concepts included in the study are presented, recent results highlighted and the status discussed. • Auxiliary systems and related R&D programme: in particular the work areas addressed in the Project (Tritium Technology, Pb-Li and Solid Breeders Technology, First Wall Design and R&D, Manufacturing) are presented, recent results highlighted and the status discussed. - Abstract: The design of a DEMO reactor requires the design of a blanket system suitable of reliable T production and heat extraction for electricity production. In the frame of the EUROfusion Consortium activities, the Breeding Blanket Project has been constituted in 2014 with the goal to develop concepts of Breeding Blankets for the EU PPPT DEMO; this includes an integrated design and R&D programme with the goal to select after 2020 concepts on fusion plants for the engineering phase. The design activities are presently focalized around a pool of solid and liquid breeder blanket with helium, water and PbLi cooling. Development of tritium extraction and control technology, as well manufacturing and development of solid and PbLi breeders are part of the programme.

  13. 18 CFR 284.303 - OCS blanket certificates.

    Science.gov (United States)

    2010-04-01

    ... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false OCS blanket certificates. 284.303 Section 284.303 Conservation of Power and Water Resources FEDERAL ENERGY REGULATORY... Pipelines on Behalf of Others § 284.303 OCS blanket certificates. Every OCS pipeline [as that term is...

  14. The TFTR lithium blanket module program

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  15. Neutronic design for the TFTR lithium blanket module

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  16. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1983-08-01

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

  17. Axial blanket enrichment optimization of the NPP Krsko fuel

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2001-01-01

    In this paper optimal axial blanket enrichment of the NPP Krsko fuel is investigated. Since the optimization is dictated by economic categories that can significantly vary in time, two step approach is applied. In the first step simple relationship between the equivalent change in enrichment of axial blankets and central fuel region is established. The relationship is afterwards processed with economic criteria and constraints to obtain optimal axial blanket enrichment. In the analysis realistic NPP Krsko conditions are considered. Except for the fuel enrichment all other fuel characteristics are the same as in the fuel used in the few most recent cycles. A typical reload cycle after the plant power uprate is examined. Analysis has shown that the current blanket enrichment is close to the optimal. Blanket enrichment reduction results in an approximately 100 000 US$ savings per fuel cycle.(author)

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  19. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

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

    1971-01-01

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

  20. Self-cooled liquid-metal blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Arheidt, K.; Barleon, L.

    1988-01-01

    A blanket concept for the Next European Torus (NET) where 83Pb-17Li serves both as breeder material and as coolant is described. The concept is based on the use of novel flow channel inserts for a decisive reduction of the magnetohydrodynamic (MHD) pressure drop and employs beryllium as neutron multiplier in order to avoid the need for breeding blankets at the inboard side of the torus. This study includes the design, neutronics, thermal hydraulics, stresses, MHDs, corrosion, tritium recovery, and safety of a self-cooled liquid-metal blanket. The results of the investigations indicate that the self-cooled blanket is an attractive alternative to other driver blanket concepts for NET and that it can be extrapolated to the conditions of a DEMO reactor

  1. Preheating of manure utilizing heat exchanger and flue gas. Forvarmning af gylle ved varmeveksling med roeggas

    Energy Technology Data Exchange (ETDEWEB)

    Weber, J.

    1987-07-15

    It has been shown that preheating of manures in biomass conversion plants to a temperature of 50-60 deg. C, before the anaerobic digestion takes place at a temperature of 35-45 deg. C, results in an increase of methane production. But the method normally involves an increase in energy consumption. The aim of the project was to develope methods of utilizing heat from flue gas emitted from the boiler connected to the plant, with the help of a heat exchanger. The heat thus recovered would be used to preheat the manure. The chosen method was to inject the flue gas directly into the manure mass, following this up with heat exchanging and condensing. In order to mix the flue gas thoroughly into the manure an ejector was used, this was driven by the manure flow. Results were satisfactory. (AB).

  2. A pre-heating method based on sinusoidal alternating current for lithium-ion battery

    Science.gov (United States)

    Fan, Wentao; Sun, Fengchun; Guo, Shanshan

    2018-04-01

    In this paper, a method of low temperature pre-heating of sinusoidal alternating current (SAC) is proposed. Generally, the lower the frequency of the AC current, the higher the heat generation rate. Yet at low frequency, there is a risk of lithium-ion deposition during the half cycle of charging. This study develops a temperature-adaptive, deposition-free AC pre-heating method. a equivalent electric circuit(EEC) model is established to predict the heat generation rate and temperature status, whose parameters are calibrated from the EIS impedance measurements. The effects of current frequency and amplitude on the heating effect are investigated respectively. A multistep temperature-adaptive amplitude strategy is proposed and the cell can be heated from -20°C to 5°C within 509s at 100Hz frequency with this method.

  3. Computational fluid dynamic on the temperature simulation of air preheat effect combustion in propane turbulent flame

    Science.gov (United States)

    Elwina; Yunardi; Bindar, Yazid

    2018-04-01

    this paper presents results obtained from the application of a computational fluid dynamics (CFD) code Fluent 6.3 to modelling of temperature in propane flames with and without air preheat. The study focuses to investigate the effect of air preheat temperature on the temperature of the flame. A standard k-ε model and Eddy Dissipation model are utilized to represent the flow field and combustion of the flame being investigated, respectively. The results of calculations are compared with experimental data of propane flame taken from literature. The results of the study show that a combination of the standard k-ε turbulence model and eddy dissipation model is capable of producing reasonable predictions of temperature, particularly in axial profile of all three flames. Both experimental works and numerical simulation showed that increasing the temperature of the combustion air significantly increases the flame temperature.

  4. Influence of the Previous Preheating Temperature on the Static Coefficient of Friction with Lubrication

    Directory of Open Access Journals (Sweden)

    M. Živković

    2016-12-01

    Full Text Available Experimental investigations static coefficient of friction in lubricated conditions and pre-heating of the sample pin at high temperatures is discussed in this paper. The static coefficient of friction was measured in the sliding steel copper pins per cylinder of polyvinylchloride. Pins are previously heated in a special chamber from room temperature to a temperature of 800 oC with a step of 50 °C. Tribological changes in the surface layer of the pins caused by pre-heating the pins at high temperatures and cooling systems have very significantly influenced the increase in the coefficient of static friction. The results indicate the possibility of improving the friction characteristics of metal materials based on their thermal treatment at elevated temperatures.

  5. Thermodynamic analysis and conceptual design for partial coal gasification air preheating coal-fired combined cycle

    Science.gov (United States)

    Xu, Yue; Wu, Yining; Deng, Shimin; Wei, Shirang

    2004-02-01

    The partial coal gasification air pre-heating coal-fired combined cycle (PGACC) is a cleaning coal power system, which integrates the coal gasification technology, circulating fluidized bed technology, and combined cycle technology. It has high efficiency and simple construction, and is a new selection of the cleaning coal power systems. A thermodynamic analysis of the PGACC is carried out. The effects of coal gasifying rate, pre-heating air temperature, and coal gas temperature on the performances of the power system are studied. In order to repower the power plant rated 100 MW by using the PGACC, a conceptual design is suggested. The computational results show that the PGACC is feasible for modernizing the old steam power plants and building the new cleaning power plants.

  6. Microshear bond strength of preheated silorane- and methacrylate-based composite resins to dentin.

    Science.gov (United States)

    Demirbuga, Sezer; Ucar, Faruk Izzet; Cayabatmaz, Muhammed; Zorba, Yahya Orcun; Cantekin, Kenan; Topçuoğlu, Hüseyin Sinan; Kilinc, Halil Ibrahim

    2016-01-01

    The aim of this study was to investigate the effect of preheating on microshear bond strength (MSBS) of silorane and methacrylate-based composite resins to human dentin. The teeth were randomly divided into three main groups: (1) composite resins were heated upto 68 °C; (2) cooled to 4 °C; and (3) control [room temperature (RT)]. Each group was then randomly subdivided into four subgroups according to adhesive system used [Solobond M (Voco), All Bond SE (Bisco), Clearfil SE Bond (CSE) (Kuraray), Silorane adhesive system (SAS) (3M ESPE)]. Resin composite cylinders were formed (0.9 mm diameter × 0.7 mm length) and MSBS of each specimen was tested. The preheated groups exhibited the highest MSBS (p composite resins may be an alternative way to increase the MSBS of composites on dentin. © Wiley Periodicals, Inc.

  7. Acid skim milk gels: The gelation process as affected by preheated pH

    NARCIS (Netherlands)

    Lakemond, C.M.M.; Vliet, van T.

    2008-01-01

    The effect of preheating milk (10 min 80 [degree sign]C) at pH values from 6.20 to 6.90 on formation of acid skim milk gels was studied by dynamic oscillation measurements. Up to pH 6.65 a higher pH of heating (pHheating) resulted in a higher G'. Since below pH 4.9 the development of

  8. Energetic, Exergetic, and Economic Analysis of MED-TVC Water Desalination Plant with and without Preheating

    Directory of Open Access Journals (Sweden)

    Nuri Eshoul

    2018-03-01

    Full Text Available Desalination is the sole proven technique that can provide the necessary fresh water in arid and semi-arid countries in sufficient quantities and meet the modern needs of a growing world population. Multi effect desalination with thermal vapour compression (MED-TVC is one of most common applications of thermal desalination technologies. The present paper presents a comprehensive thermodynamic model of a 24 million litres per day thermal desalination plant, using specialised software packages. The proposed model was validated against a real data set for a large-scale desalination plant, and showed good agreement. The performance of the MED-TVC unit was investigated using different loads, entrained vapour, seawater temperature, salinity and number of effects in two configurations. The first configuration was the MED-TVC unit without preheating system, and the second integrated the MED-TVC unit with a preheating system. The study confirmed that the thermo-compressor and its effects are the main sources of exergy destruction in these desalination plants, at about 40% and 35% respectively. The desalination plant performance with preheating mode performs well due to high feed water temperature leading to the production of more distillate water. The seawater salinity was proportional to the fuel exergy and minimum separation work. High seawater salinity results in high exergy efficiency, which is not the case with membrane technology. The plant performance of the proposed system was enhanced by using a large number of effects due to greater utilisation of energy input and higher generation level. From an economic perspective, both indicators show that using a preheating system is more economically attractive.

  9. Resistive vs. total power depositions by Alfven modes in pre-heated low aspect ratio tokamaks

    International Nuclear Information System (INIS)

    Cuperman, S.; Bruma, C.; Komoshvili, K.

    2004-01-01

    The power deposition of fast waves launched by a LFS located antenna in a pre-heated, strongly non-uniform low aspect ratio tokamak (START) is investigated. The rigorous computational results indicate a total power deposition by far larger than that predicted for Alfven continuum eigenmodes in cylindrical plasmas. For toroidal wave numbers |N| > 1, the resistive and total power depositions are almost equal. (author)

  10. Effect of preheating and light-curing unit on physicochemical properties of a bulk fill composite

    Directory of Open Access Journals (Sweden)

    Theobaldo JD

    2017-05-01

    Full Text Available Jéssica Dias Theobaldo,1 Flávio Henrique Baggio Aguiar,1 Núbia Inocencya Pavesi Pini,2 Débora Alves Nunes Leite Lima,1 Priscila Christiane Suzy Liporoni,3 Anderson Catelan3 1Department of Restorative Dentistry, Piracicaba Dental School, University of Campinas, Piracicaba, 2Ingá University Center, Maringá, 3Departament of Dentistry, University of Taubaté, Taubaté, Brazil Objective: The aim of this study is to evaluate the effect of composite preheating and polymerization mode on degree of conversion (DC, microhardness (KHN, plasticization (P, and depth of polymerization (DP of a bulk fill composite.Methods: Forty disc-shaped samples (n = 5 of a bulk fill composite were prepared (5 × 4 mm thick and randomly divided into 4 groups according to light-curing unit (quartz–tungsten–halogen [QTH] or light-emitting diode [LED] and preheating temperature (23 or 54 °C. A control group was prepared with a flowable composite at room temperature. DC was determined using a Fourier transform infrared spectrometer, KHN was measured with a Knoop indenter, P was evaluated by percentage reduction of hardness after 24 h of ethanol storage, and DP was obtained by bottom/top ratio. Data were statistically analyzed by analysis of variance and Tukey’s test (α = 0.05.Results: Regardless of light-curing, the highest preheating temperature increased DC compared to room temperature on bottom surface. LED showed a higher DC compared to QTH. Overall, DC was higher on top surface than bottom. KHN, P, and DP were not affected by curing mode and temperature, and flowable composite showed similar KHN, and lower DC and P, compared to bulk fill.Conclusion: Composite preheating increased the polymerization degree of 4-mm-increment bulk fill, but it led to a higher plasticization compared to the conventional flowable composite evaluated. Keywords: composite resins, physicochemical phenomena, polymerization, hardness, heating

  11. Formation of toroidal pre-heat plasma without residual magnetic field for high-beta pinch experiments

    International Nuclear Information System (INIS)

    Ikeda, Nagayasu; Tamaru, Ken; Nagata, Akiyoshi.

    1979-01-01

    Formation of toroidal pre-heat plasma was studied. The pre-heat plasma without residual magnetic field was made by chopping the current for pre-heat, A small toroidal-pinch system was used for the experiment. The magnetic field was measured with a magnetic probe. One turn loop was used for the measurement of the toroidal one-turn electric field. A pair of Rogoski coil was used for the measurement of plasma current. The dependence of residual magnetic field on chopping time was measured. By fast chopping of the primary current in the pre-heating circuit, the poloidal magnetic field was reduced to several percent within 5 microsecond. After chopping, no instability was observed in the principal discharge plasma produced within several microsecond. As the conclusion, it can be said that the control of residual field can be made by current chopping. (Kato, T.)

  12. Phase change material thermal storage for biofuel preheating in micro trigeneration application: A numerical study

    International Nuclear Information System (INIS)

    Wu, Dawei; Chen, Junlong; Roskilly, Anthony P.

    2015-01-01

    Highlights: • Engine exhaust heat driven phase change material thermal storage. • Fuel preheating for direct use of straight plant oil on diesel engine. • CFD aided design of the phase change material thermal storage. • Melting and solidification model considering natural convection. - Abstract: A biofuel micro trigeneration prototype has been developed to utilise local energy crop oils as fuel in rural areas and developing countries. Straight plant oils (SPOs) only leave behind very little carbon footprint during its simply production process compared to commercial biodiesels in refineries, but the high viscosity of SPOs causes difficulties at engine cold starts, which further results in poor fuel atomisation, compromised engine performance and fast engine deterioration. In this study, a phase change material (PCM) thermal storage is designed to recover and store engine exhaust heat to preheat SPOs at cold starts. High temperature commercial paraffin is selected as the PCM to meet the optimal preheating temperature range of 70–90 °C, in terms of the SPO property study. A numerical model of the PCM thermal storage is developed and validated by references. The PCM melting and solidification processes with the consideration of natural convection in liquid zone are simulated in ANSYS-FLUENT to verify the feasibility of the PCM thermal storage as a part of the self-contained biofuel micro trigeneration prototype

  13. Experimental investigation of laminar LPG-H{sub 2} jet diffusion flame with preheated reactants

    Energy Technology Data Exchange (ETDEWEB)

    D.P. Mishra; P. Kumar [Indian Institute of Technology, Kanpur (India). Combustion Laboratory, Department of Aerospace Engineering

    2008-10-15

    This paper presents an experimental investigation of the effect of H{sub 2} addition on flame length, soot free length fraction (SFLF), flame radiant fraction, gas temperature and emission level in LPG-H{sub 2} composite fuel jet diffusion flame for two preheated cases namely, (i) preheated air and (ii) preheated air and fuel. Results show that the H{sub 2} addition leads to a reduction in flame length which may be caused due to an increased gas temperature. Besides this, the flame length is also observed to be reduced with increasing reactants temperature. The soot free length fraction (SFLF) increases as H{sub 2} is added to fuel stream. This might have been caused by decrease in the C/H ratio in the flame and is favorable to attenuate PAH formation rate. Interestingly, the SFLF is observed to be reduced with increasing reactants temperature that may be due to reduction in induction period of soot formation caused by enhanced flame temperature. Moreover, the decreased radiant heat fraction with hydrogen addition is pertinent with the reduction in soot concentration level. The reduction in NOx emission level with H{sub 2} addition to the fuel stream is also observed. On the contrary, NOx emission level is found to be enhanced significantly with reactant temperature that can be attributed to the increase in thermal NOx through Zeldovich mechanism. 31 refs., 4 figs., 2 tabs.

  14. Delayed coking unit preheat train optimization; Otimizacao do preaquecimento das Unidades de Coque

    Energy Technology Data Exchange (ETDEWEB)

    Marins, Edson R; Geraldelli, Washington O; Barros, Francisco C [PETROBRAS, Rio de Janeiro, RJ (Brazil). Centro de Pesquisas (CENPES)

    2004-07-01

    The oil industry has been investing in research and development of new techniques and process improvements with the objective to increase the residual fraction profitability and to fulfill the market demands. The adequacy of the refining scheme has led to the development of bottom of the barrel processes that has the objective to convert heavy fractions into products of higher aggregate value. In this context, the process of Delayed Coking presents a great importance in the production of distillates in the diesel range as well as the processing of heavy residues, mostly in the markets where the fuel oil consumption is being reduced. With the approach to help PETROBRAS decide which route to follow during new designs of Delayed Coking units, this work presents a comparative study of the preheat train performance among the energy recovery to preheat the feed, in contrast with preheating the feed and generating steam, simultaneously. In this study the Pinch Technology methodology was used as a procedure for heat integration with the objective of getting the maximum energy recovery from the process, finding the best trade-off between operational cost and investment cost. The alternative of steam generation aims to provide an appropriate flexibility in Delayed Coking units design and operation. (author)

  15. Liquid-phase synthesis of vertically aligned carbon nanotubes and related nanomaterials on preheated alloy substrates

    Science.gov (United States)

    Yamagiwa, Kiyofumi

    2018-02-01

    Carbon nanotubes (CNTs) and related nanocarbons were selectively synthesized on commercially available alloy substrates by a simple liquid-phase technique. Fe- and Ni-rich stainless-steel (JIS SUS316L and Inconel®600, respectively) and Ni-Cu alloy (Monel®400) substrates were used for the synthesis, and each substrate was preheated in air to promote the self-formation of catalyst nanolayers on the surface. The substrates were resistance heated in ethanol without any addition of catalysts to grow CNTs. The yield of the CNTs effectively increased when the preheating process was employed. Highly aligned CNT arrays grew on the SUS316L substrate, while non-aligned CNTs and distinctive twisted fibers were observed on the other substrates. An Fe oxide layer was selectively formed on the preheated SUS316L substrate promoting the growth of the CNT arrays. Characterizations including cyclic voltammetry for the arrays revealed that the CNTs possess a comparatively defect-rich surface, which is a desirable characteristic for its application such as electrode materials for capacitors.

  16. Pre-HEAT: submillimeter site testing and astronomical spectra from Dome A, Antarctica

    Science.gov (United States)

    Kulesa, C. A.; Walker, C. K.; Schein, M.; Golish, D.; Tothill, N.; Siegel, P.; Weinreb, S.; Jones, G.; Bardin, J.; Jacobs, K.; Martin, C. L.; Storey, J.; Ashley, M.; Lawrence, J.; Luong-Van, D.; Everett, J.; Wang, L.; Feng, L.; Zhu, Z.; Yan, J.; Yang, J.; Zhang, X.-G.; Cui, X.; Yuan, X.; Hu, J.; Xu, Z.; Jiang, Z.; Yang, H.; Li, Y.; Sun, B.; Qin, W.; Shang, Z.

    2008-07-01

    Pre-HEAT is a 20 cm aperture submillimeter-wave telescope with a 660 GHz (450 micron) Schottky diode heterodyne receiver and digital FFT spectrometer for the Plateau Observatory (PLATO) developed by the University of New South Wales. In January 2008 it was deployed to Dome A, the summit of the Antarctic plateau, as part of a scientific traverse led by the Polar Research Institute of China and the Chinese Academy of Sciences. Dome A may be one of the best sites in the world for ground based Terahertz astronomy, based on the exceptionally cold, dry and stable conditions which prevail there. Pre-HEAT is measuring the 450 micron sky opacity at Dome A and mapping the Galactic Plane in the 13CO J=6-5 line, constituting the first submillimeter measurements from Dome A. It is field-testing many of the key technologies for its namesake -- a successor mission called HEAT: the High Elevation Antarctic Terahertz telescope. Exciting prospects for submillimeter astronomy from Dome A and the status of Pre-HEAT will be presented.

  17. Modelling of preheated regenerative chain in Cernavoda NPP using MMS calculation code

    International Nuclear Information System (INIS)

    Bigu, M.; Nita, I.; Prisecaru, I.; Dupleac, D.

    2005-01-01

    Full text: In this work it was studied operation of preheated regenerative chain from NPP Cernavoda. To obtain this analysis coupled analyses of condensate system, water supply system, and drain cooler system were effected. The analysis boundaries are: Upstream: - Steam condensers - Turbine Bleed Steam Downstream: - Steam Generators. The analysis was made in two steps: 1) Getting of hydraulic characteristic of pipe network from steam condensers to steam generators at nominal regime; this step was obtained with hydraulic package called PIPENET. 2) Real thermal hydraulic analyses were done based on hydraulic characteristic of pipe network and supplementary data required for heat transfer calculation in equipment of preheated regenerative chain. Thermal analyses were done using MMS package and refered to normal operating regimes, namely, nominal operating regime required for calibration of calculating model, shutdown regime, start-up regime from zero power hot to nominal power and to abnormal operating regimes, namely, turbine trip, reactor trip and loss of two condensate pumps. The results were compared with already existing analysis and showed the largest differences at interface areas (i.e. 5%). This led us to idea of extending analysis to all secondary circuits in order to reduce the number of boundary conditions which can generate uncertainty in analysis. In this analysis we obtained an advanced model of preheated regenerative chain of secondary circuit in Cernavoda NPP which could be extended up to cover the whole secondary circuit by including the analysis of steam generators, turbine, and steam condenser. (authors)

  18. Modelling of preheated regenerative chain in Cernavoda NPP using MMS calculation code

    International Nuclear Information System (INIS)

    Bigu, M.; Nita, I.; Prisecaru, I.; Dupleac, D.

    2005-01-01

    In this work it was studied operation of preheated regenerative chain from NPP Cernavoda. To obtain this analysis coupled analyses of condensate system, water supply system, and drain cooler system were effected. The analysis boundaries are: Upstream: - Steam condensers - Turbine Bleed Steam Downstream: - Steam Generators. The analysis was made in two steps: 1) Getting of hydraulic characteristic of pipe network from steam condensers to steam generators at nominal regime; this step was obtained with hydraulic package called PIPENET. 2) Real thermal hydraulic analyses were done based on hydraulic characteristic of pipe network and supplementary data required for heat transfer calculation in equipment of preheated regenerative chain. Thermal analyses were done using MMS package and referred to normal operating regimes, namely, nominal operating regime required for calibration of calculating model, shutdown regime, start-up regime from zero power hot to nominal power and to abnormal operating regimes, namely, turbine trip, reactor trip and loss of two condensate pumps. The results were compared with already existing analysis and showed the largest differences at interface areas (i.e. 5%). This led US to idea of extending analysis to all secondary circuits in order to reduce the number of boundary conditions which can generate uncertainty in analysis. In this analysis we obtained an advanced model of preheated regenerative chain of secondary circuit in Cernavoda NPP which could be extended up to cover the whole secondary circuit by including the analysis of steam generators, turbine, and steam condenser. (authors)

  19. Plasma formation and target preheating by prepulse of PW laser light

    Science.gov (United States)

    Sentoku, Yasuhiko; Iwata, Natsumi; Koga, James; Dover, Nicholas; Nishiuchi, Mamiko

    2017-10-01

    An intense short pulse laser with intensity over 1021 W/cm2 has become available, i.e. J-KAREN-P at QST. Although the contrast of the short pulse is improved to be of the order of 10-11, there is an unavoidable prepulse, which has multiple spikes (ps) on top of an exponential profile with intensity greater than 1014 W/cm2 about 50 ps in front of the main pulse. The prepulse preheats the target and also produces tenuous plasmas in front of a target before the main pulse arrives. It is critical to understand such preheating of the target, where the nonlocal heat transport is essential at intensity >1014 W/cm2, since the target condition might totally change before the interaction with the main pulse. Using a hydro code, FLASH, and a collisional particle-in-cell code, PICLS, we study the preplasma formation and target preheating over tens of picoseconds timescale, and discuss the prepulse effects on the main pulse interaction. Work supported by the JSPS KAKENHI under Grant No. JP15K21767.

  20. Plan for radionuclide tracer studies of the residence time distribution in the Wilsonville dissolver and preheater

    International Nuclear Information System (INIS)

    Jolley, R.L.; Begovich, J.M.; Brashear, H.R.

    1983-12-01

    Stimulus-response measurements using radiotracers to measure residence time distribution (RTD) and hydrodynamic parameters for the preheaters and dissolvers at the Ft. Lewis Solvent Refined Coal (SRC) and the Exxon Donor Solvent (EDS) coal conversion pilot plants are reviewed. A plan is also presented for a series of radioactive tracer studies proposed for the Advanced Coal Liquefaction Facility at Wilsonville, Alabama, to measure the RTD for the preheater and dissolvers in the SRC-I mode. The tracer for the gas phase will be 133 Xe, and 198 Au (on carbonized resin or as an aqueous colloidal suspension) will be used as the slurry tracer. Four experimental phases are recommended for the RTD tracer studies: (1) preheater; (2) dissolver with 100% takeoff; (3) dissolver with 100% takeoff and solids withdrawal; and (4) dissolver with 50% takeoff. Eighteen gas-tracer and 22 liquid-tracer injections are projected to accomplish the four experimental phases. Two to four tracer injections are projected for preliminary tests to ensure the capability of safe injection of the radiotracers and the collection of statistically significant data. A complete projected cost and time schedule is provided, including procurement of necessary components, preparation of the radiotracers, assembly and testing of tracer injection apparatus and detection systems, onsite work and tracer injections, laboratory experimentation, data analysis, and report writing

  1. Tekken tests in a steel 'ASTM A 514 GR B' to determine the preheating temperature

    International Nuclear Information System (INIS)

    Quesada, Hector Juan; Zalazar, Monica; Asta, Eduardo Pablo

    2004-01-01

    Cold fissure tests are used to determine the proper preheating temperature in order to prevent fissures during the steel welding process. Tekken tests were carried out on a quenched and tempered high resistance 25.4 mm thick steel (ASTM A514 Gr.B) used in structural applications. The welding was carried out using a FCAW semiautomatic process with gas protection and low hydrogen tubular electrode E110T5-K4. Similar parameters and splicing design were later applied in production. The microstructures of the base material and the welding were determined by optic and electron microscopy. The thermal cycles of the welding were recorded in order to relate the preheating temperature with the cooling time from 800 o C - 500 o C (t 8/5 ) and from 800 o C - 100 o C (tg/1) and the presence or not of fissures. Preheating at 150 o C and t 8/5 greater than 17 s was found to guarantee fissure free welding (CW)

  2. Tekken testing to determine the preheating temperature on ASTM A514 GR B steel

    International Nuclear Information System (INIS)

    Asta, Eduardo; Zalazar, Monica; Quesada, Hector

    2003-01-01

    The cold cracking test methods are used to determine the preheating temperature in order to avoid cracking in steel welding.In this work Tekken tests on high strength quenching and tempering (ASTM A514 GrB) structural steel with a thickness of 25 mm have been made.The welds were done using a FCAW process with gas shielding and basic low hydrogen cored wire E 110T5-K4.The welding parameters and joint design applied in this work are similar to the ones used on site production.The base metal, HAZ and weld metal microstructure have been evaluated by optical and SEM microscopy.Thermal cycles records of each welding have been made to relate preheat temperature with the cooling time on the range of 800-500 degC (t8/5) or 800-100degC (t8/1) and the evidence of crack or no crack condition.Finally, a preheat temperature of 150degC and the cooling time larger than 17 s improve a welding integrity without cracks

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

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    1995-01-01

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

  5. Is a Blanket Elective Single Embryo Transfer Policy Defensible?

    Directory of Open Access Journals (Sweden)

    Eli Y. Adashi

    2017-04-01

    Full Text Available For the purpose of reducing maternal and neonatal morbidity, elective single transfer (eSET in in vitro fertilization (IVF was first proposed in 1999. The purpose of this review is to summarize recent oral debate between a proponent and an opponent of expanded eSET utilization in an attempt to determine whether a blanket eSET policy, as is increasingly considered, is defensible. While eSET is preferable when possible, and agreed upon by provider and patient, selective double embryo transfer (DET must be seriously entertained if deemed more appropriate or is desired by the patient. Patient autonomy, let alone prolonged infertility and advancing age, demand nothing less. Importantly, IVF-generated twins represent only 15.7% of the national twin birth rate in the United States. Non-IVF fertility treatments have been identified as the main cause of all multiple births for quite some time. However, educational and regulatory efforts over the last decade, paradoxically, have exclusively only been directed at the practice of IVF, although IVF patient populations are rapidly aging. It is difficult to understand why non-IVF fertility treatments, usually applied to younger women, have so far escaped attention. This debate on eSET utilization in association with IVF may contribute to a redirection of priorities.

  6. Research on a Household Dual Heat Source Heat Pump Water Heater with Preheater Based on ASPEN PLUS

    Directory of Open Access Journals (Sweden)

    Xiang Gou

    2016-12-01

    Full Text Available This article proposes a dual heat source heat pump bathroom unit with preheater which is feasible for a single family. The system effectively integrates the air source heat pump (ASHP and wastewater source heat pump (WSHP technologies, and incorporates a preheater to recover shower wastewater heat and thus improve the total coefficient of performance (COP of the system, and it has no electric auxiliary heating device, which is favorable to improve the security of the system operation. The process simulation software ASPEN PLUS, widely used in the design and optimization of thermodynamic systems, was used to simulate various cases of system use and to analyze the impact of the preheater on the system. The average COP value of a system with preheater is 6.588 and without preheater it is 4.677. Based on the optimization and analysis, under the standard conditions of air at 25 °C, relative humidity of 70%, wastewater at 35 °C, wastewater flow rate of 0.07 kg/s, tap water at 15 °C, and condenser outlet water temperature at 50 °C, the theoretical COP of the system can reach 9.784 at an evaporating temperature of 14.96 °C, condensing temperature of 48.74 °C, and preheated water temperature of 27.19 °C.

  7. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1983-06-01

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

  8. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1984-02-01

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

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

    International Nuclear Information System (INIS)

    Ishitsuka, E.

    2002-01-01

    Advanced solid breeding blanket design in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high dose of neutron irradiation. Therefore, the development of such advanced blanket materials is indispensable. In this paper, the cooperation activities among JAERI, universities and industries in Japan on the development of these advanced materials are reported. Advanced tritium breeding material to prevent the grain growth in high temperature had to be developed because the tritium release behavior degraded by the grain growth. As one of such materials, TiO 2 -doped Li 2 TiO 3 has been studied, and TiO 2 -doped Li 2 TiO 3 pebbles was successfully fabricated. For the advanced neutron multiplier, the beryllium intermetallic compounds that have high melting point and good chemical stability have been studied. Some characterization of Be 12 Ti was studied. The pebble fabrication study for Be 12 Ti was also performed and Be 12 Ti pebbles were successfully fabricated. From these activities, the bright prospect to realize the DEMO blanket by the application of TiO 2 -doped Li 2 TiO 3 and beryllium intermetallic compounds was obtained. (author)

  10. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-12-01

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

  11. Heat transfer problems in gas-cooled solid blankets

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  12. Neutronics analysis for aqueous self-cooled fusion reactor blankets

    International Nuclear Information System (INIS)

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

    1986-06-01

    The tritium breeding performance of several Aqueous Self-Cooled Blanket (ASCB) configurations for fusion reactors has been evaluated. The ASCB concept employs small amounts of lithium compound dissolved in light or heavy water to serve as both coolant and breeding medium. The inherent simplicity of this concept allows the development of blankets with minimal technological risk. The tritium breeding performance of the ASCB concept is a critical issue for this family of blankets. Contrary to conventional blanket designs there will be a significant contribution to the tritium breeding ratio (TBR) in the water coolant/breeder of duct shields, and the 3-D TBR will therefore be similar to the 1-D TBR. The tritium breeding performance of an ASCB for a MARS-like-tandem reactor and an ASCB based breeding-shield for the Next European Torus (NET) are assessed. Two design options for the MARS-like blanket are discussed. One design employs a vanadium first wall, and zircaloy for the structural material. The trade-offs between light water and heavy water cooling options for this zircaloy blanket are discussed. The second design option for MARS relies on the use of a vanadium alloy as the stuctural material, and heavy water as the coolant. It is demonstrated that both design options lead to low-activation blankets that allow class C burial. The breeder-shield for NET consists of a water-cooled stainless steel shield

  13. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

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

    1992-06-01

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

  14. Overview of EU activities on DEMO liquid metal breeder blanket

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Malang, S.; Reimann, J.; Perujo, A.

    1994-01-01

    The present paper gives an overview of both design and experimental activities within the European Union (EU) concerning the development of liquid metal breeder blankets for DEMO. After several years of studies on breeding blankets, two blanket concepts are presently considered, both using the eutectic Pb-17Li: the dual-coolant concept and the water-cooled concept. The analysis of such concepts has permitted to identify the experimental areas where further data are required. Tritium control and MHD-issues are, at present, the activities on which is devoted the greatest effort within the EU. (authors). 4 figs., 4 tabs., 39 refs

  15. Availability analysis of the ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

    2015-01-01

    The ITER blanket remote handling system (BRHS) is required to replace 440 blanket first wall panels in a two-year maintenance period. To investigate this capability, an availability analysis of the system was carried out. Following the analysis procedure defined by the ITER organization, the availability analysis consists of a functional analysis and a reliability block diagram analysis. In addition, three measures to improve availability were implemented: procurement of spare parts, in-vessel replacement of cameras, and simultaneous replacement of umbilical cables. The availability analysis confirmed those measures improve the availability and capability of the BRHS to replace 440 blanket first wall panels in two years. (author)

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

    International Nuclear Information System (INIS)

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

    1985-02-01

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

  17. Conceptual design of blanket structures for fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-03-01

    Conceptual design study for in-vessel components including tritium breeding blanket of FER has been carried out. The objective of this study is to obtain the engineering and technological data for selecting the reactor concept and for its construction by investigating fully and broadly. The design work covers in-vessel components (such as tritium breeding blanket, first wall, shield, divertor and blanket test module), remote handling system and tritium system. The designs of those components and systems are accomplished in consideration of their accomodation to whole reactor system and problems for furthur study are clarified. (author)

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

    International Nuclear Information System (INIS)

    1978-01-01

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

  19. A Li-particulate blanket concept for ITER

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Werner, R.W.

    1977-01-01

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

  1. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Grief, Andrew; Merrill, Brad J.; Humrickhouse, Paul; Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon; Poitevin, Yves; Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard

    2016-01-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  2. Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Grief, Andrew [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Merrill, Brad J.; Humrickhouse, Paul [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID (United States); Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom)

    2016-11-01

    Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

  3. The effect of preheated tendon as a lean meat replacement on the properties of fine emulsion sausages.

    Science.gov (United States)

    Sadler, D H; Young, O A

    1993-01-01

    Tendon from beef hind leg muscles was used to replace some of the lean in a conventional emulsion formulation. The tendon was homogenized and either used raw or preheated for 2·5 h at a range of temperatures (50, 60, 70, 80°C) before use. Texture analysis and sensory evaluation were performed on cylinders of cooked sausage. Texture analysis was carried out on formulations which had 20% of meat protein replaced by 20% tendons which were raw or had been preheated to 50, 60, 70, or 80°C. Fracturability decreased by about 40% with raw tendon, but was restored to within 20% of the no-replacement control if the tendon had been preheated. Hardness was approximately doubled by replacement with raw tendon or tendon heated at 50°C. At temperatures higher than that, hardness returned to approximately no-replacement levels. For sensory evaluation (0-25% replacement; preheating at 70°C), sausages were assessed by a 12-member panel for texture, flavour and overall acceptability. All attributes decreased with increasing collagen content, the decrease being less marked with preheated tendon. Thus more connective tissue could be added for the same panel score if the tissue was preheated. Comparison of the texture profile and the panel scores for texture at the same lean replacement level suggested that reduced fracturability was the texture parameter that panellists objected to when heated tendon replaced some of the lean. Other researchers have shown that connective tissue preheated to 100°C before addition in emulsion sausages results in improved yields and better sensory attributes, but the present results show that temperatures as low as 60°C can be effective for beef tendon. Copyright © 1993. Published by Elsevier Ltd.

  4. Optimization of up-flow anaerobic sludge blanket reactor for ...

    African Journals Online (AJOL)

    Optimization of up-flow anaerobic sludge blanket reactor for treatment of composite ... AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... Granules grown in the bottom part of UASB reactor were more compact and tense ...

  5. Electromagnetic effects involving a tokamak reactor first wall and blanket

    International Nuclear Information System (INIS)

    Turner, L.R.; Evans, K. Jr.; Gelbard, E.; Prater, R.

    1980-01-01

    Four electromagnetic effects experienced by the first wall and blanket of a tokamak reactor are considered. First, the first wall provides reduction of the growth rate of vertical axisymmetric instability and stabilization of low mode number interval kink modes. Second, if a rapid plasma disruption occurs, a current will be induced on the first wall, tending to maintain the field formerly produced by the plasma. Third, correction of plasma movement can begin on a time scale much faster than the L/R time of the first wall and blanket. Fourth, field changes, especially those from plasma disruption or from rapid discharge of a toroidal field coil, can cause substantial eddy current forces on elements of the first wall and blanket. These effects are considered specifically for the first wall and blanket of the STARFIRE commercial reactor design study

  6. Blanket options for high-efficiency fusion power

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  7. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  8. Fusion blanket for high-efficiency power cycles

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  9. Fusion blankets for high-efficiency power cycles

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-11-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  12. Aqueous self-cooled blanket concepts for fusion reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Dudziak, D.J.

    1977-09-01

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

  15. Application of vanadium alloys to a fusion reactor blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-05-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1983-01-01

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

  18. Evaluation of organic moderator/coolants for fusion breeder blankets

    International Nuclear Information System (INIS)

    Romero, J.B.

    1980-03-01

    Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process

  19. Main features and potentialities of gas-blanket systems

    International Nuclear Information System (INIS)

    Lehnert, B.

    1977-02-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  2. Overview of first wall/blanket/shield technology

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1983-04-01

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

  3. Applications of the Aqueous Self-Cooled Blanket concept

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  4. Progress on DEMO blanket attachment concept with keys and pins

    International Nuclear Information System (INIS)

    Vizvary, Zsolt; Iglesias, Daniel; Cooper, David; Crowe, Robert; Riccardo, Valeria

    2015-01-01

    Highlights: • DEMO blanket attachment system with keys and pins (without using bolts). • Blanket segments are preloaded by progressively designed springs. • Blanket back plate flexibility has a major impact on spring design. • Mechanical analysis of other components indicates no unresolvable issues. • Thermal analysis indicates acceptable temperatures for the support system. - Abstract: The blanket attachment has to cope with gravity, thermal and electromagnetic loads, also it has to be installed and serviced by remote handling. Pre-stressed components suffer from stress relaxation in irradiated environments such as DEMO. To circumvent this problem pre-stressed component should be either avoided or shielded, and where possible keys and pins should be used. This strategy has been proposed for the DEMO multi-module segments (MMS). The blanket segments are held by two tapered keys each, designed to allow thermal expansions while providing contact with the vacuum vessel and to resist the poloidal and radial moments the latter being dominant at 9.1 MNm inboard and 15 MNm outboard. On the top of the blanket segment there is a pin which provides vertical support. At the bottom another vertical support has to lock them in position after installation and manage the pre-load on the segments. The pre-load is required to deal with the electromagnetic loads during disruption. This is provided by a set of springs, which require shielding as they are preloaded. These are sized to cope with the force (3 MN inboard, 1.4 MN outboard) due to halo currents and the toroidal moment which can reverse. Calculations show that the flexibility of the blanket segment itself plays a significant role in defining the required support system. The blanket segment acts as a preloaded spring and it has to be part of the attachment design as well.

  5. Electrical connectors for blanket modules in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Poddubnyi, I., E-mail: poddubnyyii@nikiet.ru [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Khomiakov, S.; Kolganov, V. [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Sadakov, S.; Calcagno, B.; Chappuis, Ph.; Roccella, R.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Danilov, I.; Leshukov, A.; Strebkov, Y. [Open Joint-Stock Company “N.A. Dollezhal Research and Development Institute of Power Engineering”, 107140, Malaya Krasnoselskaya Street 2/8, Moscow (Russian Federation); Ulrickson, M. [Sandia National Laboratories MS-1129, PO Box 5800, Albuquerque, NM 87185 (United States)

    2014-10-15

    Highlights: • Analysis of static and cyclic strength for L-shaped and Z-shaped ES has been performed. • Analysis results do show that for L-shaped ES static and cyclic strength criteria are not satisfied. • Static and cyclic strength criteria are met well by ES with Z-shaped elastic elements. • ES with Z-shaped elastic elements has been adopted as a new baseline design for ITER. - Abstract: Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2 mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137 kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9 T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak. The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was

  6. THE INFLUENCE OF PRE-HEAT TREATMENT ON WHITE CAST IRONS PLASTICITY

    Directory of Open Access Journals (Sweden)

    T. M. Myronova

    2013-11-01

    Full Text Available Purpose. The development of heat treatment modes of white cast irons for structure changes in their eutectic constituent, namely in disturbing the monolithic structure of ledeburite colonies cementite structure and eutectic net continuity. Also the mentioned heat treatment modes are targeted to the eutectic net shift for the most suitable position from the point of plastic deforming. Methodology. The hypoeutectic white cast irons with 2.92…3.35 % carbon content and additionally alloyed by 3.18 % vanadium have been used as the research materials. The mentioned alloys have been pre-heat treated and hot twist tested. Findings. The research results showed that the carbide net breaking by plastic deforming leads to cast irons mechanical properties increasing but has difficulties in implementation due to the white cast irons low plasticity. The influence of different pre-heat treatment modes on structure and plasticity of white hypoeutectic cast irons have been investigated. They include the isotherm soaking under the different temperatures as well as multiply soakings and thermo-cycling. The influence of eutectic level, as well as pre heat treatment modes on different composition white cast irons hot plasticity have been investigated. Originality. It was determined that the heat treatment, which leads to double α→γ recrystallization under 860 – 950 °С and reperlitization under 720-680 °С results in significant increase of plasticity, as well as in un-alloyed and alloyed by vanadium white cast irons. It takes place due to carbide matrix phase separation in ledeburite colonies by new phase boundaries forming especially due to carbide transformations under vanadium alloying. Practical value. The implementation of pre-heat treatment with phase recrystallization resulted in hypoeutectic white cast irons plasticity increasing. The obtained level of cast iron plasticity corresponds to the one of carbide class steels, which ensures the successful

  7. Thermal energy analysis of a lime production process: Rotary kiln, preheater and cooler

    International Nuclear Information System (INIS)

    Shahin, Hamed; Hassanpour, Saeid; Saboonchi, Ahmad

    2016-01-01

    Highlights: • The integrated model for lime production unit which includes cooler, preheater and rotary kiln is developed. • The effect of residence time in each section on efficiency is investigated. • Influence of material feed rate and excess air on specific fuel consumption is analyzed. • The significant effect of particle size on efficiency and specific fuel consumption is shown. - Abstract: In this paper, thermal energy analysis of three zones of a lime production process, which are preheater, rotary kiln and cooler, is performed. In order to perform a proper quantitative estimation, the system was modeled using energy balance equations including coupled heat transfer and chemical reaction mechanisms. A mathematical model was developed, and consequently, the thermal and chemical behavior of limestone was investigated. The model was verified using empirical data. After model confirmation, the variation of Specific Fuel Consumption (SFC) versus production rate was predicted and the optimum condition was determined. Subsequently, fuel consumption was calculated regarding to altered residence time inside each zone of lime production process, for a constant output. Results indicate that increasing the residence time inside each zone of lime production process, will enhance thermal efficiency and saves fuel consumption. Relative enhancement will be the same for different sizes of limestone. It was found that a 10-min increase in material residence time inside the preheater or rotary kiln can reduce fuel consumption by around two percent. Whereas, a 5-min increase in material residence time inside the cooler would be enough to obtain a similar result. Finally, the ratio of air-to-fuel and production rate are changed in such a way that the same product is achieved. The model predicts that lowering excess air from 15% to 10% leads to a 2.5% reduction of Specific Fuel Consumption (SFC).

  8. Minimizing scatter-losses during pre-heat for magneto-inertial fusion targets

    Science.gov (United States)

    Geissel, Matthias; Harvey-Thompson, Adam J.; Awe, Thomas J.; Bliss, David E.; Glinsky, Michael E.; Gomez, Matthew R.; Harding, Eric; Hansen, Stephanie B.; Jennings, Christopher; Kimmel, Mark W.; Knapp, Patrick; Lewis, Sean M.; Peterson, Kyle; Schollmeier, Marius; Schwarz, Jens; Shores, Jonathon E.; Slutz, Stephen A.; Sinars, Daniel B.; Smith, Ian C.; Speas, C. Shane; Vesey, Roger A.; Weis, Matthew R.; Porter, John L.

    2018-02-01

    The size, temporal and spatial shape, and energy content of a laser pulse for the pre-heat phase of magneto-inertial fusion affect the ability to penetrate the window of the laser-entrance-hole and to heat the fuel behind it. High laser intensities and dense targets are subject to laser-plasma-instabilities (LPI), which can lead to an effective loss of pre-heat energy or to pronounced heating of areas that should stay unexposed. While this problem has been the subject of many studies over the last decades, the investigated parameters were typically geared towards traditional laser driven Inertial Confinement Fusion (ICF) with densities either at 10% and above or at 1% and below the laser's critical density, electron temperatures of 3-5 keV, and laser powers near (or in excess of) 1 × 1015 W/cm2. In contrast, Magnetized Liner Inertial Fusion (MagLIF) [Slutz et al., Phys. Plasmas 17, 056303 (2010) and Slutz and Vesey, Phys. Rev. Lett. 108, 025003 (2012)] currently operates at 5% of the laser's critical density using much thicker windows (1.5-3.5 μm) than the sub-micron thick windows of traditional ICF hohlraum targets. This article describes the Pecos target area at Sandia National Laboratories using the Z-Beamlet Laser Facility [Rambo et al., Appl. Opt. 44(12), 2421 (2005)] as a platform to study laser induced pre-heat for magneto-inertial fusion targets, and the related progress for Sandia's MagLIF program. Forward and backward scattered light were measured and minimized at larger spatial scales with lower densities, temperatures, and powers compared to LPI studies available in literature.

  9. Preheating ablation effects on the Rayleigh-Taylor instability in the weakly nonlinear regime

    International Nuclear Information System (INIS)

    Wang, L. F.; Ye, W. H.; He, X. T.; Sheng, Z. M.; Don, Wai-Sun; Li, Y. J.

    2010-01-01

    The two-dimensional Rayleigh-Taylor instability (RTI) with and without thermal conduction is investigated by numerical simulation in the weakly nonlinear regime. A preheat model κ(T)=κ SH [1+f(T)] is introduced for the thermal conduction [W. H. Ye, W. Y. Zhang, and X. T. He, Phys. Rev. E 65, 057401 (2002)], where κ SH is the Spitzer-Haerm electron thermal conductivity coefficient and f(T) models the preheating tongue effect in the cold plasma ahead of the ablation front. The preheating ablation effects on the RTI are studied by comparing the RTI with and without thermal conduction with identical density profile relevant to inertial confinement fusion experiments. It is found that the ablation effects strongly influence the mode coupling process, especially with short perturbation wavelength. Overall, the ablation effects stabilize the RTI. First, the linear growth rate is reduced, especially for short perturbation wavelengths and a cutoff wavelength is observed in simulations. Second, the second harmonic generation is reduced for short perturbation wavelengths. Third, the third-order negative feedback to the fundamental mode is strengthened, which plays a stabilization role. Finally, on the contrary, the ablation effects increase the generation of the third harmonic when the perturbation wavelengths are long. Our simulation results indicate that, in the weakly nonlinear regime, the ablation effects are weakened as the perturbation wavelength is increased. Numerical results obtained are in general agreement with the recent weakly nonlinear theories as proposed in [J. Sanz, J. Ramirez, R. Ramis et al., Phys. Rev. Lett. 89, 195002 (2002); J. Garnier, P.-A. Raviart, C. Cherfils-Clerouin et al., Phys. Rev. Lett. 90, 185003 (2003)].

  10. Design and development of ceramic breeder demo blanket

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Science.gov (United States)

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

    2017-12-01

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

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

    International Nuclear Information System (INIS)

    Sako, K.; Minato, A.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Tanaka, Satoru

    1987-01-01

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

  14. Damage to Preheated Tungsten Targets after Multiple Plasma Impacts Simulating ITER ELMs

    Energy Technology Data Exchange (ETDEWEB)

    Garkusha, I.E.; Bandura, A.N.; Byrka, O.V.; Chebotarev, V.V.; Makhlay, V.A.; Tereshin, V.I. [Kharkov Inst. of Physics and Technology, Inst. of Plasma Physics of National Science Center, Akademicheskaya street, 1, 61108 Kharkov (Ukraine); Landman, I.; Pestchanyi, S. [FZK-Forschungszentrum Karlsruhe, Association Euratom-FZK, Technik und Umwelt, Postfach 3640, D-7602 1 Karlsruhe (Germany)

    2007-07-01

    Full text of publication follows: The energy loads onto ITER divertor surfaces associated with the Type I ELMs are expected to be up to 1 MJ/m{sup 2} during 0.1-0.5 ms, with the number of pulses about 103 per discharge. Tungsten is a candidate material for major part of the surface, but its brittleness can result in substantial macroscopic erosion after the repetitive heat loads. To minimize the brittle destruction, tungsten may be preheated above the ductile-to-brittle transition temperature. In this work the behavior of preheated tungsten targets under repetitive ELM-like plasma pulses is studied in simulation experiments with the quasi-stationary plasma accelerator QSPA Kh-50. The targets have been exposed up to 450 pulses of the duration 0.25 ms and the heat loads either 0.45 MJ/m{sup 2} or 0.75 MJ/m{sup 2}, which is respectively below and above the melting threshold. During the exposures the targets were permanently kept preheated at 650 deg. C by a heater at target backside. In the course of exposures the irradiated surfaces were examined after regular numbers of pulses using the SEM and the optical microscopy. The profilometry, XRD, microhardness and weight loss measurements have been performed, as well as comparisons of surface damages after the heat loads both below and above the melting threshold. It is obtained that macro-cracks do not develop on the preheated surface. After the impacts with surface melting, a fine mesh of intergranular microcracks has appeared. The width of fine intergranular cracks grows with pulse number, achieving 1-1.5 microns after 100 pulses, and after 210 pulses the crack width increases up to 20 microns, which is comparable with grain sizes. Threshold changes in surface morphology resulting in corrugation structures and pits on the surface as well as importance of surface tension in resulted 'micro-brush' structures are discussed. Further evolution of the surface pattern is caused by loss of separated grains on exposed

  15. Experimental and analytical evaluation of preheating temperature during multipass repair welding

    Directory of Open Access Journals (Sweden)

    Sedmak Aleksandar S.

    2017-01-01

    Full Text Available Experimental measurement and analytical calculation of preheating, i. e. interpass temperature during multi-pass repair welding has been presented. Analytical calculation is based on heat transfer analysis, whereas measurements have been performed by thermovision camera. Repair welding was performed on crane wheels in the Steelworks Smederevo. Comparison of results indicated that analytical calculation is good enough as the first approximation, but it needs further elaboration, e. g. taking into account the radiation component of heat dissipation and/or temperature dependence of material thermomechanical properties.

  16. Pretreatment and preheating of scrap. Tarkastelu koskien romun esikaesittely- ja esikuumennusmenetelmiae

    Energy Technology Data Exchange (ETDEWEB)

    Hooli, P.; Hanni, J. (Outokumpu Oy Tornion Tehtaat, Tornio (Finland))

    1990-01-01

    As a background for this study has been those demands for scrap treatments and transportation, which are coming with increasing production of melting shop of Outokumpu Oy's Tornio works and also problems caused by snow among productionrate. Different pretreatment-, transport-, and preheatingmethods and some alternatives has been studied to arrange those as a functioning complete. Also very exact plannings for some pretreatmentmethods has been made. From preheatingmethods some methods, which are concerned to be effective and possible in the future has been studied. In addition those parameters, which are involved to the effectivity of preheating process in melting shop of Outokumpu Oy's Tornio works has been examined.

  17. Pretreatment and preheating of scrap; Tarkastelu koskien romun esikaesittely- ja esikuumennusmenetelmiae

    Energy Technology Data Exchange (ETDEWEB)

    Hooli, P.; Hanni, J. [Outokumpu Oy Tornion Tehtaat, Tornio (Finland)

    1990-12-31

    As a background for this study has been those demands for scrap treatments and transportation, which are coming with increasing production of melting shop of Outokumpu Oy`s Tornio works and also problems caused by snow among productionrate. Different pretreatment-, transport-, and preheatingmethods and some alternatives has been studied to arrange those as a functioning complete. Also very exact plannings for some pretreatmentmethods has been made. From preheatingmethods some methods, which are concerned to be effective and possible in the future has been studied. In addition those parameters, which are involved to the effectivity of preheating process in melting shop of Outokumpu Oy`s Tornio works has been examined.

  18. Transmutation blanket design for a Tokamak system

    International Nuclear Information System (INIS)

    Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria A. Fortini; Costa, Antonella L.

    2011-01-01

    Sub-critical advanced reactor with a D-T fusion neutron source based on Tokamak technology is an innovative type of nuclear system. Due to the high quantity of neutrons produced by fusion reactions, it could be well spent in the transmutation process of the transuranic elements. Nevertheless, to achieve a successful transmutation, it is necessary to know the neutron fluence along the radial axis and its characteristics. In this work, it evaluated the neutron flux and interaction frequency along the radial axis changing the material of the first wall. W-alloy, beryllium and the combination of both were studied and regions more suitable to transmutation were determined. The results demonstrated that the better zone to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements W-alloy/W-alloy and W-alloy/Beryllium would be able to hold the requirements of high fluence and hardening spectrum needed to transuranic transmutation. The system was simulated using the MCNP5 code, the ITER Final Design Report, 2001, and the FENDL/MC-2.1 nuclear data library. (author)

  19. A simple method to prevent hard X-ray-induced preheating effects inside the cone tip in indirect-drive fast ignition implosions

    International Nuclear Information System (INIS)

    Liu, Dongxiao; Shan, Lianqiang; Zhou, Weimin; Wu, Yuchi; Zhu, Bin; Zhang, Feng; Bi, Bi; Zhang, Bo; Zhang, Zhimeng; Shui, Min; He, Yingling; Gu, Yuqiu; Zhang, Baohan; Peng, Xiaoshi; Xu, Tao; Wang, Feng; Yang, Zhiwen; Chen, Tao; Chen, Li; Chen, Ming

    2016-01-01

    During fast-ignition implosions, preheating of inside the cone tip caused by hard X-rays can strongly affect the generation and transport of hot electrons in the cone. Although indirect-drive implosions have a higher implosion symmetry, they cause stronger preheating effects than direct-drive implosions. To control the preheating of the cone tip, we propose the use of indirect-drive fast-ignition targets with thicker tips. Experiments carried out at the ShenGuang-III prototype laser facility confirmed that thicker tips are effective for controlling preheating. Moreover, these results were consistent with those of 1D radiation hydrodynamic simulations.

  20. A simple method to prevent hard X-ray-induced preheating effects inside the cone tip in indirect-drive fast ignition implosions

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Dongxiao; Shan, Lianqiang; Zhou, Weimin; Wu, Yuchi; Zhu, Bin; Zhang, Feng; Bi, Bi; Zhang, Bo; Zhang, Zhimeng; Shui, Min; He, Yingling; Gu, Yuqiu, E-mail: yqgu@caep.cn; Zhang, Baohan [Science and Technology on Plasma Physics Laboratory, China Academy of Engineering Physics, Mianyang 621900 (China); Research Center of Laser Fusion, China Academy of Engineering Physics, Mianyang 621900 (China); Peng, Xiaoshi; Xu, Tao; Wang, Feng; Yang, Zhiwen; Chen, Tao; Chen, Li; Chen, Ming [Research Center of Laser Fusion, China Academy of Engineering Physics, Mianyang 621900 (China); and others

    2016-06-15

    During fast-ignition implosions, preheating of inside the cone tip caused by hard X-rays can strongly affect the generation and transport of hot electrons in the cone. Although indirect-drive implosions have a higher implosion symmetry, they cause stronger preheating effects than direct-drive implosions. To control the preheating of the cone tip, we propose the use of indirect-drive fast-ignition targets with thicker tips. Experiments carried out at the ShenGuang-III prototype laser facility confirmed that thicker tips are effective for controlling preheating. Moreover, these results were consistent with those of 1D radiation hydrodynamic simulations.

  1. The effect of repeated preheating of dimethacrylate and silorane-based composite resins on marginal gap of class V restorations.

    Science.gov (United States)

    Alizadeh Oskoee, Parnian; Pournaghi Azar, Fatemeh; Jafari Navimipour, Elmira; Ebrahimi Chaharom, Mohammad Esmaeel; Naser Alavi, Fereshteh; Salari, Ashkan

    2017-01-01

    Background. One of the problems with composite resin restorations is gap formation at resin‒tooth interface. The present study evaluated the effect of preheating cycles of silorane- and dimethacrylate-based composite resins on gap formation at the gingival margins of Class V restorations. Methods. In this in vitro study, standard Class V cavities were prepared on the buccal surfaces of 48 bovine incisors. For restorative procedure, the samples were randomly divided into 2 groups based on the type of composite resin (group 1: di-methacrylate composite [Filtek Z250]; group 2: silorane composite [Filtek P90]) and each group was randomly divided into 2 subgroups based on the composite temperature (A: room temperature; B: after 40 preheating cycles up to 55°C). Marginal gaps were measured using a stereomicroscope at ×40 and analyzed with two-way ANOVA. Inter- and intra-group comparisons were analyzed with post-hoc Tukey tests. Significance level was defined at P composite resin type, preheating and interactive effect of these variables on gap formation were significant (Pcomposite resins (Pcomposite resins at room temperature compared to composite resins after 40 preheating cycles (Pcomposite re-sins. Preheating of silorane-based composites can result in the best marginal adaptation.

  2. Potential of roof-integrated solar collectors for preheating air at drying facilities in Northern Thailand

    Energy Technology Data Exchange (ETDEWEB)

    Roman, Franz; Nagle, Marcus; Leis, Hermann; Mueller, Joachim [Institute of Agricultural Engineering 440e, University of Hohenheim, Garbenstrasse 9, 70599 Stuttgart (Germany); Janjai, Serm [Department of Physics, Silpakorn University, Nakhon Pathom (Thailand); Mahayothee, Busarakorn [Department of Food Technology, Silpakorn University, Nakhon Pathom (Thailand); Haewsungcharoen, Methinee [Department of Food Engineering, Chiang Mai University, Chiang Mai (Thailand)

    2009-07-15

    Longan is one of the most widely cropped fruits in Northern Thailand, where a significant amount of the annual harvest is commercially dried and exported as a commodity. Liquefied petroleum gas is generally used as the energy source for heating the drying air, but concern is growing as fuel prices are expected to increase for the foreseeable future. Meanwhile, with the ample solar radiation in Thailand, the roofs of drying facilities could be adapted to serve as solar collectors to preheat the drying air, thus reducing the energy requirement from fossil fuels. In this study, a simulation program for a flat-plate solar air heater was used to estimate the potential to preheat drying air given the conditions of several longan drying facilities. Results showed that solar collectors can replace up to 19.6% of the thermal energy demand during the drying season. Bigger collectors and smaller air channels result in more useful heat, but attention has to be paid to costs and pressure drop, respectively. Annual monetary savings can reach up to THB 56,000 ({approx}US$ 1800 at US$ 1 THB 31). (author)

  3. Experimental study of a single fuel jet in conditions of highly preheated air combustion

    Energy Technology Data Exchange (ETDEWEB)

    Lille, Simon; Blasiak, W. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Metallurgy

    2000-04-01

    Highly Preheated Air Combustion (HPAC) is a technique to reduce consumption of fuel and decrease NO{sub x} formation in furnaces. The main change that occur in the furnace chamber is that the flow pattern of flue gases changes dramatically resulting in a more uniform heat transfer. The usefulness of regenerative combustion is very clear, but the advantages have so far been accompanied by high levels of pollutants, such as NO{sub x}. The combination of the regeneration technique and internal flue gas recirculation, thus decreasing NO{sub x} and keeping the other advantages, has made HPAC a very attractive combustion technology with application to heat treatment reheating and melting processes. This work gives an introduction to regenerative combustion with diluted air, including theory on flame stabilization. Furthermore, a description of a new test furnace is given with results from a parametric study and from tests using schlieren color visualization, direct photography, and laser Doppler anemometry. In the parametric study NO{sub x}-emission, CO-emission, lift-off, fluctuations, and some flame characteristics are related to nozzle diameter, oxygen concentration, and preheat temperature. For the schlieren technique and direct photography, both still and high-speed cameras were used.

  4. Instrumentation strategies for energy conservation in broiler barns with ventilation air solar pre-heaters

    Energy Technology Data Exchange (ETDEWEB)

    Cordeau, Sebastien; Barrington, Suzelle [Department of Bioresource Engineering, Macdonald Campus of McGill University, 21 111 Lakeshore, Ste Anne de Bellevue, Quebec H9X 3V9 (Canada)

    2010-08-15

    At the present consumption rate, world fossil-fuel reserves are expected to be depleted by 2050 unless their consumption is optimized and supplemented with renewable energy sources. The objective of this project was to evaluate the performance of a simple data acquisition system installed to conduct an energy balance and identify energy saving strategies in two commercial broilers barns with ventilation air solar pre-heaters. Located near Montreal, Canada, the two identical barns were instrumented for inside and outside air conditions, ventilation rate and energy recovery by the solar air pre-heaters. Whereas the temperature, relative humidity and radiation sensors were reliable, inside air temperature stratification complicated energy balance analyses and broiler heat production rate calculations. Lack of room air mixing resulted in the loss of 25 and 15% of the generated heater load and recovered solar energy. The proper monitoring of all environmental conditions required their measurement every 5 rather than 20 min. Instead of using a data transmission service found to be unreliable in rural areas, all data loggers were downloaded onto a portable computer every 45 days during regular instrument maintenance. Accordingly, room air mixing is recommended to facilitate energy balance studies and improve the efficient use of heating energies. (author)

  5. Gravitational waves from Abelian gauge fields and cosmic strings at preheating

    International Nuclear Information System (INIS)

    Dufaux, Jean-Francois; Figueroa, Daniel G.; Garcia-Bellido, Juan

    2010-01-01

    Primordial gravitational waves provide a very important stochastic background that could be detected soon with interferometric gravitational wave antennas or indirectly via the induced patterns in the polarization anisotropies of the cosmic microwave background. The detection of these waves will open a new window into the early Universe, and therefore it is important to characterize in detail all possible sources of primordial gravitational waves. In this paper we develop theoretical and numerical methods to study the production of gravitational waves from out-of-equilibrium gauge fields at preheating. We then consider models of preheating after hybrid inflation, where the symmetry breaking field is charged under a local U(1) symmetry. We analyze in detail the dynamics of the system in both momentum and configuration space. We show that gauge fields leave specific imprints in the resulting gravitational wave spectra, mainly through the appearance of new peaks at characteristic frequencies that are related to the mass scales in the problem. We also show how these new features in the spectra correlate with stringlike spatial configurations in both the Higgs and gauge fields that arise due to the appearance of topological winding numbers of the Higgs around Nielsen-Olesen strings. We study in detail the time evolution of the spectrum of gauge fields and gravitational waves as these strings evolve and decay before entering a turbulent regime where the gravitational wave energy density saturates.

  6. Symbiotic potential: the integration of preheating and dry cooling in cokemaking

    Energy Technology Data Exchange (ETDEWEB)

    Barker, J E [British Carbonization Research Association, England; Bruce, J M; Kemmetmueller, R

    1978-06-01

    The expression closed energy cycle has become popular in the last decade as descriptive of industrial systems in which exhaust heat is recovered from a primary energy-conversion stage and utilized either recuperatively or regeneratively within the overall complex. An old and well-proven means of utilizing the sensible heat of the incandescent coke discharged from coke ovens is known as dry cooling. This is being practiced widely in the USSR and Japan, but not yet to any significant extent in the western world. The waste heat recovered by this system is normally used to raise steam for power generation and process use. A recent advance in the carbonization of coal for the manufacture of metallurgical coke has been the application of the technique of coal drying and preheating as a means of improving both coke quality and oven productivity, and this is usually energized by burning gas as a fuel. An alternative configuration, having practical advantages in relation to efficiency of utilization of recovered energy and to safety in operation, is represented by a combination of coal drying and preheating with dry cooling of the coke. This paper is concerned with the case for this combination and the means whereby it may be effected in practice. The energy cycle of cokemaking would thus be more nearly closed.

  7. Gasifier selection, design and gasification of oil palm fronds with preheated and unheated gasifying air.

    Science.gov (United States)

    Guangul, Fiseha M; Sulaiman, Shaharin A; Ramli, Anita

    2012-12-01

    Oil palm frond biomass is abundantly available in Malaysia, but underutilized. In this study, gasifiers were evaluated based on the available literature data and downdraft gasifiers were found to be the best option for the study of oil palm fronds gasification. A downdraft gasifier was constructed with a novel height adjustment mechanism for changing the position of gasifying air and steam inlet. The oil palm fronds gasification results showed that preheating the gasifying air improved the volumetric percentage of H(2) from 8.47% to 10.53%, CO from 22.87% to 24.94%, CH(4) from 2.02% to 2.03%, and higher heating value from 4.66 to 5.31 MJ/Nm(3) of the syngas. In general, the results of the current study demonstrated that oil palm fronds can be used as an alternative energy source in the energy diversification plan of Malaysia through gasification, along with, the resulting syngas quality can be improved by preheating the gasifying air. Copyright © 2012 Elsevier Ltd. All rights reserved.

  8. Effect of preheat repetition on color stability of methacrylate- and silorane-based composite resins.

    Science.gov (United States)

    Abed Kahnamouei, Mehdi; Gholizadeh, Sarah; Rikhtegaran, Sahand; Daneshpooy, Mehdi; Kimyai, Soodabeh; Alizadeh Oskoee, Parnian; Rezaei, Yashar

    2017-01-01

    Background. The aim of this study was to investigate the effect of preheating methacrylate- and silorane-based composite resins on their color stability up to 40 times at 55‒60°C. Methods. Seventy-six methacrylate and silorane-based composite resin samples, with a diameter of 10 mm and a height of 2 mm, were divided into 4 groups (n=19). After the samples were prepared, their color parameters were determined using a reflective spectrophotometer. The composite resin samples were separately stored in a solution of tea for 40 consecutive days. Then the samples underwent a color determination procedure again using a spectrophotometer and color changes were recorded. Finally two-way ANOVA was used to study the effect of composite temperature on its staining (Pcomposite resin samples compared to non-heated samples at P=0.005 and P=0.029 for silorane-based and Z250 composite resin samples, respectively. Results. Both composite resin type (P=0.014) and preheating (Pcomposite resin samples, up to 55‒60°C for 40 rounds, resulted in more color changes compared with unheated composite resin samples. After storage in a solution of tea the color change rate in the composite resin samples of silorane-based was higher than the Z250 composite resin samples.

  9. Non-Gaussian and nonscale-invariant perturbations from tachyonic preheating in hybrid inflation

    Science.gov (United States)

    Barnaby, Neil; Cline, James M.

    2006-05-01

    We show that in hybrid inflation it is possible to generate large second-order perturbations in the cosmic microwave background due to the instability of the tachyonic field during preheating. We carefully calculate this effect from the tachyon contribution to the gauge-invariant curvature perturbation, clarifying some confusion in the literature concerning nonlocal terms in the tachyon curvature perturbation; we show explicitly that such terms are absent. We quantitatively compute the non-Gaussianity generated by the tachyon field during the preheating phase and translate the experimental constraints on the nonlinearity parameter fNL into constraints on the parameters of the model. We also show that nonscale-invariant second-order perturbations from the tachyon field with spectral index n=4 can become larger than the inflaton-generated first-order perturbations, leading to stronger constraints than those coming from non-Gaussianity. The width of the excluded region in terms of the logarithm of the dimensionless coupling g, grows linearly with the log of the ratio of the Planck mass to the tachyon VEV, log⁡(Mp/v); hence very large regions are ruled out if the inflationary scale v is small. We apply these results to string-theoretic brane-antibrane inflation, and find a stringent upper bound on the string coupling, gs<10-4.5.

  10. Modification of preheated tungsten surface after irradiation at the GOL-3 facility

    Energy Technology Data Exchange (ETDEWEB)

    Shoshin, A.A., E-mail: shoshin@mail.ru [Budker Institute of Nuclear Physics SB RAS, Novosibirsk 630090 (Russian Federation); Novosibirsk State University, Novosibirsk 630090 (Russian Federation); Arakcheev, A.S.; Arzhannikov, A.V. [Budker Institute of Nuclear Physics SB RAS, Novosibirsk 630090 (Russian Federation); Novosibirsk State University, Novosibirsk 630090 (Russian Federation); Burdakov, A.V. [Budker Institute of Nuclear Physics SB RAS, Novosibirsk 630090 (Russian Federation); Novosibirsk State Technical University, Novosibirsk 630092 (Russian Federation); Huber, A. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung, 52425 Jülich (Germany); Ivanov, I.A. [Budker Institute of Nuclear Physics SB RAS, Novosibirsk 630090 (Russian Federation); Novosibirsk State University, Novosibirsk 630090 (Russian Federation); Kuklin, K.N. [Budker Institute of Nuclear Physics SB RAS, Novosibirsk 630090 (Russian Federation); Polosatkin, S.V.; Postupaev, V.V.; Sinitsky, S.L. [Budker Institute of Nuclear Physics SB RAS, Novosibirsk 630090 (Russian Federation); Novosibirsk State University, Novosibirsk 630090 (Russian Federation); Vasilyev, A.A. [Novosibirsk State University, Novosibirsk 630090 (Russian Federation)

    2016-12-15

    Highlights: • Preheated tungsten was irradiated at the GOL-3 facility with plasma loads corresponding to the ITER type I ELMs. • The crack pattern and the quantity of bubbles depend on the initial temperatures of the target. • The orientation of major crack networks correlates with the direction of machining of the samples. • Dust impact craters were found. - Abstract: The study is devoted to tungsten surface modification after irradiation at the GOL-3 facility with plasma loads corresponding to the ITER type I ELMs. In order to emulate heating with a steady plasma flux in the ITER divertor, some of the tungsten samples were preheated up to 500 °C. It was found out that the behavior of the surface modification (the crack pattern and the number of bubbles) depends on the initial temperature of the targets. While the orientation of major crack networks correlates with the direction of machining of the samples. Afterwards we have observed the process of craters’ formation caused by dust particle impacts.

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

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-01-01

    A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of this study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. The primary results of the study are as follows: a) the lithium-lead blanket achieves a higher TBR with a smaller blanket thickness relative to the lithium blanket; b) the lithium blanket generates more energy per fusion neutron relative to the lithium-lead blanket; c) among the possible reflector materials, the carbon reflector produces the highest TBR; d) the high-Z reflector materials (Mo, Cu, W, or steel) generate more energy per fusion neutron and produce smaller TBRs relative to the carbon reflector; e) lithium-6 enrichment is required for the lithium-lead blanket to reduce the total blanket thickness; and f) the energy deposition per fusion neutron reaches a saturation as the blanket thickness, the fraction of the high-Z material in the reflector, or the reflector zone thickness increases (this allows one to design the blanket for a specific TBR without reducing the energy production)

  12. Breeding blanket development. Tritium release from breeder

    International Nuclear Information System (INIS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nagao, Yoshiharu

    2006-01-01

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

  13. Effect of the Preheating Temperature on Process Time in Friction Stir Welding of Al 6061-T6

    DEFF Research Database (Denmark)

    Jabbari, Masoud

    2013-01-01

    This paper presents the results obtained and the deductions made from an analytical modeling involving friction stir welding of Al 6061-T6. A new database was developed to simulate the contact temperature between the tool and the workpiece. A second-order equation is proposed for simulating...... the temperature in the contact boundary and the thermal history during the plunge phase. The effect of the preheating temperature on the process time was investigated with the proposed model. The results show that an increase of the preheating time leads to a decrease in the process time up to the plunge...

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

    Science.gov (United States)

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

    2017-04-01

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

  15. The transpiration cooled first wall and blanket concept

    International Nuclear Information System (INIS)

    Barleon, Leopold; Wong, Clement

    2002-01-01

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

  16. Assessment of alkali metal coolants for the ITER blanket

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  17. Tritium transport in the water cooled Pb-17Li blanket concept of DEMO

    International Nuclear Information System (INIS)

    Reiter, F.; Tominetti, S.; Perujo, A.

    1992-01-01

    The code TIRP has been used to calculate the time dependence of tritium inventory and tritium permeation into the coolant and into the first wall boxes in the water cooled Pb-17Li blanket concept of DEMO. The calculations have been performed for the martensitic steel MANET and the austenitic steel AISI 316L as blanket structure materials, for water or helium cooling and for convective or no motion of the liquid breeder in the blanket. Tritium inventories are rather low in blankets with MANET structure and higher in those with AISI 316L structure. Tritium permeation rates are too high in both blankets. Further calculations on tritium inventory and permeation are therefore presented for blankets with TiC permeation barriers of 1 μm thickness on various surfaces of the blanket structure and for blankets with any permeation barriers in function of their thickness, tritium diffusivities, tritium surface recombination rates and atomic densities. These last calculations have been performed for a blanket with coatings on the outer surfaces of the blanket and with a tritium residence time of 10 4 s and for a blanket with coatings on both sides of the cooling tubes and stagnant Pb-17Li in the blanket. The second case for a blanket with MANET structure presents a very interesting solution for tritium recovery by permeation into and pumping from the first wall boxes. (orig.)

  18. Status of blanket design for RTO/RC ITER

    International Nuclear Information System (INIS)

    Yamada, M.; Ioki, K.; Cardella, A.; Elio, F.; Miki, N.

    2000-01-01

    Design has progressed on the FW/blanket for the RTO/RC (reduced technical objective/ reduced cost) ITER. The basic functions and structures are the same as for the 1998 ITER design. However, design and fabrication methods of the FW/blanket have been improved to achieve ∝ 50% reduction of the construction cost compared to that for the 1998 ITER design. Detailed blanket module designs with flat separable FW panels have been developed to reduce the fabrication cost and the future radioactive waste. Most of the R and D performed so far during the EDA (engineering design activity) is still applicable. Further cost reduction methods are also being investigated and additional R and D is being performed. (orig.)

  19. Computation Method Comparison for Th Based Seed-Blanket Cores

    International Nuclear Information System (INIS)

    Kolesnikov, S.; Galperin, A.; Shwageraus, E.

    2004-01-01

    This work compares two methods for calculating a given nuclear fuel cycle in the WASB configuration. Both methods use the ELCOS Code System (2-D transport code BOXER and 3-D nodal code SILWER) [4] are compared. In the first method, the cross-sections of the Seed and Blanket, needed for the 3-D nodal code are generated separately for each region by the 2-D transport code. In the second method, the cross-sections of the Seed and Blanket, needed for the 3-D nodal code are generated from Seed-Blanket Colorsets (Fig.1) calculated by the 2-D transport code. The evaluation of the error introduced by the first method is the main objective of the present study

  20. Neutronics design aspects of reference ARIES-I fusion blanket

    International Nuclear Information System (INIS)

    Cheng, E.T.

    1990-12-01

    A SiC composite blanket concept was recently conceived for a deuterium-tritium burning, 1000 MW(e) tokamak fusion reactor design, ARIES-I. SiC composite structural material was chosen due to its very low activation features. High blanket nuclear performance and thermal efficiency, adequate tritium breeding, and a low level of activation are important design requirements for the ARIES-I reactor. The major approaches, other than using SiC as structural material, in meeting these design requirements, are to employ beryllium, the only low activation neutron multiplying material, and isotopically tailored Li 2 ZrO 3 , a tritium breeding material stable at high temperature, as blanket materials. 5 refs., 4 figs., 2 tabs

  1. Direct LiT Electrolysis in a Metallic Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-30

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

  2. High temperature blankets for the production of synthetic fuels

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  3. Heating an aquaculture pond with a solar pool blanket

    Energy Technology Data Exchange (ETDEWEB)

    Wisely, B; Holliday, J E; MacDonald, R E

    1982-01-01

    A floating solar blanket of laminated bubble plastic was used to heat a 0.11 ha seawater pond of 1.3 m depth. The covered pond maintained daily temperatures 6 to 9/sup 0/C above two controls. Local air temperatures averaged 14 to 19/sup 0/C. Oysters, prawns, seasquirts, and fish in the covered pond all survived. After three weeks, the blanket separated. This was the result of pond temperatures exceeding 30/sup 0/C, the maximum manufacturer's specification. Floating blankets fabricated to higher specifications would be useful for maintaining above-ambient temperatures in small ponds or tanks in temporary situations during cold winter months and might have a more permanent use.

  4. Solid breeder test blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ying, A. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States)]. E-mail: ying@fusion.ucla.edu; Abdou, M. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Calderoni, P. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Sharafat, S. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Youssef, M. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); An, Z. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Abou-Sena, A. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Kim, E. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States); Reyes, S. [LANL, Livermore, CA (United States); Willms, S. [LANL, Los Alamos, NM (United States); Kurtz, R. [PNNL, Richland, WA (United States)

    2006-02-15

    This paper presents the design and analysis for the US ITER solid breeder blanket test articles. Objectives of solid breeder blanket testing during the first phase of the ITER operation focus on exploration of fusion break-in phenomena and configuration scoping. Specific emphasis is placed on first wall structural response, evaluation of neutronic parameters, assessment of thermomechanical behavior and characterization of tritium release. The tests will be conducted with three unit cell arrays/sub-modules. The development approach includes: (1) design the unit cell/sub-module for low temperature operations and (2) refer to a reactor blanket design and use engineering scaling to reproduce key parameters under ITER wall loading conditions, so that phenomena under investigation can be measured at a reactor-like level.

  5. Direct LiT Electrolysis in a Metallic Fusion Blanket

    International Nuclear Information System (INIS)

    Olson, Luke

    2016-01-01

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

  6. The evolution of US helium-cooled blankets

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  7. Status of fusion reactor blanket evaluation studies in France

    International Nuclear Information System (INIS)

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

    1985-03-01

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

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

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Porfiri, T.

    1996-06-01

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

  9. Damage to preheated tungsten targets after multiple plasma impacts simulating ITER ELMs

    Energy Technology Data Exchange (ETDEWEB)

    Garkusha, I.E. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine)], E-mail: garkusha@ipp.kharkov.ua; Bandura, A.N.; Byrka, O.V.; Chebotarev, V.V. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine); Landman, I. [Forschungszentrum Karlsruhe, IHM, 76021 Karlsruhe (Germany); Makhlaj, V.A. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine); Pestchanyi, S. [Forschungszentrum Karlsruhe, IHM, 76021 Karlsruhe (Germany); Tereshin, V.I. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine)

    2009-04-30

    The behavior of a preheated at 650 deg. C tungsten targets under repetitive ELM-like plasma pulses is studied in simulation experiments with the quasi-stationary plasma accelerator QSPA Kh-50. The targets have been exposed up to 350 pulses of the duration 0.25 ms and the surface heat loads either 0.45 MJ/m{sup 2} or 0.75 MJ/m{sup 2}, which is below and above the melting threshold, respectively. The development of surface morphology of the exposed targets as well as cracking and swelling at the surface is discussed. First comparisons of obtained experimental results with corresponding numerical simulations of the code PEGASUS-3D are presented.

  10. New Colloidal Lithographic Nanopatterns Fabricated by Combining Pre-Heating and Reactive Ion Etching

    Directory of Open Access Journals (Sweden)

    Cong Chunxiao

    2009-01-01

    Full Text Available Abstract We report a low-cost and simple method for fabrication of nonspherical colloidal lithographic nanopatterns with a long-range order by preheating and oxygen reactive ion etching of monolayer and double-layer polystyrene spheres. This strategy allows excellent control of size and morphology of the colloidal particles and expands the applications of the colloidal patterns as templates for preparing ordered functional nanostructure arrays. For the first time, various unique nanostructures with long-range order, including network structures with tunable neck length and width, hexagonal-shaped, and rectangular-shaped arrays as well as size tunable nanohole arrays, were fabricated by this route. Promising potentials of such unique periodic nanostructures in various fields, such as photonic crystals, catalysts, templates for deposition, and masks for etching, are naturally expected.

  11. New pre-heating system for natural gas pressure regulating stations

    International Nuclear Information System (INIS)

    Zullo, G.; Vertuani, C.; Borghesani, O.; Vignoli, F.

    1999-01-01

    Costs for running natural gas pressure regulating stations are mainly due to operation and maintenance of a natural gas preheating system, usually equipment with a hot water boiler or an armour-plated electric resistance immersed in a fluid. The article describe a system, considering a natural circulation boiler which uses steam/condensate (at 100 degrees C and 0,5 bar) as a thermal conductor, in thermodynamic balance and in absence of un condensable. This new boiler, already operating with satisfactory results in heating system for industrial buildings, does not require testing, notifications, periodical inspections by the competent authorities, constant monitoring by trained or patented staff. Besides, it allows easier installations procedures and running cost savings. The system, to be considered as static because it has no moving parts, is a good alternative to conventional forced hot water circulation or electric heating system [it

  12. Preheat effect on titanium plate fabricated by sputter-free selective laser melting in vacuum

    Science.gov (United States)

    Sato, Yuji; Tsukamoto, Masahiro; Shobu, Takahisa; Yamashita, Yorihiro; Yamagata, Shuto; Nishi, Takaya; Higashino, Ritsuko; Ohkubo, Tomomasa; Nakano, Hitoshi; Abe, Nobuyuki

    2018-04-01

    The dynamics of titanium (Ti) melted by laser irradiation was investigated in a synchrotron radiation experiment. As an indicator of wettability, the contact angle between a selective laser melting (SLM) baseplate and the molten Ti was measured by synchrotron X-rays at 30 keV during laser irradiation. As the baseplate temperature increased, the contact angle decreased, down to 28° at a baseplate temperature of 500 °C. Based on this result, the influence of wettability of a Ti plate fabricated by SLM in a vacuum was investigated. It was revealed that the improvement of wettability by preheating suppressed sputtering generation, and a surface having a small surface roughness was fabricated by SLM in a vacuum.

  13. Solid state NMR studies for a new carbonization process with high temperature preheating

    Science.gov (United States)

    Saito, Koji; Hatakeyama, Moriaki; Komaki, Ikuo; Katoh, Kenji

    2002-01-01

    A new carbonization process with rapid preheating and coke discharging at medium temperature has been developed in Japan. The result of this process shows that even when no or slightly coking coal is by 50 wt% the coking property is improved and a coking coke with cold strength usable at blast furnace can be manufactured with the new carbonization process. The mechanism of the coking property improvement was examined by coal properties using mainly solid state NMR ( 1H CRAMPS and 13C SPE/MAS, CP/MAS) and NMR imaging (single point imaging, in-situ imaging). It has been clarified that the molecular structure of coal is relaxed by the rapid heating treatment and, in addition, there is a close relation between hydrogen bonding and relaxation of the molecular structure of coal.

  14. Overview of the TFTR Lithium Blanket Module program

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-01-01

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

  15. ITER blanket module shield block design and analysis

    International Nuclear Information System (INIS)

    Mitin, D.; Khomyakov, S.; Razmerov, A.; Strebkov, Yu.

    2008-01-01

    This paper presents the alternative design of the shield block cooling path for a typical ITER blanket module with a predominantly sequential flow circuit. A number of serious disadvantages have been observed for the reference design, where the parallel flow circuit is used, which is inherent in the majority of blanket modules. The paper discusses these disadvantages and demonstrates the benefit of the alternative design based on the detailed design and the technological, hydraulic, thermal, structural and strength analyses, conducted for module no. 17

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  17. Ceramic BOT type blanket with poloidal helium cooling

    International Nuclear Information System (INIS)

    Cardella, A.; Daenenr, W.; Iseli, M.; Ferrari, M.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.

    1989-01-01

    This paper briefly describes the work done and results achieved over the past two years on the ceramic breeder BOT blanket with poloidal helium cooling. A conclusive remark on the brick/plate option described previously is followed by short descriptions of the low and high performance pebble bed options elaborated as alternatives for both NET and DEMO. The results show, togethre with those about the poloidal cooling of the First Wall, good prospects for this blanket type provided that the questions connected wiht an extensive use of beryllium find a satisfactor answer. (author). 5 refs.; 7 figs.; 1 tab

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  19. Limiter and first wall of the fusion reactor blanket

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  20. Ceramic sphere-pac breeder design for fusion blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.J.; Sullivan, J.D.

    1991-01-01

    Randomly packed beds of ceramic spheres are a practical approach to surrounding fusion plasmas with tritium-breeding material. This paper examines the general properties of sphere-pac beds for application in fusion breeder blankets. The design considerations and models are reviewed for packing, tritium breeding and recovery, thermal conductivity, purge-gas pressure drop, mechanical behavior and fabrication. The design correlations are compared against available fusion ceramic data. Specific conclusions are that ternary (three-size) beds are not attractive for fusion blankets, and that the fusion spheres should be as large as possible subject primarily to packing constraints. (orig.)

  1. An aqueous lithium salt blanket option for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Steiner, D.; Varsamis, G. (Rensselaer Polytechnic Inst., Troy, NY (USA). Dept. of Nuclear Engineering and Engineering Physics); Deutsch, L.; Rathke, J. (Grumman Corp., Bethpage, NY (USA). Advanced Energy Systems); Gierszewski, P. (Canadian Fusion Fuels Technology Project (CFFTP), Mississauga, ON (Canada))

    1989-04-01

    An aqueous lithium salt blanket (ALSB) concept is proposed which could be the basis for either a power reactor blanket or a test module in an engineering test reactor. The design is based on an austenitic stainless steel structure, a beryllium multiplier, and a salt breeder concentration of about 32 g LiNO/sub 3/ per 100 cm/sup 3/ of H/sub 2/O. To limit tritium release rates, the salt breeder solution is separated from the water coolant circuit. The overall tritium system cost for a 2400 MW (fusion power) reactor is estimated to be 180 million Dollar US87 installed. (orig.).

  2. Study of sensitivity change of OSL signals from quartz and feldspars as a function of preheat temperature

    DEFF Research Database (Denmark)

    Jungner, H.; Bøtter-Jensen, L.

    1994-01-01

    and as a result, the equivalent dose (ED) would be underestimated. A study of sensitivity changes in feldspars and quartz was carried out with emphasis on the effect of preheat and annealing on the OSL signal. Measurement results obtained are presented, and possible elimination of errors in dating caused...

  3. Experimental and Modeling Investigation of the Effect of Air Preheat on the Formation of NOx in an RQL Combustor

    Science.gov (United States)

    Samuelsen, G. S.; Brouwer, J.; Vardakas, M. A.; Holderman, J. D.

    2012-01-01

    The Rich-burn/Quick-mix/Lean-burn (RQL) combustor concept has been proposed to minimize the formation of oxides of nitrogen (NOx) in gas turbine systems. The success of this low-NOx combustor strategy is dependent upon the links between the formation of NOx, inlet air preheat temperature, and the mixing of the jet air and fuel-rich streams. Chemical equilibrium and kinetics modeling calculations and experiments were performed to further understand NOx emissions in an RQL combustor. The results indicate that as the temperature at the inlet to the mixing zone increases (due to preheating and/or operating conditions) the fuel-rich zone equivalence ratio must be increased to achieve minimum NOx formation in the primary zone of the combustor. The chemical kinetics model illustrates that there is sufficient residence time to produce NOx at concentrations that agree well with the NOx measurements. Air preheat was found to have very little effect on mixing, but preheating the air did increase NOx emissions significantly. By understanding the mechanisms governing NOx formation and the temperature dependence of key reactions in the RQL combustor, a strategy can be devised to further reduce NOx emissions using the RQL concept.

  4. Utilization of biogas released from palm oil mill effluent for power generation using self-preheated reactor

    International Nuclear Information System (INIS)

    Hosseini, Seyed Ehsan; Wahid, Mazlan Abdul

    2015-01-01

    Highlights: • A lab-scale reactor called self-preheating flameless combustion (SPFC) system is experimented. • Feasibility of power generation by POME biogas is modeled using SPFC system. • 4 MW power is available by POME biogas utilization in a typical palm oil mill with 300,000 tons production. • The rate of power generation increases when 2% hydrogen is added to POME biogas ingredients. - Abstract: In palm oil mills, for one ton crude palm oil (CPO) production, 70 m"3 biogas is released from palm oil mill effluent (POME) which can endanger the environment. Palm oil mills without appropriate strategies for biogas collection can participate in greenhouse gases (GHGs) generation actively. In this paper, a typical palm oil mill with annual capacity of 300,000 ton oil palm production and 3 MW electricity demand is considered as a pilot plant and feasibility of power generation by POME biogas is modeled by Aspen Plus considering flameless mode in combustion system. A new design of lab-scale flameless reactor called self-preheated flameless combustion (SPFC) system is presented and employed in power generation modeling. In SPFC system, the flameless chamber is employed as a heater to preheat an oxidizer over the self-ignition temperature of the fuel. A helical stainless steel pipe (called self-preheating pipe) is installed inside the chamber to conduct the oxidizer from exhaust zone to the combustion zone inside the chamber and preheat oxidizer. In the flameless mode, the diluted oxidizer is injected to the helical pipe from the exhaust zone and the preheated oxidizer at the burner is conducted to the flameless furnace through a distributor. In SPFC system external heater for preheating oxidizer is removed and the rate of power generation increases. The results show that 10.8 MW power could be generated in ultra-lean POME biogas SPFC. However, the rate of pollutant especially CO_2 and NO_x is high in this circumstances. In stoichiometric condition, 4 MW power

  5. Hydrogen preheating through waste heat recovery of an open-cathode PEM fuel cell leading to power output improvement

    International Nuclear Information System (INIS)

    Mohamed, W.A.N.W.; Kamikl, M. Haziq M.

    2016-01-01

    Highlights: • A study on the effect of hydrogen preheating using waste heat for low temperature PEM fuel cells. • Theoretical, experimental and analytical framework was established. • The maximum electrical power output increases by 8–10% under specific operating conditions. • Open loop hydrogen supply gives a better performance than closed loop. • The waste heat utilization is less than 10% due to heat capacity limitations. - Abstract: The electrochemical reaction kinetics in a Polymer Electrolyte Membrane (PEM) fuel cell is highly influenced by the reactants supply pressures and electrode temperatures. For an open cathode PEM fuel cell stack, the power output is constrained due to the use of air simultaneously as reactant and coolant. Optimal stack operation temperatures are not achieved especially at low to medium power outputs. Based on the ideal gas law, higher reactant temperatures would lead to higher pressures and subsequently improve the reaction kinetics. The hydrogen supply temperature and its pressure can be increased by preheating; thus, slightly offsetting the limitation of low operating stack temperatures. The exit air stream offers an internal source of waste heat for the hydrogen preheating purpose. In this study, a PEM open-cathode fuel cell was used to experimentally evaluate the performance of hydrogen preheating based on two waste heat recovery approaches: (1) open-loop and (2) closed loop hydrogen flow. The stack waste heat was channelled into a heat exchanger to preheat the hydrogen line before it is being supplied (open loop) or resupplied (closed loop) into the stack. At a constant 0.3 bar hydrogen supply pressure, the preheating increases the hydrogen temperature in the range of 2–13 °C which was dependant on the stack power output and cathode air flow rates. The achievable maximum stack power was increased by 8% for the closed loop and 10% for the open loop. Due to the small hydrogen flow rates, the waste heat utilization

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

    International Nuclear Information System (INIS)

    Wang, Q.; Henderson, D.L.

    1995-01-01

    Pulsed activation calculations have been performed on two blanket options considered as part of the ITER home team blanket trade-off study. The objective was to compare the activity, afterheat and waste disposal rating (WDR) results of a composite blanket-shield design for the continuous operation approximation to a pulsed operation case to determine whether the differences are at most the duty factor as predicted by the two nuclide chain model. Up to a cooling period of 100 years, the pulsed activity and afterheat values were below the continuous oepration results and well within (except for one afterheat value) the maximum deviation predicted by the two nuclide chain model. No differences in the WDR values were noted as they are, to a large extent, based on long-lived nuclides which are insensitive to short-term changes in the operation history. (orig.)

  7. A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; Trow, Martin; Dillistone, Michael

    2016-01-01

    'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

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

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Seki, Yasushi; Minato, Akio; Kobayashi, Takeshi; Mori, Seiji; Kawasaki, Hiromitsu; Sumita, Kenji.

    1983-02-01

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

  9. Design requirement on KALIMER blanket fuel assembly duct

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang, H. Y.; Nam, C.; Kim, J. O.

    1998-03-01

    This document describes design requirements which are needed for designing the blanket fuel assembly duct of the KALIMER as design guidance. The blanket fuel assembly duct of the KALIMER consists of fuel rods, mounting rail, nosepiece, duct with pad, handling socket with pad. Blanket fuel rod consists of top end plug, bottom end plug with solid ferritic-martensitic steel rod and key way blanket fuel slug, cladding, and wire wrap. In the assembly, the rods are in a triangular pitch array, and the rod bundle is attached to the nosepiece with mounting rails. The bottom end of the assembly duct is formed by a long nosepiece which provides the lower restraint function and the paths for coolant inlet. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. (author). 20 refs., 4 figs

  10. Summary of the target-blanket breakout group

    Energy Technology Data Exchange (ETDEWEB)

    Capiello, M.; Bell, C. [Los Alamos National Laboratory, NM (United States); Barthold, W.

    1995-10-01

    This breakout group discussed a number of topics and issues pertaining to target and blanket concepts for accelerator-driven systems. This major component area is one marked by a broad spectrum of technical approaches. It is therefore less defined than other major component areas such as the accelerator and is at an earlier stage of technical needs and task specification. The working group did reach a number of general conclusions and recommendations that are summarized. The Conference and the Target/Blanket Breakout Group provided a first opportunity for people working on a variety of missions and concepts to get together and exchange information. A number of subcritical systems applicable for a spectrum of missions were proposed at the Conference and discussed in the Breakout Group. Missions included plutonium disposition, energy production, waste destruction, isotope production, and neutron scattering. The Target/Blanket Breakout Group also defined areas where parameters and data should be addressed as target/blanket design activities become more detailed and sophisticated.

  11. Technical issues for beryllium use in fusion blanket applications

    International Nuclear Information System (INIS)

    McCarville, T.J.; Berwald, D.H.; Wolfer, W.; Fulton, F.J.; Lee, J.D.; Maninger, R.C.; Moir, R.W.; Beeston, J.M.; Miller, L.G.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Smith, D.L.; Mattas, R.F.; Clemmer, R.G.; Davis, J.W.

    1980-01-01

    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

  13. Thermal-hydraulic analysis of low activity fusion blanket designs

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.; Yu, W.S.

    1977-01-01

    The heat transfer aspects of fusion blankets are considered where: (a) conduction and (b) boiling and condensation are the dominant heat transfer mechanisms. In some cases, unique heat transfer problems arise and additional heat transfer data and analyses may be required

  14. Fusion blanket testing in MFTF-α + T

    International Nuclear Information System (INIS)

    Kleefeldt, K.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Lehnert, B.

    1977-12-01

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

  16. Performance evaluation on force control for ITER blanket installation

    Energy Technology Data Exchange (ETDEWEB)

    Aburadani, A., E-mail: aburadani.atsushi@jaea.go.jp [Japan Atomic Energy Agency, Mukouyama 801-1, Naka, Ibaraki 311-0193 (Japan); Takeda, N.; Shigematsu, S.; Murakami, S.; Tanigawa, H.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka, Ibaraki 311-0193 (Japan); Nakahira, M.; Hamilton, D.; Tesini, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► It is crucial issues to avoid any jamming between the blanket modules and the keys. ► Force control for AC servo motor was developed to reduce excessive loads. ► This jam prevention force control method is directly measured and controlled by AC servo motor controllers. ► In the recent test, the module was passively positioned onto keys using the torque control method. -- Abstract: The most critical issue for the ITER blanket installation is to avoid any jamming between the blanket modules and the keys as a result of excessive loading during the module installation process. This is complicated by the limited clearance of 0.5 mm between the modules and the keys. To solve these technical issues, force control, such as controlling the torque for the AC servo motors, was developed to reduce excessive loads which may have an impact on the end-effector and to defer the forces acting on the groove of the blanket. This jam prevention force control method is directly measured and controlled by AC servo motor controllers. The AC servo motors are equipped to move the manipulator and end-effector during module installation.

  17. Performance evaluation on force control for ITER blanket installation

    International Nuclear Information System (INIS)

    Aburadani, A.; Takeda, N.; Shigematsu, S.; Murakami, S.; Tanigawa, H.; Kakudate, S.; Nakahira, M.; Hamilton, D.; Tesini, A.

    2013-01-01

    Highlights: ► It is crucial issues to avoid any jamming between the blanket modules and the keys. ► Force control for AC servo motor was developed to reduce excessive loads. ► This jam prevention force control method is directly measured and controlled by AC servo motor controllers. ► In the recent test, the module was passively positioned onto keys using the torque control method. -- Abstract: The most critical issue for the ITER blanket installation is to avoid any jamming between the blanket modules and the keys as a result of excessive loading during the module installation process. This is complicated by the limited clearance of 0.5 mm between the modules and the keys. To solve these technical issues, force control, such as controlling the torque for the AC servo motors, was developed to reduce excessive loads which may have an impact on the end-effector and to defer the forces acting on the groove of the blanket. This jam prevention force control method is directly measured and controlled by AC servo motor controllers. The AC servo motors are equipped to move the manipulator and end-effector during module installation

  18. Tritium inventory in Li2ZrO3 blanket

    International Nuclear Information System (INIS)

    Nishikawa, M.; Baba, A.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Brislin, R.J.; Middleton, A.

    1979-08-01

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

  20. Effects of buffer thickness on ATW blanket performances

    International Nuclear Information System (INIS)

    Yang, Won Sik

    2001-01-01

    This paper presents the preliminary results of target and buffer design studies for a lead-bismuth eutectic (LBE) cooled accelerator transmutation of waste (ATW) system, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using an 840 MWt LBE cooled ATW design, the effects of buffer thickness on the blanket performances have been studied. Varying the buffer thickness for a given blanket configuration, system performances have been estimated by a series of calculations using MCNPX and REBUS-3 codes. The effects of source importance change are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. As the irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. The results show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable

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

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakanhira, Masataka; Matsumoto, Yasuhiro; Shibanuma, K.

    2007-01-01

    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

  2. Remote handling assessment of attachment concepts for DEMO blanket segments

    Energy Technology Data Exchange (ETDEWEB)

    Iglesias, Daniel, E-mail: daniel.iglesias@ccfe.ac.uk [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Bastow, Roger; Cooper, Dave; Crowe, Robert; Middleton-Gear, Dave [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Sibois, Romain [VTT, Technical Research Centre of Finland, Industrial Systems, ROViR, Tampere (Finland); Carloni, Dario [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT) (Germany); Vizvary, Zsolt; Crofts, Oliver [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Harman, Jon [EFDA Close Support Unit Garching, Boltzmannstaße 2, D-85748 Garching bei München (Germany); Loving, Antony [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2015-10-15

    Highlights: • Challenges are identified for the remote handling of blanket segments’ attachments. • Two attachment design approaches are assessed for remote handling (RH) feasibility. • An alternative is proposed, which potentially simplifies and speeds-up RH operations. • Up to three different assemblies are proposed for the remote handling of the attachments. • Proposed integrated design of upper port is compatible with the attachment systems. - Abstract: The replacement strategy of the massive Multi-Module Blanket Segments (MMS) is a key driver in the design of several DEMO systems. These include the blankets themselves, the vacuum vessel (VV) and its ports and the Remote Maintenance System (RMS). Common challenges to any blanket attachment system have been identified, such as the need for applying a preload to the MMS manifold, the effects of the decay heat and several uncertainties related to permanent deformations when removing the blanket segments after service. The WP12 kinematics of the MMS in-vessel transportation was adapted to the requirements of each of the supports during 2013 and 2014 design activities. The RM equipment envisaged for handling attachments and earth connections may be composed of up to three different assemblies. An In-Vessel Mover at the divertor level handles the lower support and earth bonding, and could stabilize the MMS during transportation. A Shield Plug crane with a 6 DoF manipulator operates the upper attachment and earth straps. And a Vertical Maintenance Crane is responsible for the in-vessel MMS transportation and can handle the removable upper support pins. A final proposal is presented which can potentially reduce the number of required systems, at the same time that speeds-up the RMS global operations.

  3. Effects of buffer thickness on ATW blanket performance

    International Nuclear Information System (INIS)

    Yang, W. S.; Mercatali, L.; Taiwo, T. A.; Hill, R. N.

    2001-01-01

    This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy ( and lt; 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level

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

    Science.gov (United States)

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

    2011-03-01

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

  5. Structural analysis under the Blanket Comparison and Selection Study

    International Nuclear Information System (INIS)

    Majumdar, S.

    1985-01-01

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

  6. Effects of Buffer Thickness on ATW Blanket Performance

    International Nuclear Information System (INIS)

    Yang, W.S.; Mercatali, L.; Taiwo, T.A.; Hill, R.N.

    2002-01-01

    This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level. (authors)

  7. Measuring skin necrosis in a randomised controlled feasibility trial of heat preconditioning on wound healing after reconstructive breast surgery: study protocol and statistical analysis plan for the PREHEAT trial.

    Science.gov (United States)

    Cro, Suzie; Mehta, Saahil; Farhadi, Jian; Coomber, Billie; Cornelius, Victoria

    2018-01-01

    Essential strategies are needed to help reduce the number of post-operative complications and associated costs for breast cancer patients undergoing reconstructive breast surgery. Evidence suggests that local heat preconditioning could help improve the provision of this procedure by reducing skin necrosis. Before testing the effectiveness of heat preconditioning in a definitive randomised controlled trial (RCT), we must first establish the best way to measure skin necrosis and estimate the event rate using this definition. PREHEAT is a single-blind randomised controlled feasibility trial comparing local heat preconditioning, using a hot water bottle, against standard care on skin necrosis among breast cancer patients undergoing reconstructive breast surgery. The primary objective of this study is to determine the best way to measure skin necrosis and to estimate the event rate using this definition in each trial arm. Secondary feasibility objectives include estimating recruitment and 30 day follow-up retention rates, levels of compliance with the heating protocol, length of stay in hospital and the rates of surgical versus conservative management of skin necrosis. The information from these objectives will inform the design of a larger definitive effectiveness and cost-effectiveness RCT. This article describes the PREHEAT trial protocol and detailed statistical analysis plan, which includes the pre-specified criteria and process for establishing the best way to measure necrosis. This study will provide the evidence needed to establish the best way to measure skin necrosis, to use as the primary outcome in a future RCT to definitively test the effectiveness of local heat preconditioning. The pre-specified statistical analysis plan, developed prior to unblinded data extraction, sets out the analysis strategy and a comparative framework to support a committee evaluation of skin necrosis measurements. It will increase the transparency of the data analysis for the

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

    Science.gov (United States)

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

    2017-12-01

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

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

    International Nuclear Information System (INIS)

    Zhang Mingchun; Feng Kaiming; Li Zaixin; Zhao Fengchao

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-11-01

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

  11. [Risk Factors for Oxaliplatin-Induced Phlebitis and Venous Pain, and Evaluation of the Preventive Effect of Preheating with a Hot Compress for Administration of Oxaliplatin].

    Science.gov (United States)

    Nakauchi, Kana; Kawazoe, Hitoshi; Miyajima, Risa; Waizumi, Chieko; Rokkaku, Yuki; Tsuneoka, Kikue; Higuchi, Noriko; Fujiwara, Mitsuko; Kojima, Yoh; Yakushijin, Yoshihiro

    2015-11-01

    Venous pain induced by oxaliplatin(L-OHP)is a clinical issue related to adherence to the Cape OX regimen. To prevent LOHP- induced venous pain, we provided nursing care to outpatients who were administered a preheated L -OHP diluted solution using a hot compress. We retrospectively evaluated the risk factors for colorectal cancer patients who had L -OHP induced phlebitis and venous pain. Furthermore, the preventive effect of nursing care was compared between inpatients and outpatients from January 2010 to March 2012. At the L-OHP administration site, any symptoms were defined as phlebitis, whereas pain was defined as venous pain. A total of 132 treatment courses among 31 patients were evaluated. Multivariate logistic regression analysis revealed that both phlebitis and venous pain were significantly more common in female patients (adjusted odds ratio, 2.357; 95%CI: 1.053-5.418; and adjusted odds ratio, 5.754; 95%CI: 2.119-18.567, respectively). The prevalence of phlebitis and venous pain did not differ between inpatients and outpatients (phlebitis, 61.3% vs 67.7%; venous pain, 29.0%vs 19.4%). These results suggest that administration of L-OHP via a central venous route should be considered in female patients.

  12. Surface Characteristics of Machined NiTi Shape Memory Alloy: The Effects of Cryogenic Cooling and Preheating Conditions

    Science.gov (United States)

    Kaynak, Y.; Huang, B.; Karaca, H. E.; Jawahir, I. S.

    2017-07-01

    This experimental study focuses on the phase state and phase transformation response of the surface and subsurface of machined NiTi alloys. X-ray diffraction (XRD) analysis and differential scanning calorimeter techniques were utilized to measure the phase state and the transformation response of machined specimens, respectively. Specimens were machined under dry machining at ambient temperature, preheated conditions, and cryogenic cooling conditions at various cutting speeds. The findings from this research demonstrate that cryogenic machining substantially alters austenite finish temperature of martensitic NiTi alloy. Austenite finish ( A f) temperature shows more than 25 percent increase resulting from cryogenic machining compared with austenite finish temperature of as-received NiTi. Dry and preheated conditions do not substantially alter austenite finish temperature. XRD analysis shows that distinctive transformation from martensite to austenite occurs during machining process in all three conditions. Complete transformation from martensite to austenite is observed in dry cutting at all selected cutting speeds.

  13. Scaling of Pressure with Intensity in Laser-Driven Shocks and Effects of Hot X-Ray Preheat

    International Nuclear Information System (INIS)

    Colvin, Jeffrey D.; Kalantar, Daniel H.

    2006-01-01

    To drive shocks into solids with a laser we either illuminate the material directly, or to get higher pressures, illuminate a plastic ablator that overlays the material of interest. In both cases the illumination intensity is low, <<1013 W/cm2, compared to that for traditional laser fusion targets. In this regime, the laser beam creates and interacts with a collisional, rather than a collisionless, plasma. We present scaling relationships for shock pressure with intensity derived from simulations for this low-intensity collisional plasma regime. In addition, sometimes the plastic-ablator targets have a thin flash-coating of Al on the plastic surface as a shine-through barrier; this Al layer can be a source of hot x-ray preheat. We discuss how the preheat affects the shock pressure, with application to simulating VISAR measurements from experiments conducted on various lasers on shock compression of Fe

  14. Scaling of Pressure with Intensity in Laser-Driven Shocks and Effects of Hot X-ray Preheat

    International Nuclear Information System (INIS)

    Colvin, J D; Kalantar, D H

    2005-01-01

    To drive shocks into solids with a laser we either illuminate the material directly, or to get higher pressures, illuminate a plastic ablator that overlays the material of interest. In both cases the illumination intensity is low, 13 W/cm 2 , compared to that for traditional laser fusion targets. In this regime, the laser beam creates and interacts with a collisional, rather than a collisionless, plasma. We present scaling relationships for shock pressure with intensity derived from simulations for this low-intensity collisional plasma regime. In addition, sometimes the plastic-ablator targets have a thin flashcoating of Al on the plastic surface as a shine-through barrier; this Al layer can be a source of hot x-ray preheat. We discuss how the preheat affects the shock pressure, with application to simulating VISAR measurements from experiments conducted on various lasers on shock compression of Fe

  15. X-ray emission, ablation pressure, and preheating for foils irradiated at 0. 26. mu. m wavelength

    Energy Technology Data Exchange (ETDEWEB)

    Pepin, H.; Fabbro, R.; Faral, B.; Amiranoff, F.; Virmont, J.; Cottet, F.; Romain, J.P.

    1985-11-01

    The x-ray emission, ablation pressure, and preheating for foils irradiated with a 0.26 ..mu..m laser at intensities approx.10/sup 15/ W cm/sup -2/ are studied. The foils are Al with various thicknesses, coated or uncoated with CH or Au. The x-ray emission and conversion efficiency are obtained with a multichannel x-ray diode spectrometer, the ablation pressures are deduced from shock transit times, and the rear temperatures are inferred from x-ray pyrometry. For thin foils (<<12 ..mu..m), the rear temperatures can be predicted reasonably well with the use of the front x-ray spectra. For thick foils shock preheating is dominant.

  16. X-ray emission, ablation pressure, and preheating for foils irradiated at 0.26 μm wavelength

    International Nuclear Information System (INIS)

    Pepin, H.; Fabbro, R.; Faral, B.; Amiranoff, F.; Virmont, J.; Cottet, F.; Romain, J.P.

    1985-01-01

    The x-ray emission, ablation pressure, and preheating for foils irradiated with a 0.26 μm laser at intensities approx.10 15 W cm -2 are studied. The foils are Al with various thicknesses, coated or uncoated with CH or Au. The x-ray emission and conversion efficiency are obtained with a multichannel x-ray diode spectrometer, the ablation pressures are deduced from shock transit times, and the rear temperatures are inferred from x-ray pyrometry. For thin foils (<<12 μm), the rear temperatures can be predicted reasonably well with the use of the front x-ray spectra. For thick foils shock preheating is dominant

  17. Design of a DCS Based Model for Continuous Leakage Monitoring System of Rotary Air Preheater of a Thermal Power Plant

    Directory of Open Access Journals (Sweden)

    Madan BHOWMICK

    2011-01-01

    Full Text Available The leakage in rotary air preheater makes a considerable contribution to the reduced overall efficiency of fossil-fuel-fired thermal power plants and increase the effect on environment. Since it is normal phenomenon, continuous monitoring of leakage is generally omitted in most power plants. But for accurate analysis of the operation of the thermal power plant, this leakage monitoring plays a vital role. In the present paper, design of a DCS based model for continuous leakages monitoring of rotary air preheater has been described. In the proposed model, the existing DCS based instrumentation system has been modified and online leakage monitoring system has been developed. This model has been installed in a captive power plant with high capacity boilers and very much satisfactory operation of this system has been observed. The observed online data along with their analysis results are presented in this paper.

  18. An economic and performance design study of solar preheaters for domestic hot water heaters in North Carolina

    Science.gov (United States)

    Jones, C. B.; Smetana, F. O.

    1977-01-01

    The performance and estimated material costs for several solar preheaters for domestic hot water heaters using isolation levels present in North Carolina are presented. The effects of monthly variations in isolation and the direction of incident radiation are included. Demand is assumed at 13 gallons (49.2 liters) per day per person. The study shows that a closed circulation system with 82 gallons (310 liters) of preheated storage and 53.4 cu ft (4.94 cu m) of collector surface with single cover can be expected to cost about $800 and to repay it capital cost and interest (at 8%) in 5.2 years, assuming present electric rates increase at 5% per year.

  19. Modeling of crude oil fouling in preheat exchangers of refinery distillation units

    Energy Technology Data Exchange (ETDEWEB)

    Jafari Nasr, Mohammad Reza; Majidi Givi, Mehdi [National Petrochemical Research and Technology Company (NPC-RT), P.O. Box 14385, Tehran (Iran)

    2006-10-15

    The aim of this paper is to propose a new model for crude oil fouling in preheat exchangers of crude distillation units. The experimental results of Australian light crude oil with the tube side surface temperature between 200 and 260{sup o}C and fluid velocity ranged 0.25-0.4m/s were used [Z. Saleh, R. Sheikholeslami, A.P. Watkinson, Heat exchanger fouling by a light australian crude oil, in: Heat Exchanger Fouling and Cleaning Fundamentals and Applications, Santa Fe, 2003]. The amount of activation energy depends on the surface temperature has been calculated. A new model including a term for fouling formation and a term for fouling removal due to chemical and tube wall shear stress was proposed, respectively. The main superiority of the model are independent to Pr number, thermal fouling removal and determination of {beta} based on experimental tests. Finally using the proposed model the fouling rate of Australian light crude oil has been calculated and the threshold curves to identify fouling and no fouling formation zones have been drawn. (author)

  20. Effect of substrate preheating temperature and coating thickness on residual stress in plasma sprayed hydroxyapatite coating

    International Nuclear Information System (INIS)

    Tang, Dapei

    2015-01-01

    A thermal-mechanical coupling model was developed based on thermal-elastic- plastic theory according the special process of plasma spraying Hydroxyapatite (HA) coating upon Ti-6Al-4V substrate. On the one hand, the classical Fourier transient heat conduction equation was modified by introducing the effect item of deformation on temperature, on the other hand, the Johnson-Cook model, suitable for high temperature and high strain rate conditions, was used as constitutive equation after considering temperature softening effect, strain hardening effect and strain rate reinforcement effect. Based on the above coupling model, the residual stress field within the HA coating was simulated by using finite element method (FEM). Meanwhile, the substrate preheating temperature and coating thickness on the influence of residual stress components were calculated, respectively. The failure modes of coating were also preliminary analyzed. In addition, in order to verify the reliability of calculation, the material removal measurement technique was applied to determine the residual stress of HA coating near the interface. Some important conclusions are obtained. (paper)

  1. Flat plate solar collector for water pre-heating using concentrated solar power (CSP)

    Science.gov (United States)

    Peris, Leonard Sunny; Shekh, Md. Al Amin; Sarker, Imran

    2017-12-01

    Numerous attempt and experimental conduction on different methods to harness energy from renewable sources are being conducted. This study is a contribution to the purpose of harnessing solar energy as a renewable source by using flat plate solar collector medium to preheat water. Basic theory of solar radiation and heat convection in water (working fluid) has been combined with heat conduction process by using copper tubes and aluminum absorber plate in a closed conduit, covered with a glazed through glass medium. By this experimental conduction, a temperature elevation of 35°C in 10 minutes duration which is of 61.58% efficiency range (maximum) has been achieved. The obtained data and experimental findings are validated with the theoretical formulation and an experimental demonstration model. A cost effective and simple form of heat energy extraction method for space heating/power generation has been thoroughly discussed with possible industrial implementation possibilities. Under-developed and developing countries can take this work as an illustration for renewable energy utilization for sustainable energy prospect. Also a full structure based data to derive concentrated solar energy in any geographical location of Bangladesh has been outlined in this study. These research findings can contribute to a large extent for setting up any solar based power plant in Bangladesh irrespective of its installation type.

  2. Synthesis and physical properties of zinc-oxide textured films by using a filtered preheated hydrothermal

    International Nuclear Information System (INIS)

    Qiu, Jijun; Shin, Dongmyeong; He, Weizhen; Kim, Hyungkook; Hwang, Yoonhwae; Li, Xiaomin; Gao, Xiangdong

    2014-01-01

    Axially (c-axis)-oriented ZnO thick films with a ∼8.1 μm thickness were fabricated on ZnO seed layer coated substrates by using a filtered preheated hydrothermal solution. The thick films composed of single-crystal ZnO microrods with various diameters were formed by coalescing each nanorod together along their side surfaces. From the X-ray diffraction result a biaxial stress exists was found to exist in the as-grown thick films, and the stress gradually increased with increasing annealing temperatures from 200 to 550 .deg. C due to a degradation in the crystalline quality. The biaxial stress is responsible for the red-shift of the optical band gap of the ZnO thick films. Photoluminescence and Hall results revealed that the optical and the electrical properties of the thick films were degenerated after high-temperature annealing (> 200 .deg. C), which was due to the introduction of point defects, such as oxygen interstitials and zinc vacancies.

  3. Ways to achieve optimum utilization of waste gas heat in cement kiln plants with cyclone preheaters

    Energy Technology Data Exchange (ETDEWEB)

    Steinbiss, E

    1986-02-01

    Kiln exit gases and the exhaust gases from clinker coolers often cannot be fully utilized in drying plants. In such cases a part of the heat content of the gases should be utilized for water heating. In addition, it is possible to utilize the waste gas heat in conventional steam boilers, with which, depending on design, it is possible to generate electricity at a rate of between 10-30 kWh/t (net output). A new and promising method of utilization of waste gas heat is provided by precalcining systems with bypass, in which up to 100% of the kiln exit gases can be economically bypassed and be utilized in a steam boiler, without requiring any cooling. A development project, already started, gives information on the operational behaviour of such a plant and on the maximum energy recoverable. Alternatively, the bypass gases may, after partial cooling with air or preheater exit gas, be dedusted and then utilized in a grinding/drying plant. Furthermore, they can be used in the cement grinding process for the drying of wet granulated blastfurnace slag or other materials. For this it is not necessary to dedust the bypass gases.

  4. Synthesis and physical properties of zinc-oxide textured films by using a filtered preheated hydrothermal

    Energy Technology Data Exchange (ETDEWEB)

    Qiu, Jijun [Pusan National University, Busan (Korea, Republic of); Shin, Dongmyeong; He, Weizhen; Kim, Hyungkook; Hwang, Yoonhwae [Pusan National University, Miryang (Korea, Republic of); Li, Xiaomin; Gao, Xiangdong [Chinese Academy of Sciences, Shanghai (China)

    2014-11-15

    Axially (c-axis)-oriented ZnO thick films with a ∼8.1 μm thickness were fabricated on ZnO seed layer coated substrates by using a filtered preheated hydrothermal solution. The thick films composed of single-crystal ZnO microrods with various diameters were formed by coalescing each nanorod together along their side surfaces. From the X-ray diffraction result a biaxial stress exists was found to exist in the as-grown thick films, and the stress gradually increased with increasing annealing temperatures from 200 to 550 .deg. C due to a degradation in the crystalline quality. The biaxial stress is responsible for the red-shift of the optical band gap of the ZnO thick films. Photoluminescence and Hall results revealed that the optical and the electrical properties of the thick films were degenerated after high-temperature annealing (> 200 .deg. C), which was due to the introduction of point defects, such as oxygen interstitials and zinc vacancies.

  5. Combustion analysis of preheated crude sunflower oil in an IDI diesel engine

    Energy Technology Data Exchange (ETDEWEB)

    Canakci, Mustafa; Ozsezen, Ahmet Necati; Turkcan, Ali [Department of Mechanical Education, Kocaeli University, 41380 Izmit (Turkey); Alternative Fuels R and D Center, Kocaeli University, 41040 Izmit (Turkey)

    2009-05-15

    In this study, preheated crude sunflower oil (PCSO) was tested for combustion and emission properties against petroleum based diesel fuel (PBDF) in a naturally aspirated, indirect injection (IDI) engine. The cylinder gas pressure and heat release curves for PCSO at 75 C were similar to those of PBDF. The ignition delays for the PCSO were longer and the start of injection timing was earlier than for PBDF. The difference in the average brake torque was a decrease of 1.36% for PCSO though this was statistically insignificant. The brake specific fuel consumption increased by almost 5% more or less in proportion to the difference in calorific value, so that the 1.06% increase in thermal efficiency was again statistically insignificant. The emission test results showed that the decreases in CO{sub 2} emissions and smoke opacity 2.05% and 4.66%, respectively; however, this was not statistically significant, though in line with the apparent increase in thermal efficiency. There was a significant 34% improvement in the emissions of unburnt hydrocarbons. Carbon monoxide increased by 1.77% again the result was not statistically significant given the small number of repeat tests. The use of PCSO does not have any negative effects on the engine performance and emissions in short duration engine testing. (author)

  6. Useful work and the thermal efficiency in the ideal Lenolr cycle with regenerative preheating

    Science.gov (United States)

    Georgiou, Demos P.

    2000-11-01

    In the existing thermal engine concepts negative work transfer (usually needed to drive a compression process) is supplied by the work produced by the engine itself. The remaining difference (i.e., the net work transfer) becomes the useful work, since it is available for external consumption. The thermal efficiency is the parameter that compares this against the heat input into the system. It forms the main optimization parameter in any engine design. The objective of the present study is to show that for the case of the Lenoir cycle with regenerative preheating the entire positive work is available for external consumption, since the negative (i.e., the compression) work is supplied by the atmospheric air. Not only this, but, during the compression process and due to the pressure difference across the two sides of the moving piston, an additional (useful) work transfer may be generated. Thus, the proposed power plant may be considered as a combination of a thermal engine and a wind turbine. In the ideal cycle limit (at least), the total amount of useful work exceeds the heat entering the system. This leads to the definition of a new parameter for the efficiency (called the technical efficiency), which compares the combined positive work transfer (i.e., the useful one) against the heat entering the system and which may exceed the 100% level.

  7. Application of ground-to-air heat exchanger for preheating of supply air

    Science.gov (United States)

    Sorokins, Juris; Borodinecs, Anatolijs; Zemitis, Jurgis

    2017-10-01

    This study focuses on assessing the contribution of the passive ground-coupled air heat exchanger system to decreasing the energy consumption of air conditioning and ventilation systems for office buildings in the Latvian climate conditions. The theoretical part of the thesis deals with methods of office building ventilation, supply air preheating and heat recovery as well as particularities of using ground-coupled air heat exchangers, their design parameters and their joint impact on the thermal performance. The engineering project part includes a ventilation system for an office building with an integrated ground-coupled air heat exchanger. By simulating energy consumption of the ventilation system for a duration of one year, the thesis analyzes the contribution of the heat exchanger to the overall energy consumption, which totals 9.53 MWh and 4.02 MWh a year, according to the desired parameters of the indoor climate. The possible alternative heat recovery solutions are investigated to reach by European Regional Development Fund project Nr.1.1.1.1/16/A/048 “NEARLY ZERO ENERGY SOLUTIONS FOR UNCLASSIFIED BUILDINGS”.

  8. FAILURE ANALYSIS IN TUBING OF AIR PREHEATER OF BOILER FROM A SUGARCANE MILL

    Directory of Open Access Journals (Sweden)

    Joner Oliveira Alves

    2014-10-01

    Full Text Available The increased demand for energy from sugarcane bagasse has made the sugar and alcohol mills search alternatives to reduce maintenance of the boilers, releasing more time to the production. The stainless steel use has become one of the main tools for such reduction. However, specification errors can lead to premature failures. This work reports the factors that led tubes of AISI 409 stainless steel fail after half season when applied in a air preheater of boiler from a sugarcane mill. In such application, the AISI 304 lasts about 15 seasons and the carbon steel about 3. A tube sent by the sugar mill was characterized by wet chemical analysis, optical microscopy and EDS. Results indicated chloride formation on the internal walls of the tube, which combined with the environment, accelerated the corrosion process. The carbon steel showed high lifetime due to a 70% higher thickness. Due to the work condictions is recommended the use of stainless steels with higher corrosion resistance, such as the traditional AISI 304 or the ferritic AISI 444, the last presents better thermal exchange.

  9. Establishment of welding process without PWHT and preheating in SGV480 plate for nuclear reactor containment vessel

    International Nuclear Information System (INIS)

    Watanabe, Nozomu; Higashikubo, Tomohiro; Nagamura, Takafumi; Yoshimoto Kentaro

    2000-01-01

    Ordinances of Japan's Ministry of International Trade and Industry provide that welded joints more than 38 mm thick used in nuclear reactor containment vessels undergo Post Weld Heat Treatment (PWHT). PWHT is difficult to apply in the field, however. We made SGV480 plate tougher and more weldable by using a Thermo-Mechanical Control Process (TMCP) in rolling. Such plate can be used without PWHT or preheating up to 55 mm thick at lowest service temperature -19degC. (author)

  10. Effect of pre-heating on the chemical oxidation efficiency: implications for the PAH availability measurement in contaminated soils.

    Science.gov (United States)

    Biache, Coralie; Lorgeoux, Catherine; Andriatsihoarana, Sitraka; Colombano, Stéfan; Faure, Pierre

    2015-04-09

    Three chemical oxidation treatments (KMnO4, H2O2 and Fenton-like) were applied on three PAH-contaminated soils presenting different properties to determine the potential use of these treatments to evaluate the available PAH fraction. In order to increase the available fraction, a pre-heating (100 °C under N2 for one week) was also applied on the samples prior oxidant addition. PAH and extractable organic matter contents were determined before and after treatment applications. KMnO4 was efficient to degrade PAHs in all the soil samples and the pre-heating slightly improved its efficiency. H2O2 and Fenton-like treatments presented low efficiency to degrade PAH in the soil presenting poor PAH availability, however, the PAH degradation rates were improved with the pre-heating. Consequently H2O2-based treatments (including Fenton-like) are highly sensitive to contaminant availability and seem to be valid methods to estimate the available PAH fraction in contaminated soils. Copyright © 2014 Elsevier B.V. All rights reserved.

  11. Effect of bond coat and preheat on the microstructure, hardness, and porosity of flame sprayed tungsten carbide coatings

    Science.gov (United States)

    Winarto, Winarto; Sofyan, Nofrijon; Rooscote, Didi

    2017-06-01

    Thermally sprayed coatings are used to improve the surface properties of tool steel materials. Bond coatings are commonly used as intermediate layers deposited on steel substrates (i.e. H13 tool steel) before the top coat is applied in order to enhance a number of critical performance criteria including adhesion of a barrier coating, limiting atomic migration of the base metal, and corrosion resistance. This paper presents the experimental results regarding the effect of nickel bond coat and preheats temperatures (i.e. 200°C, 300°C and 400°C) on microstructure, hardness, and porosity of tungsten carbide coatings sprayed by flame thermal coating. Micro-hardness, porosity and microstructure of tungsten carbide coatings are evaluated by using micro-hardness testing, optical microscopy, scanning electron microscopy, and X-ray diffraction. The results show that nickel bond coatings reduce the susceptibility of micro crack formation at the bonding area interfaces. The percentage of porosity level on the tungsten carbide coatings with nickel bond coat decreases from 5.36 % to 2.78% with the increase of preheat temperature of the steel substrate of H13 from 200°C to 400°C. The optimum hardness of tungsten carbide coatings is 1717 HVN in average resulted from the preheat temperature of 300°C.

  12. Influence of preheating on API 5L-X80 pipeline joint welding with self shielded flux-cored wire

    International Nuclear Information System (INIS)

    Cooper, R.; Silva, J. H. F.; Trevisan, R. E.

    2004-01-01

    The present work refers to the characterization of API 5L-X80 pipeline joints welded with self-shielded flux cored wire. This process was evaluated under preheating conditions, with an uniform and steady heat input. All joints were welded in flat position (1G), with the pipe turning and the torch still. Tube dimensions were 762 mm in external diameter and 16 mm in thickness. Welds were applied on single V-groove, with six weld beads, along with three levels of preheating temperatures (room temperature, 100 degree centigree, 160 degree centigree). These temperatures were maintained as inter pass temperature. The filler metal E71T8-K6 with mechanical properties different from parent metal was used in under matched conditions. The weld characterization is presented according to the mechanical test results of tensile strength, hardness and impact test. The mechanical tests were conducted according to API 1104, AWS and ASTM standards. API 1104 and API 51 were used as screening criteria. According to the results obtained, it was possible to remark that it is appropriate to weld API 5L-X80 steel ducts with Self-shielded Flux Cored wires, in conformance to the API standards and no preheat temperature is necessary. (Author) 22 refs

  13. APT 3He target/blanket. Topical report

    International Nuclear Information System (INIS)

    1995-03-01

    The 3 He 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 D 2 O. 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. Improved modules for the blanket of RTO/RC ITER

    International Nuclear Information System (INIS)

    Elio, F.; Ioki, K.; Cardella, A.

    2000-01-01

    This paper describes innovative design aspects that are considered to optimise the blanket modules for the reduced technical objective/reduced cost international thermonuclear experimental reactor. The blanket modules have a vertical straight profile facing the plasma, and the first wall is built in small and flat panels. Copper may be applied only in front of the first row of cooling passages. The radial cooling of the shield block avoids a complex by-pass at the back and opens up the possibility to use cast instead of forged steel. Slits in the shield block and in the first wall reduce the electromagnetic forces enough to allow the support of the modules on the vessel and the mechanical attachment of the first wall panels

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  16. Blanket handling concepts for future fusion power plants

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    Science.gov (United States)

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

    1990-01-01

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

  18. Structural performance of a graphite blanket in fusion reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  19. Stability properties of cold blanket systems for current driven modes

    International Nuclear Information System (INIS)

    Ohlsson, D.

    1977-12-01

    The stability problem of the boundary regions of cold blanket systems with induced currents parallel to the lines of force is formulated. Particular interest is focused on two types of modes: first electrostatic modes driven by the combined effects of a transverse resistivity gradient due to a spatially non-uniform electron temperature and a longitudinal current, second electromagnetic kink like modes driven by the torque arising from a transverse current density gradient and magnetic field perturbations. It is found that the combination of various dissipative and neutral gas effects introduces strong stabilizing effects within specific parameter ranges. For particular steady-state models investigated it is shown that these effects become of importance in laboratory plasmas at relatively high densities, low temperatures and moderate magnetic field strengths. Stability diagrams based on specific steady-state cold plasma blanket models will be presented

  20. APT {sup 3}He target/blanket. Topical report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    The {sup 3}He 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 D{sub 2}O. 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.

  1. Engineering test station for TFTR blanket module experiments

    International Nuclear Information System (INIS)

    Jassby, D.L.; Leinoff, S.

    1979-12-01

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

  2. Hard x-ray (>100 keV) imager to measure hot electron preheat for indirectly driven capsule implosions on the NIF.

    Science.gov (United States)

    Döppner, T; Dewald, E L; Divol, L; Thomas, C A; Burns, S; Celliers, P M; Izumi, N; Kline, J L; LaCaille, G; McNaney, J M; Prasad, R R; Robey, H F; Glenzer, S H; Landen, O L

    2012-10-01

    We have fielded a hard x-ray (>100 keV) imager with high aspect ratio pinholes to measure the spatially resolved bremsstrahlung emission from energetic electrons slowing in a plastic ablator shell during indirectly driven implosions at the National Ignition Facility. These electrons are generated in laser plasma interactions and are a source of preheat to the deuterium-tritium fuel. First measurements show that hot electron preheat does not limit obtaining the fuel areal densities required for ignition and burn.

  3. RAMI analysis for DEMO HCPB blanket concept cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Dongiovanni, Danilo N., E-mail: danilo.dongiovanni@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Pinna, Tonio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Carloni, Dario [KIT, Institute of Neutron Physics and Reactor Technology (INR) – KIT (Germany)

    2015-10-15

    Highlights: • RAMI (reliability, availability, maintainability and inspectability) preliminary assessment for HCPB blanket concept cooling system. • Reliability block diagram (RBD) modeling and analysis for HCPB primary heat transfer system (PHTS), coolant purification system (CPS), pressure control system (PCS), and secondary cooling system. • Sensitivity analysis on system availability performance. • Failure models and repair models estimated on the base of data from the ENEA fusion component failure rate database (FCFRDB). - Abstract: A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the HCPB (helium cooled pebble bed) blanket cooling system based on currently available design for DEMO fusion power plant is presented. The following sub-systems were considered in the analysis: blanket modules, primary cooling loop including pipework and steam generators lines, pressure control system (PCS), coolant purification system (CPS) and secondary cooling system. For PCS and CPS systems an extrapolation from ITER Test Blanket Module corresponding systems was used as reference design in the analysis. Helium cooled pebble bed (HCPB) system reliability block diagrams (RBD) models were implemented taking into account: system reliability-wise configuration, operating schedule currently foreseen for DEMO, maintenance schedule and plant evolution schedule as well as failure and corrective maintenance models. A simulation of plant activity was then performed on implemented RBDs to estimate plant availability performance on a mission time of 30 calendar years. The resulting availability performance was finally compared to availability goals previously proposed for DEMO plant by a panel of experts. The study suggests that inherent availability goals proposed for DEMO PHTS system and Tokamak auxiliaries are potentially achievable for the primary loop of the HCPB concept cooling system, but not for the secondary loop. A

  4. 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.

  5. Tritium breeding blanket device of D-T reactors

    International Nuclear Information System (INIS)

    Chevereau, G.

    1984-01-01

    This blanket device uses solid tritium breeding materials as those which include, in a known manner, near a neutron breeding plasma, a neutron multiplier medium and a tritium breeding medium, cooled by a cooling fluid circulation. This device is characterized by the fact that the association of the multiplier media and the tritium breeding media is realized by pellet alternated piling up of each of those both media, help in close contact on all their lateral surfaces [fr

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

    International Nuclear Information System (INIS)

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    1984-09-01

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

  8. Evaluation of US demo helium-cooled blanket options

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  9. Heat Loads Due To Small Penetrations In Multilayer Insulation Blankets

    Science.gov (United States)

    Johnson, W. L.; Heckle, K. W.; E Fesmire, J.

    2017-12-01

    The main penetrations (supports and piping) through multilayer insulation systems for cryogenic tanks have been previously addressed by heat flow measurements. Smaller penetrations due to fasteners and attachments are now experimentally investigated. The use of small pins or plastic garment tag fasteners to ease the handling and construction of multilayer insulation (MLI) blankets goes back many years. While it has long been understood that penetrations and other discontinuities degrade the performance of the MLI blanket, quantification of this degradation has generally been lumped into gross performance multipliers (often called degradation factors or scale factors). Small penetrations contribute both solid conduction and radiation heat transfer paths through the blanket. The conduction is down the stem of the structural element itself while the radiation is through the hole formed during installation of the pin or fastener. Analytical models were developed in conjunction with MLI perforation theory and Fourier’s Law. Results of the analytical models are compared to experimental testing performed on a 10 layer MLI blanket with approximately 50 small plastic pins penetrating the test specimen. The pins were installed at ∼76-mm spacing inches in both directions to minimize the compounding of thermal effects due to localized compression or lateral heat transfer. The testing was performed using a liquid nitrogen boil-off calorimeter (Cryostat-100) with the standard boundary temperatures of 293 K and 78 K. Results show that the added radiation through the holes is much more significant than the conduction down the fastener. The results are shown to be in agreement with radiation theory for perforated films.

  10. New concepts for controlled fusion reactor blanket design

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Avci, H.; El-Maghrabi, M.

    1975-01-01

    Several new concepts for fusion reactor blanket design based on the idea of shifting, or tailoring, the neutron spectrum incident on the first structural wall are presented. The spectral shifter is a nonstructural element which can be made of graphite, silicon carbide, or three dimensionally woven carbon fibers (and containing other materials as appropriate) placed between the neutron source and the first structural wall. The softened neutron spectrum incident on the structural components leads to lower gas production and atom displacement rates than in more standard fusion blanket designs. In turn, this results in longer anticipated lifetimes for the structural materials and can significantly reduce radioactivity and afterheat levels. In addition, the neutron spectrum in the first structural wall can be made to approach the flux shape in fast breeder reactors. Such spectral softening means that existing radiation facilities may be more profitably used to provide relevant materials radiation damage data for the structural materials in these fusion blanket designs. This general class of blanket concepts are referred to as internal spectral shifter and energy converter, or ISSEC concepts. These specific design concepts fall into three main categories: ISSEC/EB concepts based on utilizing existing designs which breed tritium behind the first structural wall; ISSEC/IB concepts based on breeding tritium inside the first vacuum wall; and ISSEC/Bu concepts based on using boron, carbon, and perhaps, beryllium to obtain an energy multiplier and converter design that does not attempt to breed tritium or utilize lithium. The detailed analyses relate specifically to the nuclear performance of ISSEC systems and to a discussion of materials radiation damage problems in the structural material.(U.S.)

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

    International Nuclear Information System (INIS)

    1984-09-01

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

  12. Choice of economical optimum blanket of hybrid reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blinkin, V L; Novikov, V M

    1981-01-01

    The economical effectiveness of symbiotic power systems depends on the choice of the correlation between energy production and fissile fuel production in blankets of controlled thermonuclear fusion reactor (CTR), what is investigated here. It is shown that the optimum value of this correlation essentially depends on the ratio between the specific costs for energy production in hybrid thermonuclear reactors and that in fission reactors as part of the symbiotic system.

  13. Development of insulating coatings for liquid metal blankets

    International Nuclear Information System (INIS)

    Malang, S.; Borgstedt, H.U.; Farnum, E.H.; Natesan, K.; Vitkovski, I.V.

    1994-07-01

    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

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

    International Nuclear Information System (INIS)

    Tone, Tatsuzo; Iida, Hiromasa; Tanaka, Yoshihisa

    1984-01-01

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

  15. Neutronic performance issues of the breeding blanket options for the European DEMO fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion—Programme Management Unit, Boltzmannstr. 2, 85748 Garching (Germany); Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, SERMA, LPEC, 91191 Gif-sur-Yvette (France); Moro, F. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy)

    2016-11-01

    Highlights: • Breeder blanket concepts for DEMO—design features. • Neutronic characteristics of breeder blankets. • Evaluation of Tritium breeding potential. • Evaluation of shielding performance. - Abstract: This paper presents nuclear performance issues of the HCPB, HCLL, DCLL and WCLL breeder blankets, which are under development within the PPPT (Power Plant Physics and Technology) programme of EUROfusion, with the objective to assess the potential and suitability of the blankets for the application to DEMO. The assessment is based on the initial design versions of the blankets developed in 2014. The Tritium breeding potential is considered sufficient for all breeder blankets although the initial design versions of the HCPB, HCLL and DCLL blankets were shown to require further design improvements. Suitable measures have been proposed and proven to be sufficient to achieve the required Tritium Breeding Ratio (TBR) ≥ 1.10. The shielding performance was shown to be sufficient to protect the super-conducting toroidal field coil provided that efficient shielding material mixtures including WC or borated water are utilized. The WCLL blanket does not require the use of such shielding materials due to a very compact blanket support structure/manifold configuration which yet requires design verification. The vacuum vessel can be safely operated over the full anticipated DEMO lifetime of 6 full power years for all blanket concepts considered.

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

    International Nuclear Information System (INIS)

    Daly, E.; Pletzer, R.

    1992-04-01

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

  17. Moving ring field-reversed mirror blanket design considerations

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, L.; Kessel, C.; Norman, J.; Schultz, K.R.

    1981-01-01

    A blanket design for the Moving Ring Field-Reversed Mirror Reactor (MRFRM) is presented in this paper. The design emphasis is placed on minimizing the induced radioactivities in the first-wall, blanket and shield. To this end, aluminum-alloy was selected as the reference structural material, giving dose rates two weeks after shutdown that are 3 to 4 orders of magnitude lower than comparable steel structures. The aluminum first-wall is water-cooled and thermally insulated from the high temperature SiC-clad Li 2 O tritium breeding zone. A local tritium breeding ratio of 1.05 was obtained for the design. The tritium is extracted from the Li 2 O by the use of a small dry helium purge stream through the SiC tubes. About 1 ppM hydrogen is added to the helium purge stream to enhance the tritium recovery rate. Helium at 28 atmospheres pressure is circulated through the blanket and shield, with an outlet temperature of 850 0 C, which is coupled with an existing small size closed-cycle gas turbine (CCGT) power conversion system. The spatial and temporal variations of the first-wall temperature caused by the translational movement of the plasma rings along the axis of the cylindrical reactor were evaluated. The after-heat cooling problems of the first-wall were also considered

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

    International Nuclear Information System (INIS)

    Logan, B.G.

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    Science.gov (United States)

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

    1993-01-01

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

  1. Effect of blanket assembly shuffling on LMR neutronic performance

    International Nuclear Information System (INIS)

    Khalil, H.; Fujita, E.K.

    1987-01-01

    Neutronic analyses of advanced liquid-metal reactors (LMRs) have generally been performed with assemblies in different batches scatter-loaded but not shuffled among the core lattice positions between cycles. While this refueling approach minimizes refueling time, significant improvements in thermal performance are believed to be achievable by blanket assembly shuffling. These improvements, attributable to mitigation of the early-life overcooling of the blankets, include reductions in peak clad temperatures and in the temperature gradients responsible for thermal striping. Here the authors summarize results of a study performed to: (1) assess whether the anticipated gains in thermal performance can be realized without sacrificing core neutronic performance, particularly the burnup reactivity swing rho/sub bu/, which determines the rod ejection worth; (2) determine the effect of various blanket shuffling operations on reactor performance; and (3) determine whether shuffling strategies developed for an equilibrium (plutonium-fueled) core can be applied during the transition from an initial uranium-fueled core as is being considered in the US advanced LMR program

  2. Investigation of aqueous slurries as fusion reactor blankets

    International Nuclear Information System (INIS)

    Schuller, M.J.

    1985-01-01

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

  3. NOEL: a no-leak fusion blanket concept

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  4. Neutronic investigation and activation calculation for CFETR HCCB blankets

    Science.gov (United States)

    Shuling, XU; Mingzhun, LEI; Sumei, LIU; Kun, LU; Kun, XU; Kun, PEI

    2017-12-01

    The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder (HCCB) blanket were predicted for the Chinese Fusion Engineering Testing Reactor (CFETR) design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio (TBR). Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil. The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1 × 10-4 kW, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.

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

    International Nuclear Information System (INIS)

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

    2009-10-01

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

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

    International Nuclear Information System (INIS)

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

    1986-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  8. Mirror hybrid reactor blanket and power conversion system conceptual design

    International Nuclear Information System (INIS)

    Schultz, K.R.; Backus, G.A.; Baxi, C.B.; Dee, J.B.; Estrine, E.A.; Rao, R.; Veca, A.R.

    1976-01-01

    The conceptual design of the blanket and power conversion system for a gas-cooled mirror hybrid fusion-fission reactor is presented. The designs of the fuel, blanket module and power conversion system are based on existing gas-cooled fission reactor technology that has been developed at General Atomic Company. The uranium silicide fuel is contained in Inconel-clad rods and is cooled by helium gas. The fuel is contained in 16 spherical segment modules which surround the fusion plasma. The hot helium is used to raise steam for a conventional steam cycle turbine generator. The details of the method of support for the massive blanket modules and helium ducts remain to be determined. Nevertheless, the conceptual design appears to be technically feasible with existing gas-cooled technology. A preliminary safety analysis shows that with the development of a satisfactory method of primary coolant circuit containment and support, the hybrid reactor could be licensed under existing Nuclear Regulatory Commission regulations

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  10. Method of operating water cooled reactor with blanket

    International Nuclear Information System (INIS)

    Suzuki, Katsuo.

    1988-01-01

    Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

  11. Influence of pre-heating on the surface modification of powder-metallurgy processed cold-work tool steel during laser surface melting

    Energy Technology Data Exchange (ETDEWEB)

    Šturm, Roman, E-mail: roman.sturm@fs.uni-lj.si [University of Ljubljana, Faculty of Mechanical Engineering, Aškerčeva 6, 1000 Ljubljana (Slovenia); Štefanikova, Maria [University of Ljubljana, Faculty of Mechanical Engineering, Aškerčeva 6, 1000 Ljubljana (Slovenia); Steiner Petrovič, Darja [Institute of Metals and Technology, Lepi pot 11, 1000 Ljubljana (Slovenia)

    2015-01-15

    Graphical abstract: - Highlights: • Heat-treatment protocol for laser surface melting of cold-work tool steel is proposed. • The laser melted steel surface is hardened, and morphologically modified. • The pre-heating of substrate creates a crack-and pore-free steel surface. • The optimum pre-heating temperature is determined to be 350 °C. • Using pre-heating the quantity of retained austenite is reduced. - Abstract: In this study we determine the optimal parameters for surface modification using the laser surface melting of powder-metallurgy processed, vanadium-rich, cold-work tool steel. A combination of steel pre-heating, laser surface melting and a subsequent heat treatment creates a hardened and morphologically modified surface of the selected high-alloy tool steel. The pre-heating of the steel prior to the laser surface melting ensures a crack- and pore-free modified surface. Using a pre-heating temperature of 350 °C, the extremely fine microstructure, which typically evolves during the laser-melting, became slightly coarser and the volume fraction of retained austenite was reduced. In the laser-melted layer the highest values of microhardness were achieved in the specimens where a subsequent heat treatment at 550 °C was applied. The performed thermodynamic calculations were able to provide a very valuable assessment of the liquidus temperature and, especially, a prediction of the chemical composition as well as the precipitation and dissolution sequence for the carbides.

  12. Energy, exergy, environmental and economic analysis of industrial fired heaters based on heat recovery and preheating techniques

    International Nuclear Information System (INIS)

    Shekarchian, M.; Zarifi, F.; Moghavvemi, M.; Motasemi, F.; Mahlia, T.M.I.

    2013-01-01

    Highlights: • 4-E analysis of a typical industrial grade fired heater unit is studied. • This analysis is accomplished for the first time in this study. • Heat recovery and air preheating lead to substantial reduction in the fuel consumption. • The company’s current costs are tremendously reduced by these methods. • The methods lead to mitigation in GHG emission and to reduction in the associated taxes. - Abstract: Fired heaters are ubiquitous in both the petroleum and petrochemical industries, due to it being vital in their day to day operations. They form major components in petroleum refineries, petrochemical facilities, and processing units. This study was commissioned in order to analyze the economic benefits of incorporating both heat recovery and air preheating methods into the existing fired heater units. Four fired heater units were analyzed from the energy and environmental point of views. Moreover, the second law efficiency and the rate of irreversibility were also analyzed via the exergy analysis. Both analyses was indicative of the fact that the heat recovery process enhances both the first and second law efficiencies while simultaneously assisting in the production of high and low pressure water steam. The implementation and usage of the process improves the thermal and exergy efficiencies from 63.4% to 71.7% and 49.4%, to 54.8%, respectively. Additionally, the heat recovery and air preheating methods leads to a substantial reduction in fuel consumption, in the realm of up to 7.4%, while also simultaneously decreasing heat loss and the irreversibility of the unit. Nevertheless, the results of the economic analysis posits that although utilizing an air preheater unit enhances the thermal performance of the system, due to the air preheater’s capital and maintenance costs, incorporating an air preheater unit to an existing fired heater is not economically justifiable. Furthermore, the results of the sensitivity analysis and payback period

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

    International Nuclear Information System (INIS)

    Norajitra, P.

    1995-06-01

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

  14. Performance and emission study of preheated Jatropha oil on medium capacity diesel engine

    Energy Technology Data Exchange (ETDEWEB)

    Chauhan, Bhupendra Singh; Du Jun, Yong; Lee, Kum Bae [Division of Automobile and Mechanical Engineering, Kongju National University (Korea); Kumar, Naveen [Department of Mechanical Engineering, Delhi Technological University, Bawana Road, Delhi 42 (India)

    2010-06-15

    Diesel engines have proved their utility in transport, agriculture and power sector. Environmental norms and scared fossil fuel have attracted the attention to switch the energy demand to alternative energy source. Oil derived from Jatropha curcas plant has been considered as a sustainable substitute to diesel fuel. However, use of straight vegetable oil has encountered problem due to its high viscosity. The aim of present work is to reduce the viscosity of oil by heating from exhaust gases before fed to the engine, the study of effects of FIT (fuel inlet temperature) on engine performance and emissions using a dual fuel engine test rig with an appropriately designed shell and tube heat exchanger (with exhaust bypass arrangement). Heat exchanger was operated in such a way that it could give desired FIT. Results show that BTE (brake thermal efficiency) of engine was lower and BSEC (brake specific energy consumption) was higher when the engine was fueled with Jatropha oil as compared to diesel fuel. Increase in fuel inlet temperature resulted in increase of BTE and reduction in BSEC. Emissions of NO{sub x} from Jatropha oil during the experimental range were lower than diesel fuel and it increases with increase in FIT. CO (carbon monoxide), HC (hydrocarbon), CO{sub 2} (carbon dioxide) emissions from Jatropha oil were found higher than diesel fuel. However, with increase in FIT, a downward trend was observed. Thus, by using heat exchanger preheated Jatropha oil can be a good substitute fuel for diesel engine in the near future. Optimal fuel inlet temperature was found to be 80 C considering the BTE, BSEC and gaseous emissions. (author)

  15. Seawater feed reverse osmosis preheating appraisal, Part I: leading element performance

    International Nuclear Information System (INIS)

    Karameldin, A.; Saadawy, M.S.

    2006-01-01

    This paper is concerned with the seawater reverse osmosis preheating process, and presents a parametric study of the process. The basic transport equations describing the leading element are exhibited and appraised. The leading element, which governs the whole system performance, is studied and analysed. The incorporated and investigated operating parameters are the feed pressure and the temperature for different feed salt concentrations. In addition, different feed flow rates, effects on permeate flux and permeator salt rejection, together with the permeator recovery, are studied. A seawater membrane of a well-known data, for instance FT30SW380HR, is used to perform the study. The membrane water permeability coefficient K w is determined and correlated. Furthermore, the membrane salt permeability coefficient K s from the manufacturer system analysis program (ROSA) is given and discussed. The transport governing equations are programmed in a way that facilitates the achievement of a realistic parametric study. The results showed that the permeate flux increases significantly as the feed pressure increases. Also, it increases significantly as the feed salt concentration decreases, and also as the feed temperature and pressure increase. Meanwhile, the permeator salt rejection increases significantly as the feed pressure increases, and decreases significantly as the feed temperature increases. The study of the leading element of the array showed that there are constraints that must be considered, such as maximum membrane flux, maximum applied feed pressure, maximum feed flow rate and maximum feed temperature. Therefore, to attain the maximum membrane flux, the applied feed pressure must be lowered when the feed temperature is increased. In the case where the feed temperature is increased from 18 deg.. C to 45 deg.. C, a pressure saving of between 7% and 26% is achieved, according to the feed salt concentration and feed flow rate. (author)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  17. 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.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    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. (paper)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  20. Composite beryllium-ceramics breeder pin elements for a gas cooled solid blanket

    International Nuclear Information System (INIS)

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

    1986-06-01

    Helium coolant have main advantages compared to water for solid blankets. But limitations exist too and the development of attractive helium cooled blankets based on breeder pin assemblies has been essentially made possible by the derivation from recent CEA neutronic studies of an optimized composite beryllium/ceramics breeder arrangement. Description of the proposed toroidal blanket layout for Net is made together with the analysis of its main performance. Merits of the considered composite Be/ceramics breeder elements are discussed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  2. Reducing beryllium content in mixed bed solid-type breeder blankets

    Energy Technology Data Exchange (ETDEWEB)

    Shimwell, J., E-mail: mail@jshimwell.com [Department of Physics and Astronomy, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom); Lilley, S.; Morgan, L.; Packer, L.; Kovari, M.; Zheng, S. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); McMillan, J. [Department of Physics and Astronomy, University of Sheffield, Hicks Building, Hounsfield Road, Sheffield S3 7RH (United Kingdom)

    2016-11-01

    Highlights: • The ratio of breeder ceramic to neutron multiplier of breeder blankets was varied linearly with depth. • Blankets with varying composition were found to perform better than uniform composition breeder blankets. • It was also possible to reduce the amount of beryllium required by the blanket. - Abstract: Beryllium (Be) is a precious resource with many high value uses, the low energy threshold (n,2n) reaction makes Be an excellent neutron multiplier for use in fusion breeder blankets. Estimates of Be requirements and available resources suggest that this could represent a major supply difficulty for solid-type blanket concepts. Reducing the quantity of Be required by breeder blankets would help to alleviate the problem to some extent. In addition, it is important that the reduction in the Be quantity does not diminish the blanket's performance in key aspects such as the tritium breeding ratio (TBR), energy multiplication and peak nuclear heating. Mixed pebble bed designs allow for the multiplier fraction to be varied throughout the blanket. This neutronics study used MCNP 6 to investigate linear variations of the multiplier fraction in relation to blanket depth, in order to better utilise the important multiplying Be(n,2n) and breeding reactions. Blankets with a uniform multiplier fraction showed little scope for reduction in Be mass. Blankets with varying multiplier fractions were able to simultaneously use 10% less Be, increase the energy amplification by 1%, reduce the peak heating by 7% and maintaining a sufficient TBR when compared to the performance achievable using a uniform composition.

  3. Structural design study of tritium breeding blanket with a lead layer as a neutron multiplier

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Kitamura, Kazunori; Minato, Akio; Sakamoto, Hiroki; Yamamoto, Takashi

    1980-12-01

    Thermal and structural design study of a tritium breeding blanket with a lead layer for a International Tokamak Reactor (INTOR) is carried out. Tube in shell type blanket with a lead layer is found to be promising. The volume fraction of structural material in the lead layer can be small enough to keep the neutron multiplication effect of lead. Reasonable value of shell effect is attainable due to lead layer in the front part of the blanket. (author)

  4. Preliminary conceptual design of the blanket and power conversion system for the Mirror Hybrid Reactor

    International Nuclear Information System (INIS)

    Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

    1978-01-01

    A conceptual design of a commercial Mirror Hybrid Reactor, optimized for 239 Pu production, has been completed. This design is the product of a joint effort by Lawrence Livermore Laboratory and General Atomic Company, and follows directly from earlier work on the Mirror Hybrid. This paper describes the blanket and power conversion system of the reactor design. Included are descriptions of the prestressed concrete reactor vessel that supports the magnets and contains the blanket and power conversion system components, the blanket module design, the blanket fuel design, and the power conversion system

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

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Y.; Tobita, K.; Utoh, H.; Hoshino, K.; Asakura, N.; Nakamura, M.; Tanigawa, H.; Mikio, E.; Tanigawa, H.; Nakamichi, M.; Hoshino, T., E-mail: someya.yoji@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan)

    2012-09-15

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

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    International Nuclear Information System (INIS)

    Chen Yixue; Wu Yican

    2000-01-01

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

  8. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-11-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable.

  9. Design and fabrication methods of FW/blanket and vessel for ITER-FEAT

    International Nuclear Information System (INIS)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Kalinin, G.; Miki, N.; Onozuka, M.; Osaki, T.; Rozov, V.; Sannazzaro, G.; Utin, Y.; Yamada, M.; Yoshimura, H.

    2001-01-01

    Design has progressed on the vacuum vessel and FW/blanket for ITER-FEAT. The basic functions and structures are the same as for the 1998 ITER design. Detailed blanket module designs of the radially cooled shield block with flat separable FW panels have been developed. The ITER blanket R and D program covers different materials and fabrication methods in order make a final selection based on the results. Separate manifolds have been designed and analysed for the blanket cooling. The vessel design with flexible support housings has been improved to minimise the number of continuous poloidal ribs. Most of the R and D performed so far during EDA are still applicable

  10. 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.)

  11. Conceptual design and analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Li, Min; Lv, Zhongliang; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Wang, Xiaoliang; Zheng, Jie; Ye, Minyou

    2015-10-15

    Highlights: • A helium cooled solid blanket was proposed as a candidate blanket concept for CFETR. • Material selection, basic structure and gas flow scheme of the blanket were introduced. • A series of performance analyses for the blanket were summarized. - Abstract: To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and is being designed mainly to demonstrate 50–200 MW fusion power, 30–50% duty time factor, tritium self-sustained. Because of the high demand of tritium production and the realistic engineering consideration, the design of tritium breeding blanket for CFETR is a challenging work and getting special attention. As a blanket candidate, a helium cooled solid breeder blanket has been designed with the emphasis on conservative design and realistic blanket technology. This paper introduces the basic blanket scheme, including the material selection, structural design, cooling scheme and purge gas flow path. In addition, some results of neutronics, thermal-hydraulic and stress analysis are presented.

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

    International Nuclear Information System (INIS)

    Deng Baiquan

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  14. Tritium inventory and permeation in the ITER breeding blanket

    International Nuclear Information System (INIS)

    Violante, V.; Tosti, S.; Sibilia, C.; Felli, F.; Casadio, S.; Alvani, C.

    2000-01-01

    A model has allowed us to perform the analysis of the tritium inventory and permeation in the international thermonuclear experimental reactor (ITER) breeding blanket under the hypothesis of steady state conditions. Li 2 ZrO 3 (reference) and Li 2 TiO 3 (alternative) have been studied as breeding materials. The total breeder inventory assessed is 7.64 g for the Li 2 ZrO 3 at reference temperature. The model has also been used for a parametric analysis of the tritium permeation. At reference temperature and purge helium velocity of 0.01 m/s, the HT partial pressure is ranging from 10 to 30 Pa in the breeder and 1.5x10 -3 Pa in the beryllium. At 0.1 m/s of purge helium velocity, the HT partial pressure is reduced of one order by magnitude in the breeder and becomes 5x10 -5 Pa in the beryllium. The tritium permeation into the coolant for the whole blanket is ranging from 100 to 250 mCi per day for purge helium velocity of 0.01 m/s. The analysis of the tritium inventory and permeation for the alternative Li 2 TiO 3 breeding material has been carried out too. The tritium inventory in the breeder is in the range from 6 to 375 g larger than in Li 2 ZrO 3 by about a factor 5; the tritium permeation into coolant is comparable to the Li 2 ZrO 3 one. This analysis provides indications on the influence of the operating parameters on the tritium control in the ITER breeding blanket; particularly the control of the tritium inventory by the temperature and the tritium permeation by the purge gas velocity

  15. The current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  16. Detailed 3-D nuclear analysis of ITER outboard blanket modules

    International Nuclear Information System (INIS)

    Bohm, Tim; Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul; Ulrickson, Michael; Bullock, James

    2015-01-01

    Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

  17. Detailed 3-D nuclear analysis of ITER outboard blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, Tim, E-mail: tdbohm@wisc.edu [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, Michael; Bullock, James [Formerly, Fusion Technology, Sandia National Laboratories, Albuquerque, NM (United States)

    2015-10-15

    Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

  18. Effect of Pre-heating on Microtensile Bond Strength of Composite Resin to Dentin.

    Directory of Open Access Journals (Sweden)

    Abdolrahim Davari

    2014-10-01

    Full Text Available Direct composite resin restorations are widely used and the impact of different storage temperatures on composites is not well understood. The purpose of this study was to evaluate the microtensile bond strength of composite to dentin after different pre-curing temperatures.Occlusal surfaces of 44 human molars were ground with diamond burs under water coolant and polished with 600 grit silicon carbide papers to obtain flat dentin surfaces. The dentin was etched with 37% phosphoric acid and bonded with Adper Single Bond 2 according to the manufacturer's instructions. The specimens were randomly divided into two groups (n=22 according to the composite resin applied: FiltekP60 and Filtek Z250. Each group included three subgroups of composite resin pre-curing temperatures (4°C, 23°C and 37°C. Composite resins were applied to the dentin surfaces in a plastic mold (8mm in diameter and 4mm in length incrementally and cured. Twenty-two composite-to-dentin hour-glass sticks with one mm(2 cross-sectional area per group were prepared. Microtensile bond strength measurements were made using a universal testing machine at a crosshead speed of one mm/min. For statistical analysis, t-test, one-way and two-way ANOVA were used. The level of significance was set at P<0.05.Filtek P60 pre-heated at 37ºC had significantly higher microtensile bond strength than Filtek Z250 under the same condition. The microtensile bond strengths were not significantly different at 4ºC, 23ºC and 37ºC subgroups of each composite resin group.Filtek P60 and Filtek Z250 did not have significantly different microtensile bond strengths at 4ºC and 23ºC but Filtek P60 had significantly higher microtensile bond strength at 37 ºC. Composite and temperature interactions had significant effects on the bond strength.

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

    Science.gov (United States)

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

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  1. First wall fusion blanket temperature variation - slab geometry

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1978-01-01

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

  2. Accelerator-driven molten-salt blankets: Physics issues

    International Nuclear Information System (INIS)

    Houts, M.G.; Beard, C.A.; Buksa, J.J.; Davidson, J.W.; Durkee, J.W.; Perry, R.T.; Poston, D.I.

    1994-01-01

    A number of nuclear physics issues concerning the Los Alamos molten-salt, accelerator-driven plutonium converter are discussed. General descriptions of several concepts using internal and external, moderation are presented. Burnup and salt processing requirement calculations are presented for four concepts, indicating that both the high power density externally moderated concept and an internally moderated concept achieve total plutonium burnups approaching 90% at salt processing rates of less than 2 m 3 per year. Beginning-of-life reactivity temperature coefficients and system kinetic response are also discussed. Future research should investigate the effect of changing blanket composition on operational and safety characteristics

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    Lub, J.

    1977-01-01

    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

  5. Accelerator-driven molten-salt blankets: Physics issues

    International Nuclear Information System (INIS)

    Houts, M.G.; Beard, C.A.; Buksa, J.J.; Davidson, J.W.; Durkee, J.W.; Perry, R.T.; Poston, D.I.

    1994-01-01

    A number of nuclear physics issues concerning the Los Alamos molten-salt accelerator-driven plutonium converter are discussed. General descriptions of several concepts using internal and external moderation are presented. Burnup and salt processing requirement calculations are presented for four concepts, indicating that both the high power density externally moderated concept and an internally moderated concept achieve total plutonium burnups approaching 90% at salt processing rates of less than 2 m 3 per year. Beginning-of-life reactivity temperature coefficients and system kinetic response are also discussed. Future research should investigate the effect of changing blanket composition on operational and safety characteristics

  6. Conception of divertorless tokamak reactor with turbulent plasma blanket

    International Nuclear Information System (INIS)

    Nedospasov, A.V.; Tokar, M.Z.

    1980-01-01

    The results of the calculations presented here demonstrate that, with technically reasonable degree of the magnetic field stochastisation, the turbulent plasma blanket can take the place of a divertor. It performs the three main functions of the divertor: (a) the exhaust of the helium and unburned fuel; (b) weakening of the fast particle flux to the wall surface; and (c) essential reduction of the impurity content in the active zone of the reactor. Taking into account that plasma flows to the first wall along field lines, we may figuratively say that the first wall plays the role of a divertor in our conception. (orig.)

  7. Tritium Management In HCLL-PPCS Model AB Blanket

    International Nuclear Information System (INIS)

    Ricapito, I.; Aiello, A.; Benamati, G.; Utili, M.; Ciampichetti, A.; Zucchetti, M.

    2006-01-01

    One the main issues in the HCLL blanket development for a prototype fusion reactor is the technical feasibility of the bred tritium processing system. The basis of such concern lies in the very low tritium-Pb17Li Sieverts' constant, as measured by different scientists in the past years. In the PPCS reactor 650 g/d of tritium must be generated in the breeding blanket while less than 1 g/y of tritium has to be released to the environment through the secondary cooling circuit. As a consequence, CPS (Coolant Purification System) plays a fundamental role because it has to keep at an acceptable level the tritium partial pressure in the primary HCS (Helium Cooling Circuit) limiting, therefore, the tritium environmental release through leakage and permeation into the secondary cooling circuit. On the other hand, the He mass flow-rate to be processed by CPS is linear with the tritium permeation rate from the breeder into HCS. Therefore, with the above mentioned low Sieverts' constant values and the consequent high tritium partial pressure in the liquid metal, the possibility to keep acceptable the CPS capacity depends on a highly efficient and stable performance of tritium permeation barriers, to be applied not only on the blanket cooling plates but also on the steam generator walls. However, the experimental results on the tritium permeation barriers under relevant operative conditions were so far quite disappointing. The new data on the Sieverts' constant achieved at ENEA CR Brasimone, one order of magnitude higher than those founding the past, have a big impact in relaxing the above mentioned requirements for the tritium management in PPCS model AB reactor. Besides presenting and discussing these recent experimental results, an updated assessment of the tritium permeation rate from the liquid breeder into HCS through the cooling plates and from HCS into the environment through the steam generators is given in this paper. The consequent new constraints in terms of tritium

  8. Magnetohydrodynamic research in fusion blanket engineering and metallurgical processing

    International Nuclear Information System (INIS)

    Tokuhiro, A.

    1991-11-01

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

  9. Thermal and mechanical design of WITAMIR-I blanket

    International Nuclear Information System (INIS)

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

    1980-10-01

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

  10. Sensitivity and uncertainty analysis of NET/ITER shielding blankets

    International Nuclear Information System (INIS)

    Hogenbirk, A.; Gruppelaar, H.; Verschuur, K.A.

    1990-09-01

    Results are presented of sensitivity and uncertainty calculations based upon the European fusion file (EFF-1). The effect of uncertainties in Fe, Cr and Ni cross sections on the nuclear heating in the coils of a NET/ITER shielding blanket has been studied. The analysis has been performed for the total cross section as well as partial cross sections. The correct expression for the sensitivity profile was used, including the gain term. The resulting uncertainty in the nuclear heating lies between 10 and 20 per cent. (author). 18 refs.; 2 figs.; 2 tabs

  11. Tritium transport in HCLL and WCLL DEMO blankets

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  12. Evaluation of the activity levels in fusion reactor blankets

    International Nuclear Information System (INIS)

    Gruber, J.

    1977-05-01

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

  13. Activation and afterheat analyses for the HCPB test blanket

    International Nuclear Information System (INIS)

    Pereslavtsev, P.; Fischer, U.

    2007-01-01

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

  14. Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shelley, Afroza; Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi

    2003-10-01

    Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40-50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mmx2, internal blanket of 150 mm and axial blanket of 400 mmx2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internal blanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, 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 S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mmx2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature

  15. Establishment of welding process without PWHT and preheating in SGV480 plate for nuclear reactor containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Nozomu; Higashikubo, Tomohiro; Nagamura, Takafumi [Mitsubishi Heavy Industries. Ltd., Kobe Shipyard and Machinery Works (Japan); Yoshimoto Kentaro [Mitsubishi Heavy Industries Ltd., Takasago, Hyogo (Japan). Takasago Research and Development Center

    2000-07-01

    Ordinances of Japan's Ministry of International Trade and Industry provide that welded joints more than 38 mm thick used in nuclear reactor containment vessels undergo Post Weld Heat Treatment (PWHT). PWHT is difficult to apply in the field, however. We made SGV480 plate tougher and more weldable by using a Thermo-Mechanical Control Process (TMCP) in rolling. Such plate can be used without PWHT or preheating up to 55 mm thick at lowest service temperature -19degC. (author)

  16. MgO melting curve constraints from shock temperature and rarefaction overtake measurements in samples preheated to 2300 K

    OpenAIRE

    Fat'yanov, Oleg V.; Asimow, P. D.

    2014-01-01

    Continuing our effort to obtain experimental constraints on the melting curve of MgO at 100-200 GPa, we extended our target preheating capability to 2300 K. Our new Mo capsule design holds a long MgO crystal in a controlled thermal gradient until impact by a Ta flyer launched at up to 7.5 km/s on the Caltech two-stage light-gas gun. Radiative shock temperatures and rarefaction overtake times were measured simultaneously by a 6-channel VIS/NIR pyrometer with 3 ns time resolution. The majority ...

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  18. Comparative studies on the performance and emissions of a direct injection diesel engine fueled with neem oil and pumpkin seed oil biodiesel with and without fuel preheater.

    Science.gov (United States)

    Ramakrishnan, Muneeswaran; Rathinam, Thansekhar Maruthu; Viswanathan, Karthickeyan

    2018-02-01

    In the present experimental analysis, two non-edible oils namely neem oil and pumpkin seed oil were considered. They are converted into respective biodiesels namely neem oil methyl ester (B1) and pumpkin seed oil methyl ester (B2) through transesterification process and their physical and chemical properties were examined using ASTM standards. Diesel was used as a baseline fuel in Kirloskar TV1 model direct injection four stroke diesel engine. A fuel preheater was designed and fabricated to operate at various temperatures (60, 70, and 80 °C). Diesel showed higher brake thermal efficiency (BTE) than biodiesel samples. Lower brake specific fuel consumption (BSFC) was obtained with diesel than B1 sample. B1 exhibited lower BSFC than B2 sample without preheating process. High preheating temperature (80 °C) results in lower fuel consumption for B1 sample. The engine emission characteristics like carbon monoxide (CO), hydrocarbon (HC), and smoke were found lower with B1 sample than diesel and B2 except oxides of nitrogen (NOx) emission. In preheating of fuel, B1 sample with high preheating temperature showed lower CO, HC, and smoke emission (except NOx) than B2 sample.

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

    International Nuclear Information System (INIS)

    Chapin, D.L.

    1976-03-01

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

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

    International Nuclear Information System (INIS)

    1998-02-01

    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.)

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

    International Nuclear Information System (INIS)

    Bond, A.

    1984-01-01

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

  2. 75 FR 46911 - Certain Woven Electric Blankets from the People's Republic of China: Amended Final Determination...

    Science.gov (United States)

    2010-08-04

    ... Blankets from the People's Republic of China: Amended Final Determination of Sales at Less Than Fair Value... than fair value (``LTFV'') in the antidumping investigation of certain woven electric blankets (``woven... From the People's Republic of China: Final Determination of Sales at Less Than Fair Value, 75 FR 38459...

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

    International Nuclear Information System (INIS)

    1993-08-01

    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.)

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

    International Nuclear Information System (INIS)

    1995-09-01

    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)

  5. 77 FR 76013 - Sempra LNG Marketing, LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-12-26

    ... marketing supplies of LNG. Sempra is a customer of the Cameron Terminal. On June 22, 2012, FE issued DOE/FE... DEPARTMENT OF ENERGY [FE Docket No. 12-155-LNG] Sempra LNG Marketing, LLC; Application for Blanket..., by Sempra LNG Marketing, LLC (Sempra LNG Marketing), requesting blanket authorization to export...

  6. 78 FR 4400 - Eni USA Gas Marketing LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2013-01-22

    ... in the business of purchasing and marketing supplies of LNG, and is a customer of the Cameron... DEPARTMENT OF ENERGY [FE Docket No. 12-161-LNG] Eni USA Gas Marketing LLC; Application for Blanket..., by Eni USA Gas Marketing LLC (Eni USA Gas Marketing), requesting blanket authorization to export...

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

    International Nuclear Information System (INIS)

    1994-08-01

    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)

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    Science.gov (United States)

    2012-05-24

    ... 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 501, Birmingham, Alabama 35209, filed... Commission's regulations under the Natural Gas Act (NGA), and Southern's blanket certificate issued in Docket...

  10. Effect of reactor size on the breeding economics of LMFBR blankets

    International Nuclear Information System (INIS)

    Tagishi, A.; Driscoll, M.J.

    1975-02-01

    The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MW(e) were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

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

    1987-07-01

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

  13. 18 CFR 284.284 - Blanket certificates for unbundled sales services.

    Science.gov (United States)

    2010-04-01

    ... 18 Conservation of Power and Water Resources 1 2010-04-01 2010-04-01 false Blanket certificates for unbundled sales services. 284.284 Section 284.284 Conservation of Power and Water Resources... Sales by Interstate Pipelines § 284.284 Blanket certificates for unbundled sales services. (a...

  14. Rotating liquid blanket for a toroidal fusion reator

    International Nuclear Information System (INIS)

    Moir, R.W.

    1987-01-01

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

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

    Science.gov (United States)

    Pippin, Gary; Stuckey, Wayne; Hemminger, Carol

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  17. Uranium self-shielding in fast reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

    Kadiroglu, O.K.; Driscoll, M.J.

    1976-03-01

    The effects of heterogeneity on resonance self-shielding are examined with particular emphasis on the blanket region of the fast breeder reactor and on its dominant reaction--capture in /sup 238/U. The results, however, apply equally well to scattering resonances, to other isotopes (fertile, fissile and structural species) and to other environments, so long as the underlying assumptions of narrow resonance theory apply. The heterogeneous resonance integral is first cast into a modified homogeneous form involving the ratio of coolant-to-fuel fluxes. A generalized correlation (useful in its own right in many other applications) is developed for this ratio, using both integral transport and collision probability theory to infer the form of correlation, and then relying upon Monte Carlo calculations to establish absolute values of the correlation coefficients. It is shown that a simple linear prescription can be developed for the flux ratio as a function of only fuel optical thickness and the fraction of the slowing-down source generated by the coolant. This in turn permitted derivation of a new equivalence theorem relating the heterogeneous self-shielding factor to the homogeneous self-shielding factor at a modified value of the background scattering cross section per absorber nucleus. A simple version of this relation is developed and used to show that heterogeneity has a negligible effect on the calculated blanket breeding ratio in fast reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  19. Ferritic steels for the first generation of breeder blankets

    International Nuclear Information System (INIS)

    Diegele, E.

    2009-01-01

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

  20. Effect of preheating on the damage to tungsten targets after repetitive ITER ELM-like heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Makhlay, V A [Institute of Plasma Physics of the NSC KIPT 1, Akademicheskaya, 61108 Kharkov (Ukraine); Bandura, A N [Institute of Plasma Physics of the NSC KIPT 1, Akademicheskaya, 61108 Kharkov (Ukraine); Byrka, O V [Institute of Plasma Physics of the NSC KIPT 1, Akademicheskaya, 61108 Kharkov (Ukraine); Garkusha, I E [Institute of Plasma Physics of the NSC KIPT 1, Akademicheskaya, 61108 Kharkov (Ukraine); Chebotarev, V V [Institute of Plasma Physics of the NSC KIPT 1, Akademicheskaya, 61108 Kharkov (Ukraine); Tereshin, V I [Institute of Plasma Physics of the NSC KIPT 1, Akademicheskaya, 61108 Kharkov (Ukraine); Landman, I [Forschungszentrum Karlsruhe, IHM, 76021 Karlsruhe (Germany)

    2007-03-15

    The behaviour of a preheated tungsten target under repetitive pulsed plasma impacts of the energy density 0.75 MJ m{sup -2} with the pulse duration of 0.25 ms was studied with the quasi-stationary plasma accelerator (QSPA) Kh-50. Two identical samples of pure sintered tungsten have been exposed to numbers of pulses exceeding 100. One sample was maintained at room temperature and the other sample preheated at 650 deg. C. The experiments demonstrated that on the cold surface some macro-cracks dominate, but on the hot surface they do not develop. However, in both cases some fine meshes of micro-cracks are observed. With increasing the number of exposures, the width of the micro-cracks gradually increases, achieving 0.8-1.5 {mu}m after 100 pulses. In addition, the SEM shows some cellular structure with the cell sizes about 0.3 {mu}m, and after large numbers of exposures some blisters of sizes up to 100-150 {mu}m appear.