REACTOR CONTROL ROD OPERATING SYSTEM

DOEpatents

Miller, G.

1961-12-12

A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS

DOEpatents

Paget, J.A.; Koutz, S.L.; Stone, R.S.; Stewart, H.B.

1963-12-24

An emergency shutdown or scram apparatus for use in a nuclear reactor that includes a neutron absorber suspended from a temperature responsive substance that is selected to fail at a preselected temperature in excess of the normal reactor operating temperature, whereby the neutron absorber is released and allowed to fall under gravity to a preselected position within the reactor core is presented. (AEC)

PROCESS FOR COOLING A NUCLEAR REACTOR

DOEpatents

Borst, L.B.

1962-12-11

This patent relates to the operation of a reactor cooled by liquid sulfur dioxide. According to the invention the pressure on the sulfur dioxide in the reactor is maintained at least at the critical pressure of the sulfur dioxide. Heating the sulfur dioxide to its critical temperature results in vaporization of the sulfur dioxide without boiling. (AEC)

CONTROL SYSTEM

DOEpatents

Shannon, R.H.; Williamson, H.E.

1962-10-30

A boiling water type nuclear reactor power system having improved means of control is described. These means include provisions for either heating the coolant-moderator prior to entry into the reactor or shunting the coolantmoderator around the heating means in response to the demand from the heat engine. These provisions are in addition to means for withdrawing the control rods from the reactor. (AEC)

REFLECTOR FOR NEUTRONIC REACTORS

DOEpatents

Fraas, A.P.

1963-08-01

A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

DOEpatents

Balent, R.

1963-03-12

This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

BOILING REACTORS

DOEpatents

Untermyer, S.

1962-04-10

A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

NEUTRONIC REACTOR

DOEpatents

Fraas, A.P.; Mills, C.B.

1961-11-21

A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

FLOW SYSTEM FOR REACTOR

DOEpatents

Zinn, W.H.

1963-06-11

A reactor is designed with means for terminating the reaction when returning coolant is below a predetermined temperature. Coolant flowing from the reactor passes through a heat exchanger to a lower reservoir, and then circulates between the lower reservoir and an upper reservoir before being returned to the reactor. Means responsive to the temperature of the coolant in the return conduit terminate the chain reaction when the temperature reaches a predetermined minimum value. (AEC)

PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

DOEpatents

Puechl, K.H.

1963-09-24

A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Hauth, J.J.; Anicetti, R.J.

1962-12-01

A method is described for preparing a fuel element for a nuclear reactor. According to the patent uranium dioxide is compacted in a metal tabe by directlng intense sound waves at the tabe prior to tamp packing or vibration compaction of the powder. (AEC)

NON-CORROSIVE REACTOR FUEL SYSTEM

DOEpatents

Herrick, C.C.

1962-08-14

A non-corrosive nuclear reactor fuel system was developed utilizing a molten plutonium-- iron alloy fuel having about 2 at.% carbon and contained in a tantalum vessel. This carbon reacts with the interior surface of the tantalum vessel to form a plutonium resistant self-healing tantalum carbide film. (AEC)

METHOD OF OPERATING A NEUTRONIC REACTOR

DOEpatents

Turkevich, A.

1963-01-22

This patent relates to one step in a method of operating a neutronic reactor consisting of a slurry of fissionable material in heavy water. Deuterium gas is passed through the slurry to sweep gaseous fission products therefrom and the deuterium is then separated from the gaseous fission products. (AEC)

REACTOR

DOEpatents

Szilard, L.

1963-09-10

A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

METHOD FOR OBTAINING PLUTONIUM METAL AND ALLOYS OF PLUTONIUM FROM PLUTONIUM TRICHLORIDE

DOEpatents

Reavis, J.G.; Leary, J.A.; Maraman, W.J.

1962-11-13

A process is given for both reducing plutonium trichloride to plutonium metal using cerium as the reductant and simultaneously alloying such plutonium metal with an excess of cerium or cerium and cobalt sufficient to yield the desired nuclear reactor fuel composition. The process is conducted at a temperature from about 550 to 775 deg C, at atmospheric pressure, without the use of booster reactants, and a substantial decontamination is effected in the product alloy of any rare earths which may be associated with the source of the plutonium. (AEC)

SPRING DRIVEN ACTUATING MECHANISM FOR NUCLEAR REACTOR CONTROL

DOEpatents

Bevilacqua, F.; Uecker, D.F.; Groh, E.F.

1962-01-23

l962. rod in a nuclear reactor to shut it down. The control rod or an extension thereof is wound on a drum as it is withdrawn from the reactor. When an emergency occurs requiring the reactor to be shut down, the drum is released so as to be free to rotate, and the tendency of the control rod or its extension coiled on the drum to straighten itself is used for quickly returning the control rod to the reactor. (AEC)

NUCLEAR FLASH TYPE STEAM GENERATOR

DOEpatents

Johns, F.L.; Gronemeyer, E.C.; Dusbabek, M.R.

1962-09-01

A nuclear steam generating apparatus is designed so that steam may be generated from water heated directly by the nuclear heat source. The apparatus comprises a pair of pressure vessels mounted one within the other, the inner vessel containing a nuclear reactor heat source in the lower portion thereof to which water is pumped. A series of small ports are disposed in the upper portion of the inner vessel for jetting heated water under pressure outwardly into the atmosphere within the interior of the outer vessel, at which time part of the jetted water flashes into steam. The invention eliminates the necessity of any intermediate heat transfer medium and components ordinarily required for handling that medium. (AEC)

Annual Report to Congress of the Atomic Energy Commission for 1969

DOE Office of Scientific and Technical Information (OSTI.GOV)

Seaborg, Glenn T.

1970-01-31

The document represents the 1969 Annual Report of the Atomic Energy Commission (AEC) to Congress. The report opens with ''An Introduction to the Atomic Energy Programs during 1969'' followed by 17 Chapters, 8 appendices and an index. Chapters are as follows: (1) Source, Special, and Byproduct Nuclear Materials; (2) Nuclear Materials Safeguards; (3) The Nuclear Defense Effort; (4) Naval Propulsion Reactors; (5) Reactor Development and Technology; (6) Licensing and Regulating the Atom; (7) Operational and Public Safety; (8) Space Nuclear Propulsion; (9) Specialized Nuclear Power; (10) Isotopic Radiation Applications; (11) Peaceful Nuclear Explosives; (12) International Affairs and Cooperation; (13) Informationalmore » and Related Activities; (14) Nuclear Education and Training; (15) Biomedical and Physical Research; (16) Industrial Participation Aspects; and, (17) Administrative and Management Matters.« less

Annual Report to Congress of the Atomic Energy Commission for 1968

DOE Office of Scientific and Technical Information (OSTI.GOV)

Seaborg, Glenn T.

1969-01-31

The document represents the 1968 Annual Report of the Atomic Energy Commission (AEC) to Congress. The report opens with ''An Introduction to the Atomic Energy Programs during 1968'' followed by 17 Chapters, 8 appendices and an index. Chapters are as follows: (1) Source, Special, and Nuclear Byproduct Materials; (2) Nuclear Materials Safeguards; (3) The Nuclear Defense Effort; (4) Naval Propulsion Reactors; (5) Reactor Development and Technology; (6) Licensing and Regulating the Atom; (7) Operational and Public Safety; (8) Nuclear Rocket Propulsion; (9) Specialized Nuclear Power; (10) Isotopic Radiation Applications; (11) Peaceful Nuclear Explosives; (12) International Affairs and Cooperation; (13) Informationalmore » and Related Activities; (14) Nuclear Education and Training; (15) Biomedical and Physical Research; (16) Industrial Participation Aspects; and, (17) Administrative and Management Matters.« less

CONTROL ROD ROTATING MECHANISM

DOEpatents

Baumgarten, A.; Karalis, A.J.

1961-11-28

A threaded rotatable shaft is provided which rotates in response to linear movement of a nut, the shaft being surrounded by a pair of bellows members connected to either side of the nut to effectively seal the reactor from leakage and also to store up energy to shut down the reactor in the event of a power failure. (AEC)

NUCLEAR REACTOR

DOEpatents

Starr, C.

1963-01-01

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

METHOD OF FABRICATING A GRAPHITE MODERATED REACTOR

DOEpatents

Kratz, H.R.

1963-05-01

S>A nuclear reactor formed of spaced bodies of uranium and graphite blocks is improved by diffusing helium through the graphite blocks in order to replace the air in the pores of the graphite with helium. The helium-impregnated graphite conducts heat better, and absorbs neutrons less, than the original air- impregnated graphite. (AEC)

APPARATUS FOR LOADING AND UNLOADING A MACHINE

DOEpatents

Payne, J.H. Jr.

1962-07-17

An arrangement for loading and unloading a nuclear reactor is described. Depleted fuel elements are removed from the reactor through one of a small number of holes in a shielding plug that is rotatably mounted in an eccentric annular plug rotatably mounted in the top of the reactor. The fuel elements removed are stored in a plurality of openings in a rotatable magazine or storage means rotatably mounted over the plugs. (AEC)

DENSITY CONTROL IN A REACTOR

DOEpatents

Marshall, J. Jr.

1961-10-24

A reactor is described in which natural-uranium bodies are located in parallel channels which extend through the graphite mass in a regular lattice. The graphite mass has additional channels that are out of the lattice and contain no uranium. These additional channels decrease in number per unit volume of graphite from the center of the reactor to the exterior and have the effect of reducing the density of the graphite more at the center than at the exterior, thereby spreading neutron activity throughout the reactor. (AEC)

PREPARATION OF HIGH-DENSITY, COMPACTIBLE THORIUM OXIDE PARTICLES

DOEpatents

McCorkle, K.H.; Kleinsteuber, A.T.; Schilling, C.E.; Dean, O.C.

1962-05-22

A method is given for preparing millimeter-size, highdensity thorium oxide particles suitable for fabrication into nuclear reactor feel elements by means of vibratory compaction. A thorium oxide gel containing 3.7 to 7 weight per cent residual volatile nitrate and water is prepared by drying a thorium oxide sol. The gel is then slowly heated to a temperature of about 450DEC, and the resulting gel fragments are calcined. The starting sol is prepared by repeated dispersion of oxalate-source thorium oxide in a nitrate system or by dispersion of steam-denitrated thorium oxide in water. (AEC)

Proceedings of the Conference on the Environmental Chemistry of Hydrazine Fuels (3rd) Held in Panama City Beach, Florida on 15-17 September 1987

DTIC Science & Technology

1988-01-01

for hydrauine, MMH and UDMH are 4.78 x 10-6, 10.2 x 10Ś, and 3.19 x 10-6 aecŕ, respectively. Plots of the log(area) versus time were linear and...followed first-order kinetics except for hydrauine, for which a non- linear portion was observed in the first 6 to 8 hours. This portion of the decay...As a result, the prototype flow reactor can be represented to good approximation by a linear combination of point source solutions (Reference 19). The

Fusion Studies in Japan

NASA Astrophysics Data System (ADS)

Ogawa, Yuichi

2016-05-01

A new strategic energy plan decided by the Japanese Cabinet in 2014 strongly supports the steady promotion of nuclear fusion development activities, including the ITER project and the Broader Approach activities from the long-term viewpoint. Atomic Energy Commission (AEC) in Japan formulated the Third Phase Basic Program so as to promote an experimental fusion reactor project. In 2005 AEC has reviewed this Program, and discussed on selection and concentration among many projects of fusion reactor development. In addition to the promotion of ITER project, advanced tokamak research by JT-60SA, helical plasma experiment by LHD, FIREX project in laser fusion research and fusion engineering by IFMIF were highly prioritized. Although the basic concept is quite different between tokamak, helical and laser fusion researches, there exist a lot of common features such as plasma physics on 3-D magnetic geometry, high power heat load on plasma facing component and so on. Therefore, a synergetic scenario on fusion reactor development among various plasma confinement concepts would be important.

METHOD OF FORMING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Layer, E.H. Jr.; Peet, C.S.

1962-01-23

A method is given for preparing a fuel element for a nuclear reactor. The method includes the steps of sandblasting a body of uranium dioxide to roughen the surface thereof, depositing a thin layer of carbon thereon by thermal decomposition of methane, and cladding the uranium dioxide body with zirconium by gas pressure bonding. (AEC)

Inverted Control Rod Lock-In Device

DOEpatents

Brussalis, W. G.; Bost, G. E.

1962-12-01

A mechanism which prevents control rods from dropping out of the reactor core in the event the vessel in which the reactor is mounted should capsize is described. The mechanism includes a pivoted toothed armature which engages the threaded control rod lead screw and prevents removal of the rod whenever the armature is not attracted by the provided electromagnetic means. (AEC)

NUCLEAR REACTOR FUEL ELEMENT

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.

1963-06-11

A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

REACTOR HAVING NaK-UO$sub 2$ SLURRY HELICALLY POSITIONED IN A GRAPHITE MODERATOR

DOEpatents

Rodin, M.B.; Carter, J.C.

1962-05-15

A reactor utilizing 20% enriched uranium consists of a central graphite island in cylindrical form, with a spiral coil of tubing fitting against the central island. An external graphite moderator is placed around the central island and coil. A slurry of uranium dioxide dispersed in alkali metal passes through the coil to transfer heat externally to the reactor. There are also conventional controls for regulating the nuclear reaction. (AEC)

Adsorption of plasmid DNA on anion exchange chromatography media.

PubMed

Tarmann, Christina; Jungbauer, Alois

2008-08-01

Anion exchange chromatography (AEC) is a useful and effective tool for DNA purification, but due to average pore sizes between 40 and 100 nm most AEC resins lack truly useful binding capacities for plasmid DNA (pDNA). Equilibrium binding capacities and uptake kinetics of AEC media including conventional media (Source 30 Q, Q Sepharose HP), a polymer grafted medium (Fractogel EMD DEAE (M)), media with large pores (Celbeads DEAE, PL SAX 4000 A 30 microm) and a monolithic medium (CIM-DEAE) were investigated by batch uptake or shallow bed experiments at two salt concentrations. Theoretical and experimental binding capacities suggest that the shape of the pDNA molecule can be described by a rod with a length to diameter ratio of 20:1 and that the molecule binds in upright position. The arrangement of DNA like a brush at the surface can be considered as entropy driven, kind of self-assembly process which is inherent to highly and uniformly charged DNA molecules. The initial phase of adsorption is very fast and levels off, associated with a change in mass transfer mechanism. Feed concentrations higher than 0.1 mg/mL pDNA pronounce this effect. Monolithic media showed the fastest adsorption rate and highest binding capacity with 13 mg pDNA per mL.

APPARATUS FOR CONTROL OF A BOILING REACTOR RESPONSIVE TO STEAM DEMAND

DOEpatents

Treshow, M.

1963-07-23

A method of controlling a fuel-rod-in-tube-type boilingwater reactor having nozzles at the point of water entry into the tube is described. Water is pumped into the nozzles by an auxiliary pump operated by steam from an interstage position of the associated turbine, so that the pumping speed is responsive to turbine demand. (AEC)

Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grabaskas, D.; Brunett, A.; Passerini, S.

Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. Themore » mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.« less

NUCLEAR REACTOR

DOEpatents

Young, G.

1963-01-01

This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

METHOD OF DEHYDRATING URANIUM TETRAFLUORIDE

DOEpatents

Davis, J.O.; Fogel, C.C.; Palmer, W.E.

1962-12-18

Drying and dehydration of aqueous-precipitated uranium tetrafluoride are described. The UF/sub 4/ which normally contains 3 to 4% water, is dispersed into the reaction zone of an operating reactor wherein uranium hexafluoride is being reduced to UF/sub 4/ with hydrogen. The water-containing UF/sub 4/ is dried and blended with the UF/sub 4/ produced in the reactor without interfering with the reduction reaction. (AEC)

REACTOR

DOEpatents

Spitzer, L. Jr.

1962-01-01

The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

NASA Reactor Facility Hazards Summary. Volume 2

NASA Technical Reports Server (NTRS)

1959-01-01

Supplements to volume 1 are presented herein. Included in these papers are information unavailable when volume 1 was written, an evaluation of the proposed nuclear facility, and answers to questions raised by the AEC concerning volume 1.

BOILING SLURRY REACTOR AND METHOD FO CONTROL

DOEpatents

Petrick, M.; Marchaterre, J.F.

1963-05-01

The control of a boiling slurry nuclear reactor is described. The reactor consists of a vertical tube having an enlarged portion, a steam drum at the top of the vertical tube, and at least one downcomer connecting the steam drum and the bottom of the vertical tube, the reactor being filled with a slurry of fissionabie material in water of such concentration that the enlarged portion of the vertical tube contains a critical mass. The slurry boils in the vertical tube and circulates upwardly therein and downwardly in the downcomer. To control the reactor by controlling the circulation of the slurry, a gas is introduced into the downcomer. (AEC)

NUCLEAR POWER PLANT

DOEpatents

Carter, J.C.; Armstrong, R.H.; Janicke, M.J.

1963-05-14

A nuclear power plant for use in an airless environment or other environment in which cooling is difficult is described. The power plant includes a boiling mercury reactor, a mercury--vapor turbine in direct cycle therewith, and a radiator for condensing mercury vapor. (AEC)

METHOD FOR SENSING DEGREE OF FLUIDIZATION IN FLUIDIZED BED

DOEpatents

Levey, R.P. Jr.; Fowler, A.H.

1961-12-12

A method is given for detecting, indicating, and controlling the degree of fluidization in a fluid-bed reactor into which powdered material is fed. The method comprises admitting of gas into the reactor, inserting a springsupported rod into the powder bed of the reactor, exciting the rod to vibrate at its resonant frequency, deriving a signal responsive to the amplitude of vibi-ation of the rod and spring, the signal being directiy proportional to the rate of flow of the gas through the reactor, displaying the signal to provide an indication of the degree of fluidization within the reactor, and controlling the rate of gas flow into the reactor until said signal stabilizes at a constant value to provide substantially complete fluidization within the reactor. (AEC)

PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

DOEpatents

Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

1963-09-01

A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

DUCTILE URANIUM FUEL FOR NUCLEAR REACTORS AND METHOD OF MAKING

DOEpatents

Zegler, S.T.

1963-11-01

The fabrication process for a ductile nuclear fuel alloy consisting of uranium, fissium, and from 0.25 to 1.0 wt% of silicon or aluminum or from 0.25 to 2 wt% of titanium or yttrium is presented. (AEC)

Annual Report to Congress of the Atomic Energy Commission for 1965

DOE Office of Scientific and Technical Information (OSTI.GOV)

Seaborg, Glenn T.

1966-01-31

The document represents the 1965 Annual Report of the Atomic Energy Commission (AEC) to Congress. The report opens with a Foreword - a letter from President Lyndon B. Johnson. The main portion is divided into 3 major sections for 1965, plus 10 appendices and the index. Section names and chapters are as follows. Part One reports on Developmental and Promotional Activities with the following chapters: (1) The Atomic Energy Program - 1965; (2) The Industrial Base ; (3) Industrial Relations; (4) Operational Safety; (5) Source and Special Nuclear Materials Production; (6) The Nuclear Defense Effort; (7) Civilian Nuclear Power; (8)more » Nuclear Space Applications; (9) Auxiliary Electrical Power for Land and Sea; (10) Military Reactors; (11) Advanced Reactor Technology and Nuclear Safety Research; (12) The Plowshare Program; (13) Isotopes and Radiation Development; (14) Facilities and Projects for Basic Research; (15) International Cooperation; and, (16) Nuclear Education and Information. Part Two reports on Regulatory Activities with the following chapters: (1) Licensing and Regulating the Atom; (2) Reactors and other Nuclear Facilities; and, (3) Control of Radioactive Materials. Part Three reports on Adjudicatory Activities.« less

Key Assets for a Sustainable Low Carbon Energy Future

NASA Astrophysics Data System (ADS)

Carre, Frank

2011-10-01

Since the beginning of the 21st century, concerns of energy security and climate change gave rise to energy policies focused on energy conservation and diversified low-carbon energy sources. Provided lessons of Fukushima accident are evidently accounted for, nuclear energy will probably be confirmed in most of today's nuclear countries as a low carbon energy source needed to limit imports of oil and gas and to meet fast growing energy needs. Future challenges of nuclear energy are then in three directions: i) enhancing safety performance so as to preclude any long term impact of severe accident outside the site of the plant, even in case of hypothetical external events, ii) full use of Uranium and minimization long lived radioactive waste burden for sustainability, and iii) extension to non-electricity energy products for maximizing the share of low carbon energy source in transportation fuels, industrial process heat and district heating. Advanced LWRs (Gen-III) are today's best available technologies and can somewhat advance nuclear energy in these three directions. However, breakthroughs in sustainability call for fast neutron reactors and closed fuel cycles, and non-electric applications prompt a revival of interest in high temperature reactors for exceeding cogeneration performances achievable with LWRs. Both types of Gen-IV nuclear systems by nature call for technology breakthroughs to surpass LWRs capabilities. Current resumption in France of research on sodium cooled fast neutron reactors (SFRs) definitely aims at significant progress in safety and economic competitiveness compared to earlier reactors of this type in order to progress towards a new generation of commercially viable sodium cooled fast reactor. Along with advancing a new generation of sodium cooled fast reactor, research and development on alternative fast reactor types such as gas or lead-alloy cooled systems (GFR & LFR) is strategic to overcome technical difficulties and/or political opposition specific to sodium. In conclusion, research and technology breakthroughs in nuclear power are needed for shaping a sustainable low carbon future. International cooperation is key for sharing costs of research and development of the required novel technologies and cost of first experimental reactors needed to demonstrate enabling technologies. At the same time technology breakthroughs are developed, pre-normative research is required to support codification work and harmonized regulations that will ultimately apply to safety and security features of resulting innovative reactor types and fuel cycles.

REFLECTOR CONTROL OF A BOILING-WATER REACTOR

DOEpatents

Treshow, M.

1962-05-22

A line connecting the reactor with a spent steam condenser contains a valve set to open when the pressure in the reactor exceeds a predetermined value and an orifice on the upstream side of the valve. Another line connects the reflector with this line between the orifice and the valve. An excess steam pressure causes the valve to open and the flow of steam through the line draws water out of the reflector. Provision is also made for adding water to the reflector when the steam pressure drops. (AEC)

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Stacey, W. M.

2009-09-01

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

FAST NEUTRON SPECTROMETER USING SPACED SEMICONDUCTORS FOR MEASURING TOTAL ENERGY OF NEUTRONS CAPTURED

DOEpatents

Love, T.A.; Murray, R.B.

1964-04-14

A fast neutron spectrometer was designed, which utilizes a pair of opposed detectors having a layer of /sup 6/LiF between to produce alpha and T pair for each neutron captured to provide signals, which, when combined, constitute a measure of neutron energy. (AEC)

METHOD AND MEANS FOR SUPPORTING REACTOR FUEL CONTAINERS IN AN ASSEMBLY

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.; Coombs, C.A.

1962-12-11

This patent relates to means for supporting fuelcontaining tubes in an assembly which include grid means at either end of the fuel element assembly antl improved grid means intermediate of the ends to provide support against lateral displacement. (AEC)

METHOD AND APPARATUS FOR EXAMINING FUEL ELEMENTS FOR LEAKAGE

DOEpatents

Smith, R.R.; Echo, M.W.; Doe, C.B.

1963-12-31

A process and a device for the continuous monitoring of fuel elements while in use in a liquid-metal-cooled, argonblanketed nuclear reactor are presented. A fraction of the argon gas is withdrawn, contacted with a negative electrical charge for attraction of any alkali metal formed from argon by neutron reaction, and recycled into the reactor. The electrical charge is introduced into water, and the water is examined for radioactive alkali metals. (AEC)

NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR

DOEpatents

Rasor, N.S.; Hirsch, R.L.

1963-12-01

The patent relates to the direct conversion of fission heat to electricity by use of thermionic plasma diodes having fissionable material cathodes, said diodes arranged to form a critical mass in a nuclear reactor. The patent describes a fuel element comprising a plurality of diodes each having a fissionable material cathode, an anode around said cathode, and an ionizable gas therebetween. Provision is made for flowing the gas and current serially through the diodes. (AEC)

NEUTRON FLUX INTENSITY DETECTION

DOEpatents

Russell, J.T.

1964-04-21

A method of measuring the instantaneous intensity of neutron flux in the core of a nuclear reactor is described. A target gas capable of being transmuted by neutron bombardment to a product having a resonance absorption line nt a particular microwave frequency is passed through the core of the reactor. Frequency-modulated microwave energy is passed through the target gas and the attenuation of the energy due to the formation of the transmuted product is measured. (AEC)

Synthetic Bone Substitute Engineered with Amniotic Epithelial Cells Enhances Bone Regeneration after Maxillary Sinus Augmentation

PubMed Central

Barboni, Barbara; Mangano, Carlo; Valbonetti, Luca; Marruchella, Giuseppe; Berardinelli, Paolo; Martelli, Alessandra; Muttini, Aurelio; Mauro, Annunziata; Bedini, Rossella; Turriani, Maura; Pecci, Raffaella; Nardinocchi, Delia; Zizzari, Vincenzo Luca; Tetè, Stefano; Piattelli, Adriano; Mattioli, Mauro

2013-01-01

Background Evidence has been provided that a cell-based therapy combined with the use of bioactive materials may significantly improve bone regeneration prior to dental implant, although the identification of an ideal source of progenitor/stem cells remains to be determined. Aim In the present research, the bone regenerative property of an emerging source of progenitor cells, the amniotic epithelial cells (AEC), loaded on a calcium-phosphate synthetic bone substitute, made by direct rapid prototyping (rPT) technique, was evaluated in an animal study. Material And Methods Two blocks of synthetic bone substitute (∼0.14 cm3), alone or engineered with 1×106 ovine AEC (oAEC), were grafted bilaterally into maxillary sinuses of six adult sheep, an animal model chosen for its high translational value in dentistry. The sheep were then randomly divided into two groups and sacrificed at 45 and 90 days post implantation (p.i.). Tissue regeneration was evaluated in the sinus explants by micro-computer tomography (micro-CT), morphological, morphometric and biochemical analyses. Results And Conclusions The obtained data suggest that scaffold integration and bone deposition are positively influenced by allotransplantated oAEC. Sinus explants derived from sheep grafted with oAEC engineered scaffolds displayed a reduced fibrotic reaction, a limited inflammatory response and an accelerated process of angiogenesis. In addition, the presence of oAEC significantly stimulated osteogenesis either by enhancing bone deposition or making more extent the foci of bone nucleation. Besides the modulatory role played by oAEC in the crucial events successfully guiding tissue regeneration (angiogenesis, vascular endothelial growth factor expression and inflammation), data provided herein show that oAEC were also able to directly participate in the process of bone deposition, as suggested by the presence of oAEC entrapped within the newly deposited osteoid matrix and by their ability to switch-on the expression of a specific bone-related protein (osteocalcin, OCN) when transplanted into host tissues. PMID:23696804

Alveolar epithelial cell processing of nanoparticles activates autophagy and lysosomal exocytosis.

PubMed

Sipos, Arnold; Kim, Kwang-Jin; Chow, Robert H; Flodby, Per; Borok, Zea; Crandall, Edward D

2018-05-03

Utilizing confocal microscopy, we quantitatively assessed uptake, processing and egress of near infrared (NIR)-labeled carboxylated polystyrene nanoparticles (PNP) in live alveolar epithelial cells (AEC) during interactions with primary rat AEC monolayers (RAECM). PNP fluorescence intensity (content) and colocalization with intracellular vesicles in a cell were determined over the entire cell volume via z-stacking. Isotropic cuvette-based microfluorimetry was used to determine PNP concentration ([PNP]) from anisotropic measurements of PNP content assessed by confocal microscopy. Results showed that PNP uptake kinetics and steady state intracellular content decreased as diameter increased from 20 to 200 nm. For 20 nm PNP, uptake rate and steady state intracellular content increased with increased apical [PNP], but were unaffected by inhibition of endocytic pathways. Intracellular PNP increasingly co-localized with autophagosomes and/or lysosomes over time. PNP egress exhibited fast [Ca2+]-dependent release and a slower diffusion-like process. Inhibition of microtubule polymerization curtailed rapid PNP egress, resulting in elevated vesicular and intracellular PNP content. Interference with autophagosome formation led to slower PNP uptake and markedly decreased steady state intracellular content. At steady state, cytosolic [PNP] was higher than apical [PNP] and vesicular [PNP] (~80% of intracellular PNP content) exceeded both cytosolic [PNP] and intracellular [PNP]. These data are consistent with the hypotheses that (1) autophagic processing of nanoparticles is essential for maintenance of AEC integrity, (2) altered autophagy and/or lysosomal exocytosis may lead to AEC injury and (3) intracellular [PNP] in AEC is regulable, suggesting strategies for enhancement of nanoparticle-driven AEC gene/drug delivery and/or amelioration of AEC nanoparticle-related cellular toxicity.

NON-CORROSIVE PLUTONIUM FUEL SYSTEMS

DOEpatents

Coffinberry, A.S.; Waber, J.T.

1962-10-23

An improved plutonium reactor liquid fuel is described for utilization in a nuclear reactor having a tantalum fuel containment vessel. The fuel consists of plutonium and a diluent such as iron, cobalt, nickel, cerium, cerium-- iron, cerium--cobalt, cerium--nickel, and cerium--copper, and an additive of carbon and silicon. The carbon and silicon react with the tantalum container surface to form a coating that is self-healing and prevents the corrosive action of liquid plutonium on the said tantalum container. (AEC)

FOIL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Noland, R.A.; Walker, D.E.; Spinrad, B.I.

1963-07-16

A method of making a foil-type fuel element is described. A foil of fuel metal is perforated in; regular design and sheets of cladding metal are placed on both sides. The cladding metal sheets are then spot-welded to each other through the perforations, and the edges sealed. (AEC)

FAST CHOPPER BUILDING, TRA665, INTERIOR. LOWER (DETECTOR) LEVEL. NOTE BRICKEDIN ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FAST CHOPPER BUILDING, TRA-665, INTERIOR. LOWER (DETECTOR) LEVEL. NOTE BRICKED-IN WINDOW ON MTR SIDE. USED FOR STORAGE OF LEAD BRICKS AFTER EXPERIMENTAL NEUTRON INSTRUMENTS WERE REMOVED. SIGN SAYS "IN-PROCESS LEAD SOURCE STORAGE." INL NEGATIVE NO. HD-42-2. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

REACTOR FUEL SCAVENGING MEANS

DOEpatents

Coffinberry, A.S.

1962-04-10

A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

2017-01-01

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

Safety and core design of large liquid-metal cooled fast breeder reactors

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

REACTOR CONTROL

DOEpatents

Fortescue, P.; Nicoll, D.

1962-04-24

A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

REACTOR CONTROL DEVICE

DOEpatents

Graham, R.H.

1962-09-01

A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

FUEL ELEMENT CONSTRUCTION

DOEpatents

Simnad, M.T.

1961-08-15

A method of preventing diffusible and volatile fission products from diffusing through a fuel element container and contaminating reactor coolant is described. More specifically, relatively volatile and diffusible fission products either are adsorbed by or react with magnesium fluoride or difluoride to form stable, less volatile, less diffusible forms. The magnesium fluoride or difluoride is disposed anywhere inwardly from the outer surface of the fuel element container in order to be contacted by the fission products before they reach and contaminate the reactor coolant. (AEC)

FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Davidson, J.K.

1963-11-19

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

NEUTRON SOURCE USING MAGNETIC COMPRESSION OF PLASMA

DOEpatents

Quinn, W.E.; Elmore, W.C.; Little, E.M.; Boyer, K.; Tuck, J.L.

1961-10-31

A fusion reactor is described that utilizes compression and heating of an ionized thermonuclear fuel by an externally applied magnetic field, thus avoiding reliance on the pinch effect and its associated instability problems. The device consists of a gas-confining ceramic container surrounded by a single circumferential coil having a shape such as to produce a magnetic mirror geometry. A sinusoidally-oscillating, exponentially-damped current is passed circumferentially around the container, through the coil, inducing a circumferential current in the gas. Maximum compression and plasma temperature are obtained at the peak of the current oscillations, coinciding with maximum magnetic field intensity. Enhanced temperatures are obtained in the second and succeeding half cycles because the thermal energy accumulates from one half cycle to the next. (AEC)

MAGNETIC END CLOSURES FOR PLASMA CONFINING AND HEATING DEVICES

DOEpatents

Post, R.F.

1963-08-20

More effective magnetic closure field regions for various open-ended containment magnetic fields used in fusion reactor devices are provided by several spaced, coaxially-aligned solenoids utilized to produce a series of nodal field regions of uniform or, preferably, of incrementally increasing intensity separated by lower intensity regions outwardly from the ends of said containment zone. Plasma sources may also be provided to inject plasma into said lower intensity areas to increase plasma density therein. Plasma may then be transported, by plasma diffusion mechanisms provided by the nodal fields, into the containment field. With correlated plasma densities and nodal field spacings approximating the mean free partl cle collision path length in the zones between the nodal fields, optimum closure effectiveness is obtained. (AEC)

Technicians Manufacture a Nozzle for the Kiwi B-1-B Engine

NASA Image and Video Library

1964-05-21

Technicians manufacture a nozzle for the Kiwi B-1-B nuclear rocket engine in the Fabrication Shop’s vacuum oven at the National Aeronautics and Space Administration (NASA) Lewis Research Center. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test basic nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The final phase of the program, called Reactor-In-Flight-Test, would be an actual launch test. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The turbopump, which pumped the fuels from the storage tanks to the engine, was the primary tool for restarting the engine. The NERVA had to be able to restart in space on its own using a safe preprogrammed startup system. Lewis researchers endeavored to design and test this system. This non-nuclear Kiwi engine, seen here, was being prepared for tests at Lewis’ High Energy Rocket Engine Research Facility (B-1) located at Plum Brook Station. The tests were designed to start an unfueled Kiwi B-1-B reactor and its Aerojet Mark IX turbopump without any external power.

History of Nuclear India

NASA Astrophysics Data System (ADS)

Chaturvedi, Ram

2000-04-01

India emerged as a free and democratic country in 1947, and entered into the nuclear age in 1948 by establishing the Atomic Energy Commission (AEC), with Homi Bhabha as the chairman. Later on the Department of Atomic Energy (DAE) was created under the Office of the Prime Minister Jawahar Lal Nehru. Initially the AEC and DAE received international cooperation, and by 1963 India had two research reactors and four nuclear power reactors. In spite of the humiliating defeat in the border war by China in 1962 and China's nuclear testing in 1964, India continued to adhere to the peaceful uses of nuclear energy. On May 18, 1974 India performed a 15 kt Peaceful Nuclear Explosion (PNE). The western powers considered it nuclear weapons proliferation and cut off all financial and technical help, even for the production of nuclear power. However, India used existing infrastructure to build nuclear power reactors and exploded both fission and fusion devices on May 11 and 13, 1998. The international community viewed the later activity as a serious road block for the Non-Proliferation Treaty and the Comprehensive Test Ban Treaty; both deemed essential to stop the spread of nuclear weapons. India considers these treaties favoring nuclear states and is prepared to sign if genuine nuclear disarmament is included as an integral part of these treaties.

Role of ADS in the back-end of the fuel cycle strategies and associated design activities: The case of Japan

NASA Astrophysics Data System (ADS)

Oigawa, Hiroyuki; Tsujimoto, Kazufumi; Nishihara, Kenji; Sugawara, Takanori; Kurata, Yuji; Takei, Hayanori; Saito, Shigeru; Sasa, Toshinobu; Obayashi, Hironari

2011-08-01

Reduction of burden caused by radioactive waste management is one of the most critical issues for the sustainable utilization of nuclear power. The Partitioning and Transmutation (P&T) technology provides the possibility to reduce the amount of the radiotoxic inventory of the high-level radioactive waste (HLW) dramatically and to extend the repository capacity. The accelerator-driven system (ADS) is regarded as a powerful tool to effectively transmute minor actinides (MAs) in the "double-strata" fuel cycle strategy. The ADS has a potential to flexibly manage MA in the transient phase from light water reactors (LWRs) to fast breeder reactors (FBRs), and can co-exist with FBR symbiotically and complementarily to enhance the reliability and the safety of the commercial FBR cycle. The concept of ADS in JAEA is a lead-bismuth eutectic (LBE) cooled, tank-type subcritical reactor with the power of 800 MWth driven by a 30 MW superconducting LINAC. By such an ADS, 250 kg of MA can be transmuted annually, which corresponds to the amount of MA produced in 10 units of LWR with 1 GWe. The design study was performed mainly for the subcritical reactor and the spallation target with a beam window. In Japan, Atomic Energy Commission (AEC) has implemented the check and review (C&R) on P&T technology from 2008 to 2009. In the C&R, the benefit of P&T technology, the current status of the R&D, and the way forward to promote it were discussed.

THREADED ADAPTOR FOR LUGGED PIPE ENDS

DOEpatents

Robb, J.E.

1962-06-01

An adaptor is designed for enabling a threaded part to be connected to a member at a region having lugs normally receiving bayonet slots of another part for attachment of the latter. It has been found desirable to replace a closure cap connected in a bayonet joint to the end of a coolant tube containing nuclear- reactor fuel elements, with a threaded valve. An adaptor is used which has J- slots receiving lugs on the end of the reactor tube, a thread for connection with the valve, and gear-tooth section enabling a gear-type of tool to rotate the adaptor to seal the valve to the end of the reactor tube. (AEC)

CONTROL ROD

DOEpatents

Walker, D.E.; Matras, S.

1963-04-30

This patent shows a method of making a fuel or control rod for a nuclear reactor. Fuel or control material is placed within a tube and plugs of porous metal wool are inserted at both ends. The metal wool is then compacted and the tube compressed around it as by swaging, thereby making the plugs liquid- impervious but gas-pervious. (AEC)

METHOD FOR PREPARATION OF SINTERABLE BERYLLIUM OXIDE

DOEpatents

Sturm, B.J.

1963-08-13

High-purity beryllium oxide for nuclear reactor applications can be prepared by precipitation of beryllium oxalate monohydrate from aqueous solution at a temperature above 50 deg C and subsequent calcination of the precipitate. Improved purification with respect to metallic impurities is obtained, and the product beryllium oxide sinters reproducibly to a high density. (AEC)

ALLOY FOR FUEL OF NEUTRONIC REACTORS

DOEpatents

Bloomster, C.H.; Katayama, Y.B.

1963-04-23

This patent deals with an aluminum alloy suitable as nuclear fuel and consisting mainly of from 1 to 10 wt% of plutonium, from 2 to 3.5 wt% of nickel, the balance being aluminum. The alloy may also contain from 0.9 to 1.1 wt% of silicon and up to 0.7% of iron. (AEC)

CATALYTIC RECOMBINATION OF RADIOLYTIC GASES IN THORIUM OXIDE SLURRIES

DOEpatents

Morse, L.E.

1962-08-01

A method for the coinbination of hydrogen and oxygen in aqueous thorium oxide-uranium oxide slurries is described. A small amount of molybdenum oxide catalyst is provided in the slurry. This catalyst is applicable to the recombination of hydrogen and/or deuterium and oxygen produced by irradiation of the slurries in nuclear reactors. (AEC)

Macroscopic barotrauma caused by stiff and soft-tipped airway exchange catheters: an in vitro case series.

PubMed

Axe, Robert; Middleditch, Alex; Kelly, Fiona E; Batchelor, Tim J; Cook, Tim M

2015-02-01

Many airway management guidelines include the use of airway exchange catheters (AECs). There are reports, however, of harm from their use, from both malpositioning and in particular from the administration of oxygen via an AEC leading to barotrauma. We used an in vitro pig lung model to investigate the safety of administering oxygen at 4 different flow rates from a high-pressure source via 2 different AECs: a standard catheter and a soft-tipped catheter. Experiments were performed with the catheters positioned either above the carina or below it at the first point of resistance to advancement (hold-up). The experiments were then repeated to produce a series of 32 cases. With an AEC positioned above the carina, we did not observe macroscopic lung damage after the administration of oxygen. The administration of oxygen through an AEC positioned below the carina resulted in macroscopic barotrauma regardless of the rate of oxygen delivery. Increasing speed of oxygen flow led to faster and more extensive damage. Use of an "injector" at 2.5 or 4 bar led to instantaneous macroscopic lung damage and advancement of the AEC through the lung tissue. Our observations were the same when both types of AECs were used. Our results are consistent with reports of harm during the use of AECs and demonstrate the risk of administering oxygen through these devices when they are positioned below the carina. An indicator, ideally made on an AEC at the time of manufacture and designed to lie at the same level as the teeth, may be useful in preventing the insertion of that AEC beyond the level of the carina and improve the safety of using such devices.

Merchant Marine Ship Reactor

DOEpatents

Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

1961-05-01

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

COATED CARBON ELEMENT FOR USE IN NUCLEAR REACTORS AND THE PROCESS OF MAKING THE ELEMENT

DOEpatents

Pyle, R.J.; Allen, G.L.

1963-01-15

S>This patent relates to a carbide-nitride-carbide coating for carbon bodies that are to be subjected to a high temperature nuclear reactor atmosphere, and a method of applying the same. This coating is a highly efficient diffusion barrier and protects the C body from corrosion and erosion by the reactor atmosphere. Preferably, the innermost coating is Zr carbide, the middle coatlng is Zr nitride, and the outermost coating is a mixture of Zr and Nb carbide. The nitride coating acts as a diffusion barrier, while the innermost carbide bonds the nitride to the C body and prevents deleterious reaction between the nitride and C body. The outermost carbide coating protects the nitride coating from the reactor atmosphere. (AEC)

An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, M.; Farmer, M.; Grabaskas, D.

The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination ofmore » gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.« less

The effect of epoch length on estimated EEG functional connectivity and brain network organisation

NASA Astrophysics Data System (ADS)

Fraschini, Matteo; Demuru, Matteo; Crobe, Alessandra; Marrosu, Francesco; Stam, Cornelis J.; Hillebrand, Arjan

2016-06-01

Objective. Graph theory and network science tools have revealed fundamental mechanisms of functional brain organization in resting-state M/EEG analysis. Nevertheless, it is still not clearly understood how several methodological aspects may bias the topology of the reconstructed functional networks. In this context, the literature shows inconsistency in the chosen length of the selected epochs, impeding a meaningful comparison between results from different studies. Approach. The aim of this study was to provide a network approach insensitive to the effects that epoch length has on functional connectivity and network reconstruction. Two different measures, the phase lag index (PLI) and the amplitude envelope correlation (AEC) were applied to EEG resting-state recordings for a group of 18 healthy volunteers using non-overlapping epochs with variable length (1, 2, 4, 6, 8, 10, 12, 14 and 16 s). Weighted clustering coefficient (CCw), weighted characteristic path length (L w) and minimum spanning tree (MST) parameters were computed to evaluate the network topology. The analysis was performed on both scalp and source-space data. Main results. Results from scalp analysis show a decrease in both mean PLI and AEC values with an increase in epoch length, with a tendency to stabilize at a length of 12 s for PLI and 6 s for AEC. Moreover, CCw and L w show very similar behaviour, with metrics based on AEC more reliable in terms of stability. In general, MST parameters stabilize at short epoch lengths, particularly for MSTs based on PLI (1-6 s versus 4-8 s for AEC). At the source-level the results were even more reliable, with stability already at 1 s duration for PLI-based MSTs. Significance. The present work suggests that both PLI and AEC depend on epoch length and that this has an impact on the reconstructed network topology, particularly at the scalp-level. Source-level MST topology is less sensitive to differences in epoch length, therefore enabling the comparison of brain network topology between different studies.

FUEL ASSAY REACTOR

DOEpatents

Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

1962-12-25

A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

METHOD OF OPERATING A HEAVY WATER MODERATED REACTOR

DOEpatents

Vernon, H.C.

1962-08-14

A method of removing fission products from the heavy water used in a slurry type nuclear reactor is described. According to the process the slurry is steam distilled with carbon tetrachloride so that at least a part of the heavy water and carbon tetrachloride are vaporized; the heavy water and carbon tetrachloride are separated; the carbon tetrachloride is returned to the steam distillation column at different points in the column to aid in depositing the slurry particles at the bottom of the column; and the heavy water portion of the condensate is purified. (AEC)

EXPERIMENTAL LIQUID METAL FUEL REACTOR

DOEpatents

Happell, J.J.; Thomas, G.R.; Denise, R.P.; Bunts, J.L. Jr.

1962-01-23

A liquid metal fuel nuclear fission reactor is designed in which the fissionable material is dissolved or suspended in a liquid metal moderator and coolant. The liquid suspension flows into a chamber in which a critical amount of fissionable material is obtained. The fluid leaves the chamber and the heat of fission is extracted for power or other utilization. The improvement is in the support arrangement for a segrnented graphite core to permit dif ferential thermal expansion, effective sealing between main and blanket liquid metal flows, and avoidance of excessive stress development in the graphite segments. (AEC)

Test of a prototype neutron spectrometer based on diamond detectors in a fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osipenko, M.; Ripani, M.; Ricco, G.

2015-07-01

A prototype of neutron spectrometer based on diamond detectors has been developed. This prototype consists of a {sup 6}Li neutron converter sandwiched between two CVD diamond crystals. The radiation hardness of the diamond crystals makes it suitable for applications in low power research reactors, while a low sensitivity to gamma rays and low leakage current of the detector permit to reach good energy resolution. A fast coincidence between two crystals is used to reject background. The detector was read out using two different electronic chains connected to it by a few meters of cable. The first chain was based onmore » conventional charge-sensitive amplifiers, the other used a custom fast charge amplifier developed for this purpose. The prototype has been tested at various neutron sources and showed its practicability. In particular, the detector was calibrated in a TRIGA thermal reactor (LENA laboratory, University of Pavia) with neutron fluxes of 10{sup 8} n/cm{sup 2}s and at the 3 MeV D-D monochromatic neutron source named FNG (ENEA, Rome) with neutron fluxes of 10{sup 6} n/cm{sup 2}s. The neutron spectrum measurement was performed at the TAPIRO fast research reactor (ENEA, Casaccia) with fluxes of 10{sup 9} n/cm{sup 2}s. The obtained spectra were compared to Monte Carlo simulations, modeling detector response with MCNP and Geant4. (authors)« less

RADIATION FACILITY FOR NUCLEAR REACTORS

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.

1961-12-12

A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

METHOD OF MANUFACTURE OF METAL ENCASED CORE MATERIAL

DOEpatents

Peters, E.J.

1963-06-01

A method of making reactor fuel elements in which the fissionable material is encased and sealed in steel or aluminum cladding having an enclosed channel from the fissionable material to the surface of the cladding at one end is described. Heat and pressure sufficient to bond the assembled fuel element are applied in a nonoxidizing atmosphere. (AEC)

METHOD OF OPPOSING IRRADIATION-INDUCED VISCOSITY INCREASE IN EMPLOYMENT OF ORGANIC FLUIDS

DOEpatents

Balt, R.O.

1961-10-24

A method is described for conducting mechanical operations necessitating the use of a lubricant in a medium operaject to reactor irradiation of 0.5 x 10/ sup 12/ to 1 x 10/sup 12/ neut rons/ cm/sup 2//sec. A thiopolyether lubricant such as 16, 19-dioxa-13, 22-dithiatetratriacontane is used. (AEC)

Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term – Trial Calculation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grabaskas, David; Bucknor, Matthew; Jerden, James

2016-10-01

The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is notmore » without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.« less

BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES

DOEpatents

Treshow, M.

1963-04-30

This patent covers the use of injection nozzles for pumping water into the lower ends of reactor fuel tubes in which water is converted directly to steam. Pumping water through fuel tubes of this type of boiling water reactor increases its power. The injection nozzles decrease the size of pump needed, because the pump handles only the water going through the nozzles, additional water being sucked into the tubes by the nozzles independently of the pump from the exterior body of water in which the fuel tubes are immersed. The resulting movement of exterior water along the tubes holds down steam formation, and thus maintains the moderator effectiveness, of the exterior body of water. (AEC)

COMPOSITE NEUTRONIC REACTOR

DOEpatents

Menke, J.R.

1963-06-11

This patent relates to a reactor having a core which comprises an inner active region and an outer active region, each region separately having a k effective less than one and a k infinity greater than one. The inner and outer regions in combination have a k effective at least equal to one and each region contributes substantially to the k effective of the reactor core. The inner region has a low moderator to fuel ratio such that the majority of fissions occurring therein are induced by neutrons having energies greater than thermal. The outer region has a high moderator to fuel ratio such that the majority of fissions occurring therein are induced by thermal neutrons. (AEC)

Human amniotic membrane-derived epithelial stem cells display anticancer activity in BALB/c female nude mice bearing disseminated breast cancer xenografts.

PubMed

Kang, Nam-Hee; Yi, Bo-Rim; Lim, So Yoon; Hwang, Kyung-A; Baek, Young Seok; Kang, Kyung-Sun; Choi, Kyung-Chul

2012-06-01

Breast cancer is one of the most common malignant tumors and the leading cause of mortality among women. In this study, we propose a human stem cell transplantation strategy, an important method for treating various cancers, as a potential breast cancer therapy. To this end, we used human amniotic membrane-derived epithelial stem cells (hAECs) as a cell source for performing human stem cell transplantation. hAECs have multipotent differentiation abilities and possess high proliferative potential. We transplanted hAECs into female BALB/c nude mice bearing tumors originating from MDA-MB-231 breast cancer cells. Co-culturred hAECs and MDA-MB-231 cells at a ratio of 1:4 or 1:8 (tumor cells to stem cells) inhibited breast cancer cell growth by 67.29 and 67.33%, respectively. In the xenograft mouse model, tumor volumes were significantly decreased by 5-flurouracil (5-FU) treatment and two different ratios of hAECs (1:4 and 1:8) by 84.33, 73.88 and 56.89%, respectively. Treatment of nude mice with hAECs (1:4) produced remarkable antitumor effects without any side-effects (e.g., weight loss, death and bruising) compared to the mice that received only 5-FU treatment. Tumor progression was significantly reduced by hAEC treatment compared to the xenograft model. On the other hand, breast tissues (e.g., the epidermis, dermis and reticular layer) appeared to be well-maintained following treatment with hAECs. Taken together, these results provide strong evidence that hAECs can be used as a safe and effective cancer-targeting cytotherapy for treating breast cancer.

Analysis of the propagation of neutrons and gamma-rays from the fast neutron source reactor YAYOI

NASA Astrophysics Data System (ADS)

Yoshida, Shigeo; Murata, Isao; Nakagawa, Tsutomu; Saito, Isao

2011-10-01

The skyshine effect is crucial for designing appropriate shielding. To investigate the skyshine effect, the propagation of neutrons was measured and analyzed at the fast neutron source reactor YAYOI. Pulse height spectra and dose distributions of neutron and secondary gamma-ray were measured outside YAYOI, and analyzed with MCNP-5 and JENDL-3.3. Comparison with the experimental results showed good agreement. Also, a semi-empirical formula was successfully derived to describe the dose distribution. The formulae can be used to predict the skyshine effect at YAYOI, and will be useful for estimating the skyshine effect and designing the shield structure for fusion facilities.

Nuclear fuel requirements for the American economy - A model

NASA Astrophysics Data System (ADS)

Curtis, Thomas Dexter

A model is provided to determine the amounts of various fuel streams required to supply energy from planned and projected nuclear plant operations, including new builds. Flexible, user-defined scenarios can be constructed with respect to energy requirements, choices of reactors and choices of fuels. The model includes interactive effects and extends through 2099. Outputs include energy provided by reactors, the number of reactors, and masses of natural Uranium and other fuels used. Energy demand, including electricity and hydrogen, is obtained from US DOE historical data and projections, along with other studies of potential hydrogen demand. An option to include other energy demand to nuclear power is included. Reactor types modeled include (thermal reactors) PWRs, BWRs and MHRs and (fast reactors) GFRs and SFRs. The MHRs (VHTRs), GFRs and SFRs are similar to those described in the 2002 DOE "Roadmap for Generation IV Nuclear Energy Systems." Fuel source choices include natural Uranium, self-recycled spent fuel, Plutonium from breeder reactors and existing stockpiles of surplus HEU, military Plutonium, LWR spent fuel and depleted Uranium. Other reactors and fuel sources can be added to the model. Fidelity checks of the model's results indicate good agreement with historical Uranium use and number of reactors, and with DOE projections. The model supports conclusions that substantial use of natural Uranium will likely continue to the end of the 21st century, though legacy spent fuel and depleted uranium could easily supply all nuclear energy demand by shifting to predominant use of fast reactors.

How to Produce a Reactor Neutron Spectrum Using a Proton Accelerator

DOE PAGES

Burns, Kimberly A.; Wootan, David W.; Gates, Robert O.; ...

2015-06-18

A method for reproducing the neutron energy spectrum present in the core of an operating nuclear reactor using an engineered target in an accelerator proton beam is proposed. The protons interact with a target to create neutrons through various (p,n) type reactions. Spectral tailoring of the emitted neutrons can be used to modify the energy of the generated neutron spectrum to represent various reactor spectra. Through the use of moderators and reflectors, the neutron spectrum can be modified to reproduce many different spectra of interest including spectra in small thermal test reactors, large pressurized water reactors, and fast reactors. Themore » particular application of this methodology is the design of an experimental approach for using an accelerator to measure the betas produced during fission to be used to reduce uncertainties in the interpretation of reactor antineutrino measurements. This approach involves using a proton accelerator to produce a neutron field representative of a power reactor, and using this neutron field to irradiate fission foils of the primary isotopes contributing to fission in the reactor, creating unstable, neutron rich fission products that subsequently beta decay and emit electron antineutrinos. A major advantage of an accelerator neutron source over a neutron beam from a thermal reactor is that the fast neutrons can be slowed down or tailored to approximate various power reactor spectra. An accelerator based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra affect parameters such as the resulting fission product beta spectrum.« less

CALANDRIA TYPE SODIUM GRAPHITE REACTOR

DOEpatents

Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

1964-02-11

A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

NASA Astrophysics Data System (ADS)

Koltashev, D. A.; Stakhanova, A. A.

2017-01-01

This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

CONTROL ROD DRIVE MECHANISM FOR A NUCLEAR REACTOR

DOEpatents

Hawke, B.C.; Liederbach, F.J.; Lones, W.

1963-05-14

A lead-screw-type control rod drive featuring an electric motor and a fluid motor arranged to provide a selectably alternative driving means is described. The electric motor serves to drive the control rod slowly during normal operation, while the fluid motor, assisted by an automatic declutching of the electric motor, affords high-speed rod insertion during a scram. (AEC)

PRODUCTION OF TRITIUM

DOEpatents

Jenks, G.H.; Shapiro, E.M.; Elliott, N.; Cannon, C.V.

1963-02-26

This invention relates to a process for the production of tritium by subjecting comminuted solid lithium fluoride containing the lithium isotope of atomic mass number 6 to neutron radiation in a self-sustaining neutronic reactor. The lithium fiuoride is heated to above 450 deg C. in an evacuated vacuum-tight container during radiation. Gaseous radiation products are withdrawn and passed through a palladium barrier to recover tritium. (AEC)

SEPARATION OF PROTACTINIUM FROM MOLTEN SALT REACTOR FUEL COMPOSITIONS

DOEpatents

Shaffer, J.H.; Strain, J.E.; Cuneo, D.R.; Kelly, M.J.

1963-11-12

A method for selectively precipitating protactinium from a neutron- irradiated fused fluoride salt composition comprising at least one metal fluoride selected from the group consisting of an alkali metal fluoride and an alkaline earth metal fluoride containing dissolved thorium-232 values is presented. An inorganic metal oxide corresponding to any of the metal fluorides of the composition is also added. (AEC)

PREPARATION OF SPHERICAL URANIUM DIOXIDE PARTICLES

DOEpatents

Levey, R.P. Jr.; Smith, A.E.

1963-04-30

This patent relates to the preparation of high-density, spherical UO/sub 2/ particles 80 to 150 microns in diameter. Sinterable UO/sub 2/ powder is wetted with 3 to 5 weight per cent water and tumbled for at least 48 hours. The resulting spherical particles are then sintered. The sintered particles are useful in dispersion-type fuel elements for nuclear reactors. (AEC)

Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

2017-09-01

The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

SODIUM DEUTERIUM REACTOR

DOEpatents

Oppenheimer, E.D.; Weisberg, R.A.

1963-02-26

This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

HIGH TEMPERATURE REACTOR

DOEpatents

Kulsrud, R.M.; Spitzer, L. Jr.

1961-12-12

An apparatus of the stellarator type for heating a plasma to high temperatures is designed. Circularizers at the end of then helical windings produce a circular magnetic surface and provide improved confining and heating of the plasma. Reverse curvature sections formed in the end loops of the reaction tube provide increased plasma pressure for a given magnetic field pressure and thereby minimize the current flow in the helical windings. (AEC)

PROCESS FOR MAKING NEUTRON-ABSORBING BODIES

DOEpatents

Schippereit, G.H.; Lang, R.M.

1961-11-14

A process for making a control element for a nuclear reactor and the control element prepared by the process are described. Equally spaced, conically shaped depressions are formed in one face of a metal plate, spheres of boron of uniform size are placed in the depressions, another plate is welded on top of this place covering the depressions, and the joined plates are rolled to the desired thickness. (AEC)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, Jon; Hayes, Steven; Walters, L. C.

This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO 2 and UO 2-PuO 2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availabilitymore » are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.« less

ANALYSIS OF THE MOMENTS METHOD EXPERIMENT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kloster, R.L.

1959-09-01

Monte Cario calculations show the effects of a plane water-air boundary on both fast neutron and gamma dose rates. Multigroup diffusion theory calculation for a reactor source shows the effects of a plane water-air boundary on thermal neutron dose rate. The results of Monte Cario and multigroup calculations are compared with experimental values. The predicted boundary effect for fast neutrons of 7.3% agrees within 16% with the measured effect of 6.3%. The gamma detector did not measure a boundary effect because it lacked sensitivity at low energies. However, the effect predicted for gamma rays of 5 to 10% is asmore » large as that for neutrons. An estimate of the boundary effect for thermal neutrons from a PoBe source is obtained from the results of muitigroup diffusion theory calcuiations for a reactor source. The calculated boundary effect agrees within 13% with the measured values. (auth)« less

Immunological compatibility status of placenta-derived stem cells is mediated by scaffold 3D structure.

PubMed

Azizian, Sara; Khatami, Fatemeh; Modaresifar, Khashayar; Mosaffa, Nariman; Peirovi, Habibollah; Tayebi, Lobat; Bahrami, Soheyl; Redl, Heinz; Niknejad, Hassan

2018-02-23

Placenta-derived amniotic epithelial cells (AECs), a great cell source for tissue engineering and stem cell therapy, are immunologically inert in their native state; however, immunological changes in these cells after culture and differentiation have challenged their applications. The aim of this study was to investigate the effect of 2D and 3D scaffolds on human lymphocyte antigens (HLA) expression by AECs. The effect of different preparation parameters including pre-freezing time and temperature was evaluated on 3D chitosan-gelatine scaffolds properties. Evaluation of MHC class I, HLA-DR and HLA-G expression in AECs after 7 d culture on 2D bed and 3D scaffold of chitosan-gelatine showed that culture of AECs on the 2D substrate up-regulated MHC class I and HLA-DR protein markers on AECs surface and down-regulated HLA-G protein. In contrast, 3D scaffold did not increase protein expression of MHC class I and HLA-DR. Moreover, HLA-G protein expression remained unchanged in 3D culture. These results confirm that 3D scaffold can remain AECs in their native immunological state and modification of physical properties of the scaffold is a key regulator of immunological markers at the gene and protein expression levels; a strategy which circumvents rejection challenge of amniotic stem cells to be translated into the clinic.

Environmental friendly technology for aluminum electrolytic capacitors recycling from waste printed circuit boards.

PubMed

Wang, Jianbo; Xu, Zhenming

2017-03-15

up to now, the recycling of e-waste should be developed towards more depth and refinement to promote industrial production of e-waste resource recovery. in the present study, the recycling of aluminum electrolytic capacitors (AECs) from waste printed circuit boards (WPCBs) is focused on. First of all, AECs are disassembled from WPCBs by a self-designed machine; meanwhile, the disassembled AECs are subjected to an integrated process, involving heating treatment, crushing, sieving, and magnetic separating, to recover aluminum and iron; finally, the off-gas and residue generated during the aforementioned processes are analyzed to evaluate environmental risks. The results indicate that 96.52% and 98.68% of aluminum and iron, respectively, can be recovered from AECs under the optimal condition. The off-gas generated during the process is mainly composed of elements of C, H, and O, indicating that the off-gas is non-toxic and could be re-utilized as clean energy source. The residue according with toxicity characteristics leaching standard can be landfilled safely in sanitary landfill site. The present study provides an environmentally friendly and industrial application potential strategy to recycle AECs to promote e-waste recycling industry. Copyright © 2016 Elsevier B.V. All rights reserved.

Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gougar, Hans David

2015-10-01

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each ofmore » the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.« less

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki; Miura, Ryosuke

2015-09-30

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design.more » The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.« less

Impacts of Heterogeneous Recycle in Fast Reactors on Overall Fuel Cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Temitope A. Taiwo; Samuel E. Bays; Abdullatif M. Yacout

2011-03-01

A study in the United States has evaluated the attributes of the heterogeneous recycle approach for plutonium and minor actinide transmutation in fast reactor fuel cycles, with comparison to the homogeneous recycle approach, where pertinent. The work investigated the characteristics, advantages, and disadvantages of the approach in the overall fuel cycle, including reactor transmutation, systems and safety impacts, fuel separation and fabrication issues, and proliferation risk and transportation impacts. For this evaluation, data from previous and ongoing national studies on heterogeneous recycle were reviewed and synthesized. Where useful, information from international sources was included in the findings. The intent ofmore » the work was to provide a comprehensive assessment of the heterogeneous recycle approach at the current time.« less

FUEL SUBASSEMBLY CONSTRUCTION FOR RADIAL FLOW IN A NUCLEAR REACTOR

DOEpatents

Treshow, M.

1962-12-25

An assembly of fuel elements for a boiling water reactor arranged for radial flow of the coolant is described. The ingress for the coolant is through a central header tube, perforated with parallel circumferertial rows of openings each having a lip to direct the coolant flow downward. Around the central tube there are a number of equally spaced concentric trays, closely fitiing the central header tube. Cylindrical fuel elements are placed in a regular pattern around the central tube, piercing the trays. A larger tube encloses the arrangement, with space provided for upward flow of coolart beyond the edge of the trays. (AEC)

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

DOEpatents

Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

1962-08-14

A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

NEUTRONIC REACTOR

DOEpatents

Daniels, F.

1962-12-18

A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

METHOD AND APPARATUS FOR EARTH PENETRATION

DOEpatents

Adams, W.M.

1963-12-24

A nuclear reactor apparatus for penetrating into the earth's crust is described. The apparatus comprises a cylindrical nuclear core operating at a temperature that is higher than the melting temperature of rock. A high-density ballast member is coupled to the nuclear core such that the overall density of the core-ballast assembly is greater than the density of molten rock. The nuclear core is thermally insulated so that its heat output is constrained to flow axially, with radial heat flow being minimized. In operation, the apparatus is placed in contact with the earth's crust at the point desired to be penetrated. The heat output of the reactor melts the underlying rock, and the apparatus sinks through the resulting magma. The fuel loading of the reactor core determines the ultimate depth of crust penetration. (AEC)

The fast neutron fluence and the activation detector activity calculations using the effective source method and the adjoint function

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hep, J.; Konecna, A.; Krysl, V.

2011-07-01

This paper describes the application of effective source in forward calculations and the adjoint method to the solution of fast neutron fluence and activation detector activities in the reactor pressure vessel (RPV) and RPV cavity of a VVER-440 reactor. Its objective is the demonstration of both methods on a practical task. The effective source method applies the Boltzmann transport operator to time integrated source data in order to obtain neutron fluence and detector activities. By weighting the source data by time dependent decay of the detector activity, the result of the calculation is the detector activity. Alternatively, if the weightingmore » is uniform with respect to time, the result is the fluence. The approach works because of the inherent linearity of radiation transport in non-multiplying time-invariant media. Integrated in this way, the source data are referred to as the effective source. The effective source in the forward calculations method thereby enables the analyst to replace numerous intensive transport calculations with a single transport calculation in which the time dependence and magnitude of the source are correctly represented. In this work, the effective source method has been expanded slightly in the following way: neutron source data were performed with few group method calculation using the active core calculation code MOBY-DICK. The follow-up neutron transport calculation was performed using the neutron transport code TORT to perform multigroup calculations. For comparison, an alternative method of calculation has been used based upon adjoint functions of the Boltzmann transport equation. Calculation of the three-dimensional (3-D) adjoint function for each required computational outcome has been obtained using the deterministic code TORT and the cross section library BGL440. Adjoint functions appropriate to the required fast neutron flux density and neutron reaction rates have been calculated for several significant points within the RPV and RPV cavity of the VVER-440 reacto rand located axially at the position of maximum power and at the position of the weld. Both of these methods (the effective source and the adjoint function) are briefly described in the present paper. The paper also describes their application to the solution of fast neutron fluence and detectors activities for the VVER-440 reactor. (authors)« less

COAXIAL TUBE COUPLING

DOEpatents

Niemoth, H.R.

1963-02-26

BS>This patent shows a device for quickly coupling coaxial tubes in metal-to-metal fashion, so as to be suitable for use in a nuclear reactor. A threaded coliar urges a tapered metal extension on the outer coaxial tube into a tapered seat in the device and simultaneously exerts pressure through a coaxial helical spring so that a similar extension on the inner tube seats in a similar seat near the other end. (AEC)

NUCLEAR REACTOR FUEL ELEMENTS AND METHOD OF PREPARATION

DOEpatents

Kingston, W.E.; Kopelman, B.; Hausner, H.H.

1963-07-01

A fuel element consisting of uranium nitride and uranium carbide in the form of discrete particles in a solid coherent matrix of a metal such as steel, beryllium, uranium, or zirconium and clad with a metal such as steel, aluminum, zirconium, or beryllium is described. The element is made by mixing powdered uranium nitride and uranium carbide with powdered matrix metal, then compacting and sintering the mixture. (AEC)

FUEL COMPOSITION FOR NUCLEAR REACTORS

DOEpatents

Andersen, J.C.

1963-08-01

A process for making refractory nuclear fuel elements involves heating uranium and silicon powders in an inert atmosphere to 1600 to 1800 deg C to form USi/sub 3/; adding silicon carbide, carbon, 15% by weight of nickel and aluminum, and possibly also molybdenum and silicon powders; shaping the mixture; and heating to 1700 to 2050 deg C again in an inert atmosphere. Information on obtaining specific compositions is included. (AEC)

CONTINUOUS TREATMENT APPARATUS

DOEpatents

Erickson, E.E.

1962-05-15

An apparatus is described for dissolving a nuclear reactor fuel element in strong acid solutlon. The vapors and entrained liquid resulting from the violent reaction of dissolution are led into a reflux condenser which discharges, not directly back into the top of the reaction vessel in the conventional manner, but by a route leading to the bottom of the apparatus, thereby utilizing the energy of the reaction to bring about a circulation of the solution. (AEC)

APPARATUS FOR HEATING A PLASMA

DOEpatents

Stix, T.H.

1962-01-01

The system contemplates the use of ion cyclotron motions for transferring energy to a plasma immersed in a confining magnetic field such as is found in thermonuclear reactors of the stellarator class. Oppositely directed windings are provided for producing ion-accelerating fields having a time and spatial periodicity and these have the advantage of producing ion cyclotron motions without the development of space charges which preclude the efficient energy transfer to the plasma. (AEC)

Calculation and benchmarking of an azimuthal pressure vessel neutron fluence distribution using the BOXER code and scraping experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holzgrewe, F.; Hegedues, F.; Paratte, J.M.

1995-03-01

The light water reactor BOXER code was used to determine the fast azimuthal neutron fluence distribution at the inner surface of the reactor pressure vessel after the tenth cycle of a pressurized water reactor (PWR). Using a cross-section library in 45 groups, fixed-source calculations in transport theory and x-y geometry were carried out to determine the fast azimuthal neutron flux distribution at the inner surface of the pressure vessel for four different cycles. From these results, the fast azimuthal neutron fluence after the tenth cycle was estimated and compared with the results obtained from scraping test experiments. In these experiments,more » small samples of material were taken from the inner surface of the pressure vessel. The fast neutron fluence was then determined form the measured activity of the samples. Comparing the BOXER and scraping test results have maximal differences of 15%, which is very good, considering the factor of 10{sup 3} neutron attenuation between the reactor core and the pressure vessel. To compare the BOXER results with an independent code, the 21st cycle of the PWR was also calculated with the TWODANT two-dimensional transport code, using the same group structure and cross-section library. Deviations in the fast azimuthal flux distribution were found to be <3%, which verifies the accuracy of the BOXER results.« less

Alternate Tritium Production Methods Using A Liquid Lithium Target

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilson, J.

For over 60 years, the Savannah River Site’s primary mission has been the production of tritium. From the beginning, the Savannah River National Laboratory (SRNL) has provided the technical foundation to ensure the successful execution of this critical defense mission. SRNL has developed most of the processes used in the tritium mission and provides the research and development necessary to supply this critical component. This project was executed by first developing reactor models that could be used as a neutron source. In parallel to this development calculations were carried out testing the feasibility of accelerator technologies that could also bemore » used for tritium production. Targets were designed with internal moderating material and optimized target was calculated to be capable of 3000 grams using a 1400 MWt sodium fast reactor, 850 grams using a 400 MWt sodium fast reactor, and 100 grams using a 62 MWt reactor, annually.« less

MULTIPLIER CIRCUIT

DOEpatents

Chase, R.L.

1963-05-01

An electronic fast multiplier circuit utilizing a transistor controlled voltage divider network is presented. The multiplier includes a stepped potentiometer in which solid state or transistor switches are substituted for mechanical wipers in order to obtain electronic switching that is extremely fast as compared to the usual servo-driven mechanical wipers. While this multiplier circuit operates as an approximation and in steps to obtain a voltage that is the product of two input voltages, any desired degree of accuracy can be obtained with the proper number of increments and adjustment of parameters. (AEC)

WATER TREATMENT

DOEpatents

Pitman, R.W.; Conley, W.R. Jr.

1962-12-01

An automated system for adding clarifying chemicals to water in a water treatment plant is described. To a sample of the floc suspension polyacrylamide or similar filter aid chemicals are added, and the sample is then put through a fast filter. The resulting filtrate has the requisite properties for monitoring in an optical turbidimeter to control the automated system. (AEC)

Investigation of the characteristics of Automatic Exposure Control (AEC) of a Computed Tomography (CT) scanner by utilising cylindrical and anthropomorphic phantoms

NASA Astrophysics Data System (ADS)

Rulaidi, W. E. P.; Huri, M. S. N.; Ng, K. H.

2017-05-01

One method to optimise the use of x-rays in CT and hence a reduction in patient dose is the application of automatic exposure control (AEC). This study measured the effective mAs, image noise and volume CT dose index (CTDIvol) as the result of changing the AEC index on a Siemens Somatom Definition 64 slices dual source CT scanner. The scans were performed on four phantoms of different geometries, namely the 16 and 32 cm cylindrical CTDI phantoms and two anthropomorphic phantoms, RANDO (20 cm effective diameter) and ATOM (19.8 cm effective diameter). Results showed that the effective mAs increased with increasing tube potential (kVp) and Quality Reference mAs (QRM), therefore increasing CTDIvol while reducing image noise. Meanwhile, no changes of radiation dose and image noise were observed when the pitch was increased. However, for the largest phantom (32 cm effective diameter), a constant effective mAs was found between 120 and 140 kVp. The same trend was also found with increasing QRM from 300 mAs to 400 mAs suggesting a certain limitation of the AEC has been reached. In conclusion, this study showed that AEC is affected by kVp and QRM but not by pitch selection. Further work is required to quantify the characteristics of the AEC system in relation to the mentioned parameters for better optimisation.

Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

PubMed

Khattab, K; Sulieman, I

2009-04-01

The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

Fast growth rate of epitaxial β-Ga2O3 by close coupled showerhead MOCVD

NASA Astrophysics Data System (ADS)

Alema, Fikadu; Hertog, Brian; Osinsky, Andrei; Mukhopadhyay, Partha; Toporkov, Mykyta; Schoenfeld, Winston V.

2017-10-01

We report on the growth of epitaxial β-Ga2O3 thin films on c-plane sapphire substrates using a close coupled showerhead MOCVD reactor. Ga(DPM)3 (DPM = dipivaloylmethanate), triethylgallium (TEGa) and trimethylgallium (TMGa) metal organic (MO) precursors were used as Ga sources and molecular oxygen was used for oxidation. Films grown from each of the Ga sources had high growth rates, with up to 10 μm/hr achieved using a TMGa precursor at a substrate temperature of 900 °C. As confirmed by X-ray diffraction, the films grown from each of the Ga sources were the monoclinic (2 bar 0 1) oriented β-Ga2O3 phase. The optical bandgap of the films was also estimated to be ∼4.9 eV. The fast growth rate of β-Ga2O3 thin films obtained using various Ga-precursors has been achieved due to the close couple showerhead design of the MOCVD reactor as well as the separate injection of oxygen and MO precursors, preventing the premature oxidation of the MO sources. These results suggest a pathway to overcoming the long-standing challenge of realizing fast growth rates for Ga2O3 using the MOCVD method.

SEPARATION OF METAL VALUES FROM NUCLEAR REACTOR

DOEpatents

Campbell, D.O.; Cathers, G.I.

1962-06-19

A method is given for separating beryllium fluoride and an alkali metal fluoride from a mixture containing same and rare earth fluorides. The method comprises contacting said mixture with a liquid hydrogen fluoride solvent containing no more than about 30 per cent water by weight and saturated with a fluoride salt characterized by its solubility in anhydrous hydrogen fluoride for a period of time sufficient to dissolve said beryllium fluoride in said solvent. (AEC)

IRRADIATION METHOD AND APPARATUS

DOEpatents

Cabell, C.P.

1962-12-18

A method and apparatus are described for changing fuel bodies into a process tube of a reactor. According to this method fresh fuel elements are introduced into one end of the tube forcing used fuel elements out the other end. When sufficient fuel has been discharged, a reel and tape arrangement is employed to pull the column of bodies back into the center of the tube. Due provision is made for providing shielding in the tube. (AEC)

Antineutrino analysis for continuous monitoring of nuclear reactors: Sensitivity study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stewart, Christopher; Erickson, Anna

This paper explores the various contributors to uncertainty on predictions of the antineutrino source term which is used for reactor antineutrino experiments and is proposed as a safeguard mechanism for future reactor installations. The errors introduced during simulation of the reactor burnup cycle from variation in nuclear reaction cross sections, operating power, and other factors are combined with those from experimental and predicted antineutrino yields, resulting from fissions, evaluated, and compared. The most significant contributor to uncertainty on the reactor antineutrino source term when the reactor was modeled in 3D fidelity with assembly-level heterogeneity was found to be the uncertaintymore » on the antineutrino yields. Using the reactor simulation uncertainty data, the dedicated observation of a rigorously modeled small, fast reactor by a few-ton near-field detector was estimated to offer reduction of uncertainty on antineutrino yields in the 3.0–6.5 MeV range to a few percent for the primary power-producing fuel isotopes, even with zero prior knowledge of the yields.« less

Nodal weighting factor method for ex-core fast neutron fluence evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chiang, R. T.

The nodal weighting factor method is developed for evaluating ex-core fast neutron flux in a nuclear reactor by utilizing adjoint neutron flux, a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV, the unit fission source, and relative assembly nodal powers. The method determines each nodal weighting factor for ex-core neutron fast flux evaluation by solving the steady-state adjoint neutron transport equation with a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV as the adjoint source, by integrating the unit fission source with a typical fission spectrum to the solved adjointmore » flux over all energies, all angles and given nodal volume, and by dividing it with the sum of all nodal weighting factors, which is a normalization factor. Then, the fast neutron flux can be obtained by summing the various relative nodal powers times the corresponding nodal weighting factors of the adjacent significantly contributed peripheral assembly nodes and times a proper fast neutron attenuation coefficient over an operating period. A generic set of nodal weighting factors can be used to evaluate neutron fluence at the same location for similar core design and fuel cycles, but the set of nodal weighting factors needs to be re-calibrated for a transition-fuel-cycle. This newly developed nodal weighting factor method should be a useful and simplified tool for evaluating fast neutron fluence at selected locations of interest in ex-core components of contemporary nuclear power reactors. (authors)« less

SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

DOEpatents

Schluderberg, D.C.; Ryon, J.W.

1962-05-01

A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

The new postirradiation examination facility of the Atomic Energy Corporation of South Africa

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walt, P.L. van der; Aspeling, J.C.; Jonker, W.D.

1992-01-01

The Pelindaba Hot Cell Complex (HCC) forms an important part of the infrastructure and support services of the Atomic Energy Corporation (AEC) of South Africa. It is a comprehensive, one-stop facility designed to make South Africa self-sufficient in the fields of spent-fuel qualification and verification, reactor pressure vessel surveillance program testing, ad hoc failure analyses for the nuclear power industry, and research and development studies in conjunction with the Safari I material test reactor (MTR) and irradiation rigs. Local technology and expertise was used for the design and construction of the HCC, which start up in 1980. The facility wasmore » commissioned in 1990.« less

THERMAL RELAY DEVICE

DOEpatents

Murdoch, R.O.; Record, F.A.

1963-01-29

This invention relates to a fast-acting spring-loaded electrical switch which can break a 1500-volt circuit in one millisecond without arcing. In particular, a springloaded shorting bar is held in tension by a fusible wire. Passage of an electrical current pulse through the fusible wire breaks the fuse thereby releasing the shorting bar to open one and close another electrical circuit. (AEC)

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated meltingmore » of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.« less

Research Program of a Super Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

FAST NEUTRON DOSIMETER FOR HIGH TEMPERATURE OPERATION BY MEASUREMENT OF THE AMOUNT OF CESIUM 137 FORMED FROM A THORIUM WIRE

DOEpatents

McCune, D.A.

1964-03-17

A method and device for measurement of integrated fast neutron flux in the presence of a large thermal neutron field are described. The device comprises a thorium wire surrounded by a thermal neutron attenuator that is, in turn, enclosed by heat-resistant material. The method consists of irradiating the device in a neutron field whereby neutrons with energies in excess of 1.1 Mev cause fast fissions in the thorium, then removing the thorium wire, separating the cesium-137 fission product by chemical means from the thorium, and finally counting the radioactivity of the cesium to determine the number of fissions which have occurred so that the integrated fast flux may be obtained. (AEC)

The Atomic Energy Commission's Annual Report to Congress for 1962. Major Activities in the Atomic Energy Programs, January - December 1962

DOE Office of Scientific and Technical Information (OSTI.GOV)

Seaborg, Glenn T.

1963-01-31

The document represents the 1962 Annual Report of the Atomic Energy Commission (AEC) to Congress. This year's report opens with a section of Highlights of the Atomic Energy Programs of 1962, followed by five parts: Part One, Commission Activities; Part Two, Nuclear Reactor Programs; Part Three, Production and Weapons Programs; Part Four, Other Major Programs; and Part Five, The Regulatory Program. Sixteen appendices are also included.

Development of Ultra-Fine Multigroup Cross Section Library of the AMPX/SCALE Code Packages

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeon, Byoung Kyu; Sik Yang, Won; Kim, Kang Seog

The Consortium for Advanced Simulation of Light Water Reactors Virtual Environment for Reactor Applications (VERA) neutronic simulator MPACT is being developed by Oak Ridge National Laboratory and the University of Michigan for various reactor applications. The MPACT and simplified MPACT 51- and 252-group cross section libraries have been developed for the MPACT neutron transport calculations by using the AMPX and Standardized Computer Analyses for Licensing Evaluations (SCALE) code packages developed at Oak Ridge National Laboratory. It has been noted that the conventional AMPX/SCALE procedure has limited applications for fast-spectrum systems such as boiling water reactor (BWR) fuels with very highmore » void fractions and fast reactor fuels because of its poor accuracy in unresolved and fast energy regions. This lack of accuracy can introduce additional error sources to MPACT calculations, which is already limited by the Bondarenko approach for resolved resonance self-shielding calculation. To enhance the prediction accuracy of MPACT for fast-spectrum reactor analyses, the accuracy of the AMPX/SCALE code packages should be improved first. The purpose of this study is to identify the major problems of the AMPX/SCALE procedure in generating fast-spectrum cross sections and to devise ways to improve the accuracy. For this, various benchmark problems including a typical pressurized water reactor fuel, BWR fuels with various void fractions, and several fast reactor fuels were analyzed using the AMPX 252-group libraries. Isotopic reaction rates were determined by SCALE multigroup (MG) calculations and compared with continuous energy (CE) Monte Carlo calculation results. This reaction rate analysis revealed three main contributors to the observed differences in reactivity and reaction rates: (1) the limitation of the Bondarenko approach in coarse energy group structure, (2) the normalization issue of probability tables, and (3) neglect of the self-shielding effect of resonance-like cross sections at high energy range such as (n,p) cross section of Cl35. The first error source can be eliminated by an ultra-fine group (UFG) structure in which the broad scattering resonances of intermediate-weight nuclides can be represented accurately by a piecewise constant function. A UFG AMPX library was generated with modified probability tables and tested against various benchmark problems. The reactivity and reaction rates determined with the new UFG AMPX library agreed very well with respect to Monte Carlo Neutral Particle (MCNP) results. To enhance the lattice calculation accuracy without significantly increasing the computational time, performing the UFG lattice calculation in two steps was proposed. In the first step, a UFG slowing-down calculation is performed for the corresponding homogenized composition, and UFG cross sections are collapsed into an intermediate group structure. In the second step, the lattice calculation is performed for the intermediate group level using the condensed group cross sections. A preliminary test showed that the condensed library reproduces the results obtained with the UFG cross section library. This result suggests that the proposed two-step lattice calculation approach is a promising option to enhance the applicability of the AMPX/SCALE system to fast system analysis.« less

Comparison of Measured and Estimated CT Organ Doses for Modulated and Fixed Tube Current:: A Human Cadaver Study.

PubMed

Padole, Atul; Deedar Ali Khawaja, Ranish; Otrakji, Alexi; Zhang, Da; Liu, Bob; Xu, X George; Kalra, Mannudeep K

2016-05-01

The aim of this study was to compare the directly measured and the estimated computed tomography (CT) organ doses obtained from commercial radiation dose-tracking (RDT) software for CT performed with modulated tube current or automatic exposure control (AEC) technique and fixed tube current (mAs). With the institutional review board (IRB) approval, the ionization chambers were surgically implanted in a human cadaver (88 years old, male, 68 kg) in six locations such as liver, stomach, colon, left kidney, small intestine, and urinary bladder. The cadaver was scanned with routine abdomen pelvis protocol on a 128-slice, dual-source multidetector computed tomography (MDCT) scanner using both AEC and fixed mAs. The effective and quality reference mAs of 100, 200, and 300 were used for AEC and fixed mAs, respectively. Scanning was repeated three times for each setting, and measured and estimated organ doses (from RDT software) were recorded (N = 3*3*2 = 18). Mean CTDIvol for AEC and fixed mAs were 4, 8, 13 mGy and 7, 14, 21 mGy, respectively. The most estimated organ doses were significantly greater (P < 0.01) than the measured organ doses for both AEC and fixed mAs. At AEC, the mean estimated organ doses (for six organs) were 14.7 mGy compared to mean measured organ doses of 12.3 mGy. Similarly, at fixed mAs, the mean estimated organ doses (for six organs) were 24 mGy compared to measured organ doses of 22.3 mGy. The differences among the measured and estimated organ doses were higher for AEC technique compared to the fixed mAs for most organs (P < 0.01). The most CT organ doses estimated from RDT software are greater compared to directly measured organ doses, particularly when AEC technique is used for CT scanning. Copyright © 2016 The Association of University Radiologists. Published by Elsevier Inc. All rights reserved.

Safety survey report EBR-II safety survey, ANL-west health protection, industrial safety and fire protection survey

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunbar, K.A.

1972-01-10

A safety survey covering the disciplines of Reactor Safety, Nuclear Criticality Safety, Health Protection and Industrial Safety and Fire Protection was conducted at the ANL-West EBR-II FEF Complex during the period January 10-18, 1972. In addition, the entire ANL-West site was surveyed for Health Protection and Industrial Safety and Fire Protection. The survey was conducted by members of the AEC Chicago Operations Office, a member of RDT-HQ and a member of the RDT-ID site office. Eighteen recommendations resulted from the survey, eleven in the area of Industrial Safety and Fire Protection, five in the area of Reactor Safety and twomore » in the area of Nuclear Criticality Safety.« less

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Bassett, C.H.

1961-11-21

A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

MODULAR CORE UNITS FOR A NEUTRONIC REACTOR

DOEpatents

Gage, J.F. Jr.; Sherer, D.B.

1964-04-01

A modular core unit for use in a nuclear reactor is described. Many identical core modules can be placed next to each other to make up a complete core. Such a module includes a cylinder of moderator material surrounding a fuel- containing re-entrant coolant channel. The re-entrant channel provides for the circulation of coolant such as liquid sodium from one end of the core unit, through the fuel region, and back out through the same end as it entered. Thermal insulation surrounds the moderator exterior wall inducing heat to travel inwardly to the coolant channel. Spaces between units may be used to accommodate control rods and support structure, which may be cooled by a secondary gas coolant, independently of the main coolant. (AEC)

FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Dickson, J.J.

1963-09-24

A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooledmore » fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the current state of knowledge is extensive, and in most areas may be sufficient. Several knowledge gaps were identified, such as uncertainty in release from molten fuel and availability of thermodynamic data for lanthanides and actinides in liquid sodium. However, the overall findings suggest that high retention rates can be expected within the fuel and primary sodium for all radionuclides other than noble gases.« less

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grabaskas, David; Bucknor, Matthew; Jerden, James

2016-02-01

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish releasemore » fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.« less

SEPARATION APPARATUS

DOEpatents

Huff, J.B.

1962-03-13

A furnace apparatus is designed for treating a nuclear reactor waste solution. The solution is sprayed onto a bed of burning petroleum coke which expels water, the more volatile fission products, and nitrogen oxides. Next, chlorine gas is introduced from below which causes aluminum to volatilize as aluminum chloride and along with it certain fission products including Nb/sup 95/ and Zr/sup 95/. These lose their radioactivity within four years and the long- lived radioactivity remains with the ash, which is stored. (AEC) V) nitrate. (P.C.H.)

ALLOY COMPOSITION FOR NEUTRONIC REACTOR CONTROL RODS

DOEpatents

Lustman, B.; Losco, E.F.; Snyder, H.J.; Eggleston, R.R.

1963-01-22

This invention relates to alloy compositons suitable as cortrol rod material consisting of, by weight, from 85% to 85% Ag, from 2% to 20% In, from up to 10% of Cd, from up to 5% Sn, and from up to 1.5% Al, the amount of each element employed being determined by the equation X + 2Y + 3Z + 3W + 4V = 1.4 and less, where X, Y, Z, W, and V represent the atom fractions of the elements Ag, Cd, In, Al and Sn. (AEC)

The 14 MeV Neutron Irradiation Facility in MARIA Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prokopowicz, R.; Pytel, K.; Dorosz, M.

2015-07-01

The MARIA reactor with thermal neutron flux density up to 3x10{sup 14} cm{sup -2} s{sup -1} and a number of vertical channels is well suited to material testing by thermal neutron treatment. Beside of that some fast neutron irradiation facilities are operated in MARIA reactor as well. One of them is thermal to 14 MeV neutron converter launched in 2014. It is especially devoted to fusion devices material testing irradiation. The ITER and DEMO research thermonuclear facilities are to be run using the deuterium - tritium fusion reaction. Fast neutrons (of energy approximately 14 MeV) resulting from the reaction aremore » essential to carry away the released thermonuclear energy and to breed tritium. However, constructional materials of which thermonuclear reactors are to be built must be specially selected to survive intense fluxes of fast neutrons. Strong sources of 14 MeV neutrons are needed if research on resistance of candidate materials to such fluxes is to be carried out effectively. Nuclear reactor-based converter capable to convert thermal neutrons into 14 MeV fast neutrons may be used to that purpose. The converter based on two stage nuclear reaction on lithium-6 and deuterium compounds leading to 14 MeV neutron production. The reaction chain is begun by thermal neutron capture by lithium-6 nucleus resulted in triton release. The neutron and triton transport calculations have been therefore carried-out to estimate the thermal to 14 MeV neutron conversion efficiency and optimize converter construction. The usable irradiation space of ca. 60 cm{sup 3} has been obtained. The released energy have been calculated. Heat transport has been asses to ensure proper device cooling. A set of thermocouples has been installed in converter to monitor its temperature distribution on-line. Influence of converter on reactor operation has been studied. Safety analyses of steady states and transients have been done. Performed calculations and analyses allow designing the converter and formulate its operation limits and conditions. During first tested operation of the converter the 14 MeV neutron flux density was estimated to 10{sup 9} cm{sup -2} s{sup -1}, whereas fast fission neutrons inside converter achieved 10{sup 12} cm{sup -2} s{sup -1}, and thermal neutrons were reduced down to 109 cm-2 s-1. Taking into account the feasibility of almost incessant converter operation for a number of months, its arisen as one of the most powerful (in terms of fluence), currently available 14 MeV neutron source. Such a converter currently under operation in the MARIA reactor core will be presented. (authors)« less

Revised FINAL–REPORT NO. 2: INDEPENDENT CONFIRMATORY SURVEY SUMMARY AND RESULTS FOR THE ENRICO FERMI ATOMIC POWER PLANT, UNIT 1, NEWPORT, MICHIGAN (DOCKET NO. 50 16; RFTA 10-004) 2018-SR-02-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Erika Bailey

2011-10-27

The Enrico Fermi Atomic Power Plant, Unit 1 (Fermi 1) was a fast breeder reactor design that was cooled by sodium and operated at essentially atmospheric pressure. On May 10, 1963, the Atomic Energy Commission (AEC) granted an operating license, DPR-9, to the Power Reactor Development Company (PRDC), a consortium specifically formed to own and operate a nuclear reactor at the Fermi 1 site. The reactor was designed for a maximum capability of 430 megawatts (MW); however, the maximum reactor power with the first core loading (Core A) was 200 MW. The primary system was filled with sodium in Decembermore » 1960 and criticality was achieved in August 1963. The reactor was tested at low power during the first couple years of operation. Power ascension testing above 1 MW commenced in December 1965 immediately following the receipt of a high-power operating license. In October 1966 during power ascension, zirconium plates at the bottom of the reactor vessel became loose and blocked sodium coolant flow to some fuel subassemblies. Two subassemblies started to melt and the reactor was manually shut down. No abnormal releases to the environment occurred. Forty-two months later after the cause had been determined, cleanup completed, and the fuel replaced, Fermi 1 was restarted. However, in November 1972, PRDC made the decision to decommission Fermi 1 as the core was approaching its burn-up limit. The fuel and blanket subassemblies were shipped off-site in 1973. Following that, the secondary sodium system was drained and sent off-site. The radioactive primary sodium was stored on-site in storage tanks and 55 gallon (gal) drums until it was shipped off-site in 1984. The initial decommissioning of Fermi 1 was completed in 1975. Effective January 23, 1976, DPR-9 was transferred to the Detroit Edison Company (DTE) as a 'possession only' license (DTE 2010a). This report details the confirmatory activities performed during the second Oak Ridge Institute for Science and Education (ORISE) site visit to Fermi 1 in November 2010. The survey was strategically planned during a Unit 2 (Fermi 2) outage to take advantage of decreased radiation levels that were observed and attributed to Fermi 2 from the operating unit during the first site visit. However, during the second visit there were elevated radiation levels observed and attributed to the partially dismantled Fermi 1 reactor vessel and a waste storage box located on the 3rd floor of the Fermi 1 Turbine Building. Confirmatory surveys (unshielded) performed directly in the line of sight of these areas were affected. The objective of the confirmatory survey was to verify that the final radiological conditions were accurately and adequately described in Final Status Survey (FSS) documentation, relative to the established release criteria. This objective was achieved by performing document reviews, as well as independent measurements and sampling. Specifically, documentation of the planning, implementation, and results of the FSS were evaluated; side-by-side FSS measurement and source comparisons were performed; site areas were evaluated relative to appropriate FSS classification; and areas were assessed for residual, undocumented contamination.« less

Fast particles in a steady-state compact FNS and compact ST reactor

NASA Astrophysics Data System (ADS)

Gryaznevich, M. P.; Nicolai, A.; Buxton, P.

2014-10-01

This paper presents results of studies of fast particles (ions and alpha particles) in a steady-state compact fusion neutron source (CFNS) and a compact spherical tokamak (ST) reactor with Monte-Carlo and Fokker-Planck codes. Full-orbit simulations of fast particle physics indicate that a compact high field ST can be optimized for energy production by a reduction of the necessary (for the alpha containment) plasma current compared with predictions made using simple analytic expressions, or using guiding centre approximation in a numerical code. Alpha particle losses may result in significant heating and erosion of the first wall, so such losses for an ST pilot plant have been calculated and total and peak wall loads dependence on the plasma current has been studied. The problem of dilution has been investigated and results for compact and big size devices are compared.

China’s Ace in the Hole Rare Earth Elements

DTIC Science & Technology

2010-01-01

before losing magnetism. 11 Europium sesquioxide (Eu203) has been tested as neutron absorbers for control rods in (fast breeder ) nuclear reactors ...sources of rare earth around the world, it could take anywhere from 10 to 15 years from the time of discovery to begin a full- scale rare earth...television/computer screens, it is being studied for possible use in nuclear reactors .11 Erbium is used as an amplifier for fiber optic data

Assessment Environment for Complex Systems Software Guide

NASA Technical Reports Server (NTRS)

2013-01-01

This Software Guide (SG) describes the software developed to test the Assessment Environment for Complex Systems (AECS) by the West Virginia High Technology Consortium (WVHTC) Foundation's Mission Systems Group (MSG) for the National Aeronautics and Space Administration (NASA) Aeronautics Research Mission Directorate (ARMD). This software is referred to as the AECS Test Project throughout the remainder of this document. AECS provides a framework for developing, simulating, testing, and analyzing modern avionics systems within an Integrated Modular Avionics (IMA) architecture. The purpose of the AECS Test Project is twofold. First, it provides a means to test the AECS hardware and system developed by MSG. Second, it provides an example project upon which future AECS research may be based. This Software Guide fully describes building, installing, and executing the AECS Test Project as well as its architecture and design. The design of the AECS hardware is described in the AECS Hardware Guide. Instructions on how to configure, build and use the AECS are described in the User's Guide. Sample AECS software, developed by the WVHTC Foundation, is presented in the AECS Software Guide. The AECS Hardware Guide, AECS User's Guide, and AECS Software Guide are authored by MSG. The requirements set forth for AECS are presented in the Statement of Work for the Assessment Environment for Complex Systems authored by NASA Dryden Flight Research Center (DFRC). The intended audience for this document includes software engineers, hardware engineers, project managers, and quality assurance personnel from WVHTC Foundation (the suppliers of the software), NASA (the customer), and future researchers (users of the software). Readers are assumed to have general knowledge in the field of real-time, embedded computer software development.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

2009-08-01

This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

Low-power lead-cooled fast reactor loaded with MOX-fuel

NASA Astrophysics Data System (ADS)

Sitdikov, E. R.; Terekhova, A. M.

2017-01-01

Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

DOEpatents

Stengel, F.G.

1963-12-24

A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

REACTOR MODERATOR STRUCTURE

DOEpatents

Fraas, A.P.; Tudor, J.J.

1963-08-01

An improved moderator structure for nuclear reactors consists of moderator blocks arranged in horizontal layers to form a multiplicity of vertically stacked columns of blocks. The blocks in each vertical column are keyed together, and a ceramic grid is disposed between each horizontal layer of blocks. Pressure plates cover- the lateral surface of the moderator structure in abutting relationship with the peripheral terminal lengths of the ceramic grids. Tubular springs are disposed between the pressure plates and a rigid external support. The tubular springs have their axes vertically disposed to facilitate passage of coolant gas through the springs and are spaced apart a selected distance such that at sonae preselected point of spring deflection, the sides of the springs will contact adjacent springs thereby causing a large increase in resistance to further spring deflection. (AEC)

COMPARTMENTED REACTOR FUEL ELEMENT

DOEpatents

Cain, F.M. Jr.

1962-09-11

A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buchanan, J.R.; Keilholtz, G.W.

This report discusses the current status of liquid-metal fast breeder (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1960 through 1969. The bibliography consists of 1560 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Honma, George

The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will bemore » used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).« less

Alternative Sources of Adult Stem Cells: Human Amniotic Membrane

NASA Astrophysics Data System (ADS)

Wolbank, Susanne; van Griensven, Martijn; Grillari-Voglauer, Regina; Peterbauer-Scherb, Anja

Human amniotic membrane is a highly promising cell source for tissue engineering. The cells thereof, human amniotic epithelial cells (hAEC) and human amniotic mesenchymal stromal cells (hAMSC), may be immunoprivileged, they represent an early developmental status, and their application is ethically uncontroversial. Cell banking strategies may use freshly isolated cells or involve in vitro expansion to increase cell numbers. Therefore, we have thoroughly characterized the effect of in vitro cultivation on both phenotype and differentiation potential of hAEC. Moreover, we present different strategies to improve expansion including replacement of animal-derived supplements by human platelet products or the introduction of the catalytic subunit of human telomerase to extend the in vitro lifespan of amniotic cells. Characterization of the resulting cultures includes phenotype, growth characteristics, and differentiation potential, as well as immunogenic and immunomodulatory properties.

Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A., E-mail: Azizov-EA@nrcki.ru

2015-12-15

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel canmore » be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.« less

Amniotic Epithelial Cells: A New Tool to Combat Aging and Age-Related Diseases?

PubMed Central

Di Germanio, Clara; Bernier, Michel; de Cabo, Rafael; Barboni, Barbara

2016-01-01

The number of elderly people is growing at an unprecedented rate and this increase of the aging population is expected to have a direct impact on the incidence of age-related diseases and healthcare-associated costs. Thus, it is imperative that new tools are developed to fight and slow age-related diseases. Regenerative medicine is a promising strategy for the maintenance of health and function late in life; however, stem cell-based therapies face several challenges including rejection and tumor transformation. As an alternative, the placenta offers an extraordinary source of fetal stem cells, including the amniotic epithelial cells (AECs), which retain some of the characteristics of embryonic stem cells, but show low immunogenicity, together with immunomodulatory and anti-inflammatory activities. Because of these characteristics, AECs have been widely utilized in regenerative medicine. This perspective highlights different mechanisms triggered by transplanted AECs that could be potentially useful for anti-aging therapies, which include: Graft and differentiation for tissue regeneration in age-related settings, anti-inflammatory behavior to combat “inflammaging,” anti-tumor activity, direct lifespan and healthspan extension properties, and possibly rejuvenation in a manner reminiscent of heterochronic parabiosis. Here, we critically discuss benefits and limitation of AECs-based therapies in age-related diseases. PMID:27921031

Fast formation of aerobic granules by combining strong hydraulic selection pressure with overstressed organic loading rate.

PubMed

Liu, Yong-Qiang; Tay, Joo-Hwa

2015-09-01

The combined strong hydraulic selection pressure (HSP) with overstressed organic loading rate (OLR) as a fast granulation strategy was used to enhance aerobic granulation. To investigate the wide applicability of this strategy to different scenarios and its relevant mechanism, different settling times, different inoculums, different exchange ratios, different reactor configurations, and different shear force were used in this study. It was found that clear granules were formed within 24 h and steady state reached within three days when the fast granulation strategy was used in a lab-scale reactor seeded with well settled activated sludge (Reactor 2). However, granules appeared after 2-week operation and reached steady state after one month at the traditional step-wise decreased settling time from 20 to 2 min with OLR of 6 g COD/L·d (Reactor 1). With the fast granulation strategy, granules appeared within 24 h even with bulking sludge as seed to start up Reactor 3, but 6-day lag phase was observed compared with Reactor 2. Both Reactor 2 and Reactor 3 experienced sigmoidal growth curve in terms of biomass accumulation and granule size increase after granulation. In addition, the reproducible results in pilot-scale reactors (Reactor 5 and Reactor 6) with diameter of 20 cm and height/diameter ratio (H/D) of 4 further proved that reactor configuration and fluid flow pattern had no effect on the aerobic granulation when the fast granulation strategy was employed, but biomass accumulation experienced a short lag phase too in Reactor 5 and Reactor 6. Although overstressed OLR was favorable for fast granulation, it also led to the fluffy granules after around two-week operation. However, the stable 6-month operation of Reactor 3 demonstrated that the rapidly formed granules were able to maintain long-term stability by reducing OLR from 12 g COD/L·d to 6 g COD/L·d. A mechanism of fast granulation with the strategy of combined strong HSP and OLR was proposed to explain results and guide the operation with this fast strategy. Copyright © 2015 Elsevier Ltd. All rights reserved.

Conceptual Design of Low-Temperature Hydrogen Production and High-Efficiency Nuclear Reactor Technology

NASA Astrophysics Data System (ADS)

Fukushima, Kimichika; Ogawa, Takashi

Hydrogen, a potential alternative energy source, is produced commercially by methane (or LPG) steam reforming, a process that requires high temperatures, which are produced by burning fossil fuels. However, as this process generates large amounts of CO2, replacement of the combustion heat source with a nuclear heat source for 773-1173K processes has been proposed in order to eliminate these CO2 emissions. In this paper, a novel method of nuclear hydrogen production by reforming dimethyl ether (DME) with steam at about 573K is proposed. From a thermodynamic equilibrium analysis of DME steam reforming, the authors identified conditions that provide high hydrogen production fraction at low pressure and temperatures of about 523-573K. By setting this low-temperature hydrogen production process upstream from a turbine and nuclear reactor at about 573K, the total energy utilization efficiency according to equilibrium mass and heat balance analysis is about 50%, and it is 75%for a fast breeder reactor (FBR), where turbine is upstream of the reformer.

Deep-Earth reactor: nuclear fission, helium, and the geomagnetic field.

PubMed

Hollenbach, D F; Herndon, J M

2001-09-25

Geomagnetic field reversals and changes in intensity are understandable from an energy standpoint as natural consequences of intermittent and/or variable nuclear fission chain reactions deep within the Earth. Moreover, deep-Earth production of helium, having (3)He/(4)He ratios within the range observed from deep-mantle sources, is demonstrated to be a consequence of nuclear fission. Numerical simulations of a planetary-scale geo-reactor were made by using the SCALE sequence of codes. The results clearly demonstrate that such a geo-reactor (i) would function as a fast-neutron fuel breeder reactor; (ii) could, under appropriate conditions, operate over the entire period of geologic time; and (iii) would function in such a manner as to yield variable and/or intermittent output power.

SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

DOEpatents

Dickinson, R.W.

1963-03-01

This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

A Poor Buy

ERIC Educational Resources Information Center

Cochran, Thomas B.; And Others

1975-01-01

Both economic and environmental factors are affecting the further development of liquid metal fast breeder reactors. Rising costs of plant production and the possible environmental effects of the breeder system have caused the U.S. to consider carefully this method of energy production. Alternative energy sources are being studied also. (MA)

The benefits of an advanced fast reactor fuel cycle for plutonium management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hannum, W.H.; McFarlane, H.F.; Wade, D.C.

1996-12-31

The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium andmore » long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a `focus area` for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed.« less

FUEL ELEMENT

DOEpatents

Bean, R.W.

1963-11-19

A ceramic fuel element for a nuclear reactor that has improved structural stability as well as improved cooling and fission product retention characteristics is presented. The fuel element includes a plurality of stacked hollow ceramic moderator blocks arranged along a tubular raetallic shroud that encloses a series of axially apertured moderator cylinders spaced inwardly of the shroud. A plurality of ceramic nuclear fuel rods are arranged in the annular space between the shroud and cylinders of moderator and appropriate support means and means for directing gas coolant through the annular space are also provided. (AEC)

PROCESS OF MAKING SHAPED FUEL FOR NUCLEAR REACTORS

DOEpatents

O'Leary, W.J.; Fisher, E.A.

1964-02-11

A process for making uranium dioxide fuel of great strength, density, and thermal conductivity by mixing it with 0.1 to 1% of a densifier oxide (tin, aluminum, zirconium, ferric, zinc, chromium, molybdenum, titanium, or niobium oxide) and with a plasticizer (0.5 to 3% of bentonite and 0.05 to 2% of methylcellulose, propylene glycol alginate, or ammonium alginate), compacting the mixture obtained, and sintering the bodies in an atmosphere of carbon monoxide or carbon dioxide, with or without hydrogen, or of a nitrogen-hydrogen mixture is described. (AEC)

NEUTRONIC REACTION SYSTEM

DOEpatents

Wigner, E.P.

1963-09-01

A nuclear reactor system is described for breeding fissionable material, including a heat-exchange tank, a high- and a low-pressure chamber therein, heat- exchange tubes connecting these chambers, a solution of U/sup 233/ in heavy water in a reaction container within the tank, a slurry of thorium dioxide in heavy water in a second container surrounding the first container, an inlet conduit including a pump connecting the low pressure chamber to the reaction container, an outlet conduit connecting the high pressure chamber to the reaction container, and means of removing gaseous fission products released in both chambers. (AEC)

FISSILE MATERIAL AND FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

Shaner, B.E.

1961-08-15

The fissile material consists of about 64 to 70% (weight) zirconium dioxide, 15 to 19% uranium dioxide, and 8 to 17% calcium oxide. The fissile material is formed into sintered composites which are disposed in a compartmented fuel element, comprising essentially a flat filler plate having a plurality of compartments therein, enclosed in cladding plates of the same material as the filler plate. The resultant fuel has good resistance to corrosion in high temperature pressurized water, good dimensional stability to elevated temperatures, and good resistance to thermal shock. (AEC)

COMPOSITE CONTROL ROD

DOEpatents

Rock, H.R.

1963-12-24

A composite control rod for use in controlling a nuclear reactor is described. The control rod is of sandwich construction in which finned dowel pins are utilized to hold together sheets of the neutron absorbing material and nonabsorbing structural material thereby eliminating the need for being dependent on the absorbing material for structural support. The dowel pins perform the function of absorbing the forces due to differential thermal expansion, seating further with the fins into the sheets of material and crushing before damage is done either to the absorbing or non-absorbing material. (AEC)

Process for Descaling and Decontaminating Metals

DOEpatents

Baybarz, R. D.

1961-04-25

The oxide scale on the surface of stainless steels and similar metals is removed by contacting the metal under an inert atmosphere with a dilute H/sub 2/ SO/sub 4/ solution containing CrSO/sub 4/. The removed oxide scale is either dissolved or disintegrated into a slurry by the solution. Preferred reagent concentrations are 0.3 to 0.5 M CrSO/sub 4/ and 0.5 to 0.6 M H/sub 2/SO/sub 4/. The process is particularly applicable to decontamination of aqueous homogeneous nuclear reactor systems. (AEC)

NEUTRONIC REACTOR OPERATIONAL METHOD AND CORE SYSTEM

DOEpatents

Winters, C.E.; Graham, C.B.; Culver, J.S.; Wilson, R.H.

1960-07-19

Homogeneous neutronic reactor systems are described wherein an aqueous fuel solution is continuously circulated through a spherical core tank. The pumped fuel solution-is injected tangentially into the hollow spherical interior, thereby maintaining vigorous rotation of the solution within the tank in the form of a vortex; gaseous radiolytic decomposition products concentrate within the axial vortex cavity. The evolved gas is continuously discharged through a gas- outlet port registering with an extremity of the vortex cavity. and the solution stream is discharged through an annular liquid outlet port concentrically encircling the gas outlet by virtue of which the vortex and its cavity are maintained precisely axially aligned with the gas outlet. A primary heat exchanger extracts useful heat from the hot effluent fuel solution before its recirculation into the core tank. Hollow cylinders and other alternative core- tank configurations defining geometric volumes of revolution about a principal axis are also covered. AEC's Homogeneous Reactor Experiment No. 1 is a preferred embodiment.

Research on stellarator-mirror fission-fusion hybrid

NASA Astrophysics Data System (ADS)

Moiseenko, V. E.; Kotenko, V. G.; Chernitskiy, S. V.; Nemov, V. V.; Ågren, O.; Noack, K.; Kalyuzhnyi, V. N.; Hagnestål, A.; Källne, J.; Voitsenya, V. S.; Garkusha, I. E.

2014-09-01

The development of a stellarator-mirror fission-fusion hybrid concept is reviewed. The hybrid comprises of a fusion neutron source and a powerful sub-critical fast fission reactor core. The aim is the transmutation of spent nuclear fuel and safe fission energy production. In its fusion part, neutrons are generated in deuterium-tritium (D-T) plasma, confined magnetically in a stellarator-type system with an embedded magnetic mirror. Based on kinetic calculations, the energy balance for such a system is analyzed. Neutron calculations have been performed with the MCNPX code, and the principal design of the reactor part is developed. Neutron outflux at different outer parts of the reactor is calculated. Numerical simulations have been performed on the structure of a magnetic field in a model of the stellarator-mirror device, and that is achieved by switching off one or two coils of toroidal field in the Uragan-2M torsatron. The calculations predict the existence of closed magnetic surfaces under certain conditions. The confinement of fast particles in such a magnetic trap is analyzed.

Assessment of Sensor Technologies for Advanced Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Korsah, Kofi; Kisner, R. A.; Britton Jr., C. L.

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope formore » fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.« less

Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

NASA Astrophysics Data System (ADS)

Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

2017-07-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

41 CFR 50-204.34 - AEC licensees-AEC contractors operating AEC plants and facilities-AEC agreement State licensees...

Code of Federal Regulations, 2014 CFR

2014-07-01

... material, or special nuclear material, as defined in the Atomic Energy Act of 1954, as amended, under a license issued by the Atomic Energy Commission and in accordance with the requirements of 10 CFR part 20... contract with the Atomic Energy Commission for the operation of AEC plants and facilities and in accordance...

41 CFR 50-204.34 - AEC licensees-AEC contractors operating AEC plants and facilities-AEC agreement State licensees...

Code of Federal Regulations, 2013 CFR

2013-07-01

... material, or special nuclear material, as defined in the Atomic Energy Act of 1954, as amended, under a license issued by the Atomic Energy Commission and in accordance with the requirements of 10 CFR part 20... contract with the Atomic Energy Commission for the operation of AEC plants and facilities and in accordance...

41 CFR 50-204.34 - AEC licensees-AEC contractors operating AEC plants and facilities-AEC agreement State licensees...

Code of Federal Regulations, 2010 CFR

2010-07-01

... Alabama, Arkansas, California, Kansas, Kentucky, Florida, Mississippi, New Hampshire, New York, North..., Florida, Mississippi, New Hampshire, New York, North Carolina, Texas, Tennessee, Oregon, Idaho, Arizona...

41 CFR 50-204.34 - AEC licensees-AEC contractors operating AEC plants and facilities-AEC agreement State licensees...

Code of Federal Regulations, 2011 CFR

2011-07-01

... Alabama, Arkansas, California, Kansas, Kentucky, Florida, Mississippi, New Hampshire, New York, North..., Florida, Mississippi, New Hampshire, New York, North Carolina, Texas, Tennessee, Oregon, Idaho, Arizona...

Neutron fluxes in test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sienicki, James J.; Grandy, Christopher

A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. Themore » various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.« less

A CFD model for biomass fast pyrolysis in fluidized-bed reactors

NASA Astrophysics Data System (ADS)

Xue, Qingluan; Heindel, T. J.; Fox, R. O.

2010-11-01

A numerical study is conducted to evaluate the performance and optimal operating conditions of fluidized-bed reactors for fast pyrolysis of biomass to bio-oil. A comprehensive CFD model, coupling a pyrolysis kinetic model with a detailed hydrodynamics model, is developed. A lumped kinetic model is applied to describe the pyrolysis of biomass particles. Variable particle porosity is used to account for the evolution of particle physical properties. The kinetic scheme includes primary decomposition and secondary cracking of tar. Biomass is composed of reference components: cellulose, hemicellulose, and lignin. Products are categorized into groups: gaseous, tar vapor, and solid char. The particle kinetic processes and their interaction with the reactive gas phase are modeled with a multi-fluid model derived from the kinetic theory of granular flow. The gas, sand and biomass constitute three continuum phases coupled by the interphase source terms. The model is applied to investigate the effect of operating conditions on the tar yield in a fluidized-bed reactor. The influence of various parameters on tar yield, including operating temperature and others are investigated. Predicted optimal conditions for tar yield and scale-up of the reactor are discussed.

Deep-Earth reactor: Nuclear fission, helium, and the geomagnetic field

PubMed Central

Hollenbach, D. F.; Herndon, J. M.

2001-01-01

Geomagnetic field reversals and changes in intensity are understandable from an energy standpoint as natural consequences of intermittent and/or variable nuclear fission chain reactions deep within the Earth. Moreover, deep-Earth production of helium, having 3He/4He ratios within the range observed from deep-mantle sources, is demonstrated to be a consequence of nuclear fission. Numerical simulations of a planetary-scale geo-reactor were made by using the SCALE sequence of codes. The results clearly demonstrate that such a geo-reactor (i) would function as a fast-neutron fuel breeder reactor; (ii) could, under appropriate conditions, operate over the entire period of geologic time; and (iii) would function in such a manner as to yield variable and/or intermittent output power. PMID:11562483

Nuclear energy in the service of biomedicine: the U.S. Atomic Energy Commission's radioisotope program, 1946-1950.

PubMed

Creager, Angela N H

2006-01-01

The widespread adoption of radioisotopes as tools in biomedical research and therapy became one of the major consequences of the "physicists' war" for postwar life science. Scientists in the Manhattan Project, as part of their efforts to advocate for civilian uses of atomic energy after the war, proposed using infrastructure from the wartime bomb project to develop a government-run radioisotope distribution program. After the Atomic Energy Bill was passed and before the Atomic Energy Commission (AEC) was formally established, the Manhattan Project began shipping isotopes from Oak Ridge. Scientists and physicians put these reactor-produced isotopes to many of the same uses that had been pioneered with cyclotron-generated radioisotopes in the 1930s and early 1940s. The majority of early AEC shipments were radioiodine and radiophosphorus, employed to evaluate thyroid function, diagnose medical disorders, and irradiate tumors. Both researchers and politicians lauded radioisotopes publicly for their potential in curing diseases, particularly cancer. However, isotopes proved less successful than anticipated in treating cancer and more successful in medical diagnostics. On the research side, reactor-generated radioisotopes equipped biologists with new tools to trace molecular transformations from metabolic pathways to ecosystems. The U.S. government's production and promotion of isotopes stimulated their consumption by scientists and physicians (both domestic and abroad), such that in the postwar period isotopes became routine elements of laboratory and clinical use. In the early postwar years, radioisotopes signified the government's commitment to harness the atom for peace, particularly through contributions to biology, medicine, and agriculture.

Application of a Self-Actuating Shutdown System (SASS) to a Gas-Cooled Fast Reactor (GCFR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Germer, J.H.; Peterson, L.F.; Kluck, A.L.

1980-09-01

The application of a SASS (Self-Actuated Shutdown System) to a GCFR (Gas-Cooled Fast Reactor) is compared with similar systems designed for an LMFBR (Liquid Metal Fast Breeder Reactor). A comparison of three basic SASS concepts is given: hydrostatic holdup, fluidic control, and magnetic holdup.

Development of Inspection and Repair Technology for Heat Exchanger Tubes in Fast Breeder Reactors

DTIC Science & Technology

2009-06-01

Technology for Heat Exchanger Tubes in Fast Breeder Reactors Akihiko NISHIMURA *1 , Takahisa SHOBU, Kiyoshi OKA, Toshihiko YAMAGUCHI, Yukihiro SHIMADA...fast breeder reactors (FBRs). It comprises a laser processing head combined with an eddy current testing unit. Ultrashort laser pulse ablation is used...be applied in the main- tenance of large structures such as nuclear reactors and chemical factories [1]. Internal access to a blanket cooling pipe

Formerly utilized MED/AEC sites remedial action program. Radiological survey of The George Herbert Jones Chemical Laboratory, The University of Chicago, Chicago, Illinois, June 13-17, 1977

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wynveen, R.A.; Smith, W.H.; Mayes, C.B.

A comprehensive radiological survey was conducted at George Herbert Jones Chemical Laboratory at the University of Chicago, Chicago, Illinois. Radiochemistry for the MED/AEC project was performed in this building in the 1940s. The building is now used as laboratories, offices, and classrooms. The survey was undertaken to determine the location and quantities of any radioactive materials remaining from the MED/AEC operations. Forty-three spots of contamination possibly resulting from MED/AEC occupancy in 17 rooms exceeded the allowable limits as given in the ANSI Standard N13.12. Under current use conditions, the potential for radiation exposure to occupants of this building from thesemore » sources of contamination is remote. Concentrations of radon daughters in the air of the building, as measured with grab-sampling techniques, were below the limit of 0.01 WL above background as given in the Surgeon General's Guidelines. No long-lived radionuclides were detected in any air sample. Concentrations of radionuclides in soil samples from near the laboratory generally indicated background levels. In order to reduce the potential for radiation exposure, remedial measures such as stabilization of the contamination in place would be applicable as a short-term measure. In order to reduce the risk in the event that building modifications take place in the future, health physics procedures and coverage are recommended. The long-term solution would involve decontamination by removal of the radioactive residues from the 17 rooms or areas where contamination possibly resulting from MED/AEC activities was detected.« less

Hybrid systems for transuranic waste transmutation in nuclear power reactors: state of the art and future prospects

NASA Astrophysics Data System (ADS)

Yurov, D. V.; Prikhod'ko, V. V.

2014-11-01

The features of subcritical hybrid systems (HSs) are discussed in the context of burning up transuranic wastes from the U-Pu nuclear fuel cycle. The advantages of HSs over conventional atomic reactors are considered, and fuel cycle closure alternatives using HSs and fast neutron reactors are comparatively evaluated. The advantages and disadvantages of two HS types with neutron sources (NSs) of widely different natures -- nuclear spallation in a heavy target by protons and nuclear fusion in magnetically confined plasma -- are discussed in detail. The strengths and weaknesses of HSs are examined, and demand for them for closing the U-Pu nuclear fuel cycle is assessed.

Results of the International Energy Agency Round Robin on Fast Pyrolysis Bio-oil Production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Elliott, Douglas C.; Meier, Dietrich; Oasmaa, Anja

An international round robin study of the production of fast pyrolysis bio-oil was undertaken. Fifteen institutions in six countries contributed. Three biomass samples were distributed to the laboratories for processing in fast pyrolysis reactors. Samples of the bio-oil produced were transported to a central analytical laboratory for analysis. The round robin was focused on validating the pyrolysis community understanding of production of fast pyrolysis bio-oil by providing a common feedstock for bio-oil preparation. The round robin included: •distribution of 3 feedstock samples from a common source to each participating laboratory; •preparation of fast pyrolysis bio-oil in each laboratory with themore » 3 feedstocks provided; •return of the 3 bio-oil products (minimum 500 ml) with operational description to a central analytical laboratory for bio-oil property determination. The analyses of interest were: density, viscosity, dissolved water, filterable solids, CHN, S, trace element analysis, ash, total acid number, pyrolytic lignin, and accelerated aging of bio-oil. In addition, an effort was made to compare the bio-oil components to the products of analytical pyrolysis through GC/MS analysis. The results showed that clear differences can occur in fast pyrolysis bio-oil properties by applying different reactor technologies or configurations. The comparison to analytical pyrolysis method suggested that Py-GC/MS could serve as a rapid screening method for bio-oil composition when produced in fluid-bed reactors. Furthermore, hot vapor filtration generally resulted in the most favorable bio-oil product, with respect to water, solids, viscosity, and total acid number. These results can be helpful in understanding the variation in bio-oil production methods and their effects on bio-oil product composition.« less

Nuclear Data Needs for Generation IV Nuclear Energy Systems

NASA Astrophysics Data System (ADS)

Rullhusen, Peter

2006-04-01

Nuclear data needs for generation IV systems. Future of nuclear energy and the role of nuclear data / P. Finck. Nuclear data needs for generation IV nuclear energy systems-summary of U.S. workshop / T. A. Taiwo, H. S. Khalil. Nuclear data needs for the assessment of gen. IV systems / G. Rimpault. Nuclear data needs for generation IV-lessons from benchmarks / S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn. Core design issues of the supercritical water fast reactor / M. Mori ... [et al.]. GFR core neutronics studies at CEA / J. C. Bosq ... [et al]. Comparative study on different phonon frequency spectra of graphite in GCR / Young-Sik Cho ... [et al.]. Innovative fuel types for minor actinides transmutation / D. Haas, A. Fernandez, J. Somers. The importance of nuclear data in modeling and designing generation IV fast reactors / K. D. Weaver. The GIF and Mexico-"everything is possible" / C. Arrenondo Sánchez -- Benmarks, sensitivity calculations, uncertainties. Sensitivity of advanced reactor and fuel cycle performance parameters to nuclear data uncertainties / G. Aliberti ... [et al.]. Sensitivity and uncertainty study for thermal molten salt reactors / A. Biduad ... [et al.]. Integral reactor physics benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) / J. B. Briggs, D. W. Nigg, E. Sartori. Computer model of an error propagation through micro-campaign of fast neutron gas cooled nuclear reactor / E. Ivanov. Combining differential and integral experiments on [symbol] for reducing uncertainties in nuclear data applications / T. Kawano ... [et al.]. Sensitivity of activation cross sections of the Hafnium, Tanatalum and Tungsten stable isotopes to nuclear reaction mechanisms / V. Avrigeanu ... [et al.]. Generating covariance data with nuclear models / A. J. Koning. Sensitivity of Candu-SCWR reactors physics calculations to nuclear data files / K. S. Kozier, G. R. Dyck. The lead cooled fast reactor benchmark BREST-300: analysis with sensitivity method / V. Smirnov ... [et al.]. Sensitivity analysis of neutron cross-sections considered for design and safety studies of LFR and SFR generation IV systems / K. Tucek, J. Carlsson, H. Wider -- Experiments. INL capabilities for nuclear data measurements using the Argonne intense pulsed neutron source facility / J. D. Cole ... [et al.]. Cross-section measurements in the fast neutron energy range / A. Plompen. Recent measurements of neutron capture cross sections for minor actinides by a JNC and Kyoto University Group / H. Harada ... [et al.]. Determination of minor actinides fission cross sections by means of transfer reactions / M. Aiche ... [et al.] -- Evaluated data libraries. Nuclear data services from the NEA / H. Henriksson, Y. Rugama. Nuclear databases for energy applications: an IAEA perspective / R. Capote Noy, A. L. Nichols, A. Trkov. Nuclear data evaluation for generation IV / G. Noguère ... [et al.]. Improved evaluations of neutron-induced reactions on americium isotopes / P. Talou ... [et al.]. Using improved ENDF-based nuclear data for candu reactor calculations / J. Prodea. A comparative study on the graphite-moderated reactors using different evaluated nuclear data / Do Heon Kim ... [et al.].

Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jensen, C.; Wachs, D.; Carmack, J.

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, andmore » salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.« less

FLOW GATING

DOEpatents

Poppelbaum, W.J.

1962-12-01

BS>This invention is a fast gating system for eiectronic flipflop circuits. Diodes connect the output of one circuit to the input of another, and the voltage supply for the receiving flip-flop has two alternate levels. When the supply is at its upper level, no current can flow through the diodes, but when the supply is at its lower level, current can flow to set the receiving flip- flop to the same state as that of the circuit to which it is connected. (AEC)

FAST OPENING SWITCH

DOEpatents

Bender, M.; Bennett, F.K.; Kuckes, A.F.

1963-09-17

A fast-acting electric switch is described for rapidly opening a circuit carrying large amounts of electrical power. A thin, conducting foil bridges a gap in this circuit and means are provided for producing a magnetic field and eddy currents in the foil, whereby the foil is rapidly broken to open the circuit across the gap. Advantageously the foil has a hole forming two narrow portions in the foil and the means producing the magnetic field and eddy currents comprises an annular coil having its annulus coaxial with the hole in the foil and turns adjacent the narrow portions of the foil. An electrical current flows through the coil to produce the magnetic field and eddy currents in the foil. (AEC)

The effective dose result of 18F-FDG PET-CT paediatric patients

NASA Astrophysics Data System (ADS)

Hussin, D.; Said, M. A.; Ali, N. S.; Tajuddin, A. A.; Zainon, R.

2017-05-01

Paediatric patient received high exposure from both CT and PET examination. Automatic Exposure Control (AEC) is important in CT dose reduction. This study aimed to compare the effective dose obtained from PET-CT scanner with and without the use of AEC function. In this study, 68 patients underwent PET-CT examination without the use of AEC function, while 25 patients used the AEC function during the examination. Patients involved in this study were between 2 to 15 years old with varies of malignancies and epilepsy diseases. The effective dose obtained from PET and CT examinations was calculated based on recommendation from International Commission on Radiological Protection (ICRP) Publication 106 and ICRP publication 102. The outcome of this study shows that the radiation dose was reduced up to 20% with the use of AEC function. The mean average of effective dose result obtained from PET and CT examinations without the use of AEC and AEC function were found to be as 6.67 mSv, 6.77 mSv, 6.03mSv and 4.96 mSv respectively. Where total effective dose result of PET-CT with non-AEC and AEC were found to be 13.44 mSv and 10.99 mSv respectively. Conclusion of this study is, the installation of AEC function in PET-CT machine does play important role in CT dose reduction especially for paediatric patient.

Some Implications of the Three Mile Island Accident for LMFBR Safety and Licensing: The Design Basis Issue

DTIC Science & Technology

1980-08-01

metal fast breeder reactor (LMFBR) design. It also re-examines the impact of the accident at Three Mile Island on the design basis concept, and how...Water Reactors : ImpZications for Liquid MetaZ Fast Breeder Reactors , by W. E. Kastenberg and K. A. Solomon, July 1979. v SUNMARY The 1979 accident...the liquid metal fast breeder reactor (LMFBR). This Note assesses the impact of the TMI-2 accident on the LMFBR. Specifically, it: o Reviews the

FUEL ELEMENT INTERLOCKING ARRANGEMENT

DOEpatents

Fortescue, P.; Nicoll, D.

1963-01-01

This patent relates to a system for mutually interlocking a multiplicity of elongated, parallel, coextensive, upright reactor fuel elements so as to render a laterally selfsupporting bundle, while admitting of concurrent, selective, vertical withdrawal of a sizeable number of elements without any of the remaining elements toppling, Each element is provided with a generally rectangular end cap. When a rank of caps is aligned in square contact, each free edge centrally defines an outwardly profecting dovetail, and extremitally cooperates with its adjacent cap by defining a juxtaposed half of a dovetail- receptive mortise. Successive ranks are staggered to afford mating of their dovetails and mortises. (AEC)

SAFETY SYSTEM FOR CONTROL ROD

DOEpatents

Paget, J.A.

1963-05-14

A structure for monitoring the structural continuity of a control rod foi a neutron reactor is presented. A electric conductor readily breakable under mechanical stress is fastened along the length of the control rod at a plurality of positions and forms a closed circuit with remote electrical components responsive to an open circuit. A portion of the conductor between the control rod and said components is helically wound to allow free and normally unrestricted movement of the segment of conductor secured to the control rod relative to the remote components. Any break in the circuit is indicative of control rod breakage. (AEC)

Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mamivand, Mahmood; Yang, Ying; Busby, Jeremy T.

The current work combines the Cluster Dynamics (CD) technique and CALPHAD-based precipitation modeling to address the second phase precipitation in cold-worked (CW) 316 stainless steels (SS) under irradiation at 300–400 °C. CD provides the radiation enhanced diffusion and dislocation evolution as inputs for the precipitation model. The CALPHAD-based precipitation model treats the nucleation, growth and coarsening of precipitation processes based on classical nucleation theory and evolution equations, and simulates the composition, size and size distribution of precipitate phases. We benchmark the model against available experimental data at fast reactor conditions (9.4 × 10 –7 dpa/s and 390 °C) and thenmore » use the model to predict the phase instability of CW 316 SS under light water reactor (LWR) extended life conditions (7 × 10 –8 dpa/s and 275 °C). The model accurately predicts the γ' (Ni 3Si) precipitation evolution under fast reactor conditions and that the formation of this phase is dominated by radiation enhanced segregation. The model also predicts a carbide volume fraction that agrees well with available experimental data from a PWR reactor but is much higher than the volume fraction observed in fast reactors. We propose that radiation enhanced dissolution and/or carbon depletion at sinks that occurs at high flux could be the main sources of this inconsistency. The integrated model predicts ~1.2% volume fraction for carbide and ~3.0% volume fraction for γ' for typical CW 316 SS (with 0.054 wt% carbon) under LWR extended life conditions. Finally, this work provides valuable insights into the magnitudes and mechanisms of precipitation in irradiated CW 316 SS for nuclear applications.« less

Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

DOE PAGES

Mamivand, Mahmood; Yang, Ying; Busby, Jeremy T.; ...

2017-03-11

The current work combines the Cluster Dynamics (CD) technique and CALPHAD-based precipitation modeling to address the second phase precipitation in cold-worked (CW) 316 stainless steels (SS) under irradiation at 300–400 °C. CD provides the radiation enhanced diffusion and dislocation evolution as inputs for the precipitation model. The CALPHAD-based precipitation model treats the nucleation, growth and coarsening of precipitation processes based on classical nucleation theory and evolution equations, and simulates the composition, size and size distribution of precipitate phases. We benchmark the model against available experimental data at fast reactor conditions (9.4 × 10 –7 dpa/s and 390 °C) and thenmore » use the model to predict the phase instability of CW 316 SS under light water reactor (LWR) extended life conditions (7 × 10 –8 dpa/s and 275 °C). The model accurately predicts the γ' (Ni 3Si) precipitation evolution under fast reactor conditions and that the formation of this phase is dominated by radiation enhanced segregation. The model also predicts a carbide volume fraction that agrees well with available experimental data from a PWR reactor but is much higher than the volume fraction observed in fast reactors. We propose that radiation enhanced dissolution and/or carbon depletion at sinks that occurs at high flux could be the main sources of this inconsistency. The integrated model predicts ~1.2% volume fraction for carbide and ~3.0% volume fraction for γ' for typical CW 316 SS (with 0.054 wt% carbon) under LWR extended life conditions. Finally, this work provides valuable insights into the magnitudes and mechanisms of precipitation in irradiated CW 316 SS for nuclear applications.« less

Californium-252: a remarkable versatile radioisotope

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne-Lee, I.W.; Alexander, C.W.

A product of the nuclear age, Californium-252 ({sup 252}Cf) has found many applications in medicine, scientific research, industry, and nuclear science education. Californium-252 is unique as a neutron source in that it provides a highly concentrated flux and extremely reliable neutron spectrum from a very small assembly. During the past 40 years, {sup 252}Cf has been applied with great success to cancer therapy, neutron radiography of objects ranging from flowers to entire aircraft, startup sources for nuclear reactors, fission activation for quality analysis of all commercial nuclear fuel, and many other beneficial uses, some of which are now ready formore » further growth. Californium-252 is produced in the High Flux Isotope Reactor (HFIR) and processed in the Radiochemical Engineering Development Center (REDC), both of which are located at the Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee. The REDC/HFIR facility is virtually the sole supplier of {sup 252}Cf in the western world and is the major supplier worldwide. Extensive exploitation of this product was made possible through the {sup 252}Cf Market Evaluation Program, sponsored by the United States Department of Energy (DOE) [then the Atomic Energy Commission (AEC) and later the Energy Research and Development Administration (ERDA)]. This program included training series, demonstration centers, seminars, and a liberal loan policy for fabricated sources. The Market Evaluation Program was instituted, in part, to determine if large-quantity production capability was required at the Savannah River Laboratory (SRL). Because of the nature of the product and the means by which it is produced, {sup 252}Cf can be produced only in government-owned facilities. It is evident at this time that the Oak Ridge research facility can meet present and projected near-term requirements. The production, shipment, and sales history of {sup 252}Cf from ORNL is summarized herein.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Craig, D.F.

The division was formed in 1946 at the suggestion of Dr. Eugene P. Wigner to attack the problem of the distortion of graphite in the early reactors due to exposure to reactor neutrons, and the consequent radiation damage. It was called the Metallurgy Division and assembled the metallurgical and solid state physics activities of the time which were not directly related to nuclear weapons production. William A. Johnson, a Westinghouse employee, was named Division Director in 1946. In 1949 he was replaced by John H Frye Jr. when the Division consisted of 45 people. He was director during most ofmore » what is called the Reactor Project Years until 1973 and his retirement. During this period the Division evolved into three organizational areas: basic research, applied research in nuclear reactor materials, and reactor programs directly related to a specific reactor(s) being designed or built. The Division (Metals and Ceramics) consisted of 204 staff members in 1973 when James R. Weir, Jr., became Director. This was the period of the oil embargo, the formation of the Energy Research and Development Administration (ERDA) by combining the Atomic Energy Commission (AEC) with the Office of Coal Research, and subsequent formation of the Department of Energy (DOE). The diversification process continued when James O. Stiegler became Director in 1984, partially as a result of the pressure of legislation encouraging the national laboratories to work with U.S. industries on their problems. During that time the Division staff grew from 265 to 330. Douglas F. Craig became Director in 1992.« less

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-11-03

separation list: ! 8 indigenous Indian power reactors ! Fast Breeder test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction...facilities like reprocessing and enrichment plants and breeder reactors could be viewed as providing a significant nonproliferation benefit because the... breeder reactors would support the 2002 U.S. National Strategy to Combat Weapons of Mass Destruction, in which the United States pledged to “continue to

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-10-02

8 indigenous Indian power reactors ! Fast Breeder test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction ! Enrichment... breeder reactors could be viewed as providing a significant nonproliferation benefit because the materials produced by these plants are a few steps closer...to potential use in a bomb. In addition, safeguards on enrichment, reprocessing plants, and breeder reactors would support the 2002 U.S. National

NUCLEAR SUPERHEATER FOR BOILING WATER REACTOR

DOEpatents

Holl, R.J.; Klecker, R.W.; Graham, C.B.

1962-05-15

A description is given of a boiling water reactor having a superheating region integral with the core. The core consists essentially of an annular boiling region surrounding an inner superheating region. Both regions contain fuel elements and are separated by a cylindrical wall, perforations being provided in the lower portion of the cylindrical wall to permit circulation of a common water moderator between the two regions. The superheater region comprises a plurality of tubular fuel assemblies through which the steam emanating from the boiling region passes to the steam outlet. Each superheater fuel assembly has an outer double-walled cylinder, the double walls being concentrically spaced and connected together at their upper ends but open at the bottom to provide for differential thermal expansion of the inner and outer walls. Gas is entrapped in the annulus between the walls which acts as an insulating space between the fissionable material inside and the moderator outside. (AEC)

Eddy Current Flow Measurements in the FFTF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nielsen, Deborah L.; Polzin, David L.; Omberg, Ronald P.

2017-02-02

The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following shut down of the Clinch River Breeder Reactor Plant (CRBRP) project in 1983, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuablemore » information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor control and operations. This report provides HEDL-TC-1344, “ECFM Flow Measurements in the FFTF Using Phase-Sensitive Detectors”, March 1979.« less

Overview of past and current activities on fuels for fast reactors at the Institute for Transuranium Elements

NASA Astrophysics Data System (ADS)

Fernandez, A.; McGinley, J.; Somers, J.; Walter, M.

2009-07-01

Nuclear energy has the potential to provide a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gas emissions. The renewal of interest in fast neutron spectra reactors to meet more ambitious sustainable development criteria (i.e., resource maximisation and waste minimisation), opens a favourable framework for R&D activities in this area. The Institute for Transuranium Elements has extensive experience in the fabrication, characterization and irradiation testing (Phénix, Dounreay, Rapsodie) of fast reactor fuels, in oxide, nitride and carbide forms. An overview of these past and current activities on fast reactor fuels is presented.

Benchmark tests of JENDL-3.2 for thermal and fast reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takano, Hideki; Akie, Hiroshi; Kikuchi, Yasuyuki

1994-12-31

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k{sub eff} and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k{sub eff} reactivity worths of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments.

Correction method for influence of tissue scattering for sidestream dark-field oximetry using multicolor LEDs

NASA Astrophysics Data System (ADS)

Kurata, Tomohiro; Oda, Shigeto; Kawahira, Hiroshi; Haneishi, Hideaki

2016-12-01

We have previously proposed an estimation method of intravascular oxygen saturation (SO_2) from the images obtained by sidestream dark-field (SDF) imaging (we call it SDF oximetry) and we investigated its fundamental characteristics by Monte Carlo simulation. In this paper, we propose a correction method for scattering by the tissue and performed experiments with turbid phantoms as well as Monte Carlo simulation experiments to investigate the influence of the tissue scattering in the SDF imaging. In the estimation method, we used modified extinction coefficients of hemoglobin called average extinction coefficients (AECs) to correct the influence from the bandwidth of the illumination sources, the imaging camera characteristics, and the tissue scattering. We estimate the scattering coefficient of the tissue from the maximum slope of pixel value profile along a line perpendicular to the blood vessel running direction in an SDF image and correct AECs using the scattering coefficient. To evaluate the proposed method, we developed a trial SDF probe to obtain three-band images by switching multicolor light-emitting diodes and obtained the image of turbid phantoms comprised of agar powder, fat emulsion, and bovine blood-filled glass tubes. As a result, we found that the increase of scattering by the phantom body brought about the decrease of the AECs. The experimental results showed that the use of suitable values for AECs led to more accurate SO_2 estimation. We also confirmed the validity of the proposed correction method to improve the accuracy of the SO_2 estimation.

Nuclear modules for space electric propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1998-01-01

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takeda, T.; Shimazu, Y.; Hibi, K.

2012-07-01

Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of thismore » project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)« less

Simulator platform for fast reactor operation and safety technology demonstration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vilim, R. B.; Park, Y. S.; Grandy, C.

2012-07-30

A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe responsemore » to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.« less

Anomalies in the Charge Yields of Fission Fragments from the U ( n , f ) 238 Reaction

DOE PAGES

Wilson, J. N.; Lebois, M.; Qi, L.; ...

2017-06-01

Fast-neutron-induced fission of 238U at an energy just above the fission threshold is studied with a novel technique which involves the coupling of a high-efficiency γ-ray spectrometer (MINIBALL) to an inverse-kinematics neutron source (LICORNE) to extract charge yields of fission fragments via γ-γ coincidence spectroscopy. Experimental data and fission models are compared and found to be in reasonable agreement for many nuclei; however, significant discrepancies of up to 600% are observed, particularly for isotopes of Sn and Mo. This indicates that these models significantly overestimate the standard 1 fission mode and suggests that spherical shell effects in the nascent fissionmore » fragments are less important for low-energy fast-neutron-induced fission than for thermal neutron-induced fission. Finally, this has consequences for understanding and modeling the fission process, for experimental nuclear structure studies of the most neutron-rich nuclei, for future energy applications (e.g., Generation IV reactors which use fast-neutron spectra), and for the reactor antineutrino anomaly.« less

PRECISION INTEGRATOR FOR MINUTE ELECTRIC CURRENTS

DOEpatents

Hemmendinger, A.; Helmer, R.J.

1961-10-24

An integrator is described for measuring the value of integrated minute electrical currents. The device consists of a source capacitor connected in series with the source of such electrical currents, a second capacitor of accurately known capacitance and a source of accurately known and constant potential, means responsive to the potentials developed across the source capacitor for reversibly connecting the second capacitor in series with the source of known potential and with the source capacitor and at a rate proportional to the potential across the source capacitor to maintain the magnitude of the potential across the source capacitor at approximately zero. (AEC)

Conditioned medium derived from rat amniotic epithelial cells confers protection against inflammation, cancer, and senescence.

PubMed

Di Germanio, Clara; Bernier, Michel; Petr, Michael; Mattioli, Mauro; Barboni, Barbara; de Cabo, Rafael

2016-06-28

Amniotic epithelial cells (AECs) are a class of fetal stem cells that derives from the epiblast and resides in the amnion until birth. AECs are suitable candidates for regenerative medicine because of the ease of collection, their low immunogenicity and inability to form tumors after transplantation. Even though human AECs have been widely investigated, the fact remains that very little is known about AECs isolated from rat, one of the most common animal models in medical testing. In this study, we showed that rat AECs retained stemness properties and plasticity, expressed the pluripotency markers Sox2, Nanog, and Oct4 and were able to differentiate toward the osteogenic lineage. The addition of conditioned medium collected from rat AECs to lipopolysaccharide-activated macrophages elicited anti-inflammatory properties through a decrease of Tnfa expression and slowed tumor cell proliferation in vitro and in vivo. The senescence-associated secretory phenotype was also significantly lower upon incubation of senescent human IMR-90 fibroblast cells with conditioned medium from rat AECs. These results confirm the potential of AECs in the modulation of inflammatory mechanisms and open new therapeutic possibilities for regenerative medicine and anti-aging therapies as well.

Design and implementation of a hybrid sub-band acoustic echo canceller (AEC)

NASA Astrophysics Data System (ADS)

Bai, Mingsian R.; Yang, Cheng-Ken; Hur, Ker-Nan

2009-04-01

An efficient method is presented for implementing an acoustic echo canceller (AEC) that makes use of hybrid sub-band approach. The hybrid system is comprised of a fixed processor and an adaptive filter in each sub-band. The AEC aims at reducing the echo resulting from the acoustic feedback in loudspeaker-enclosure-microphone (LEM) systems such as teleconferencing and hands-free systems. In order to cancel the acoustical echo efficiently, various processing architectures including fixed filters, hybrid processors, and sub-band structure are investigated. A double-talk detector is incorporated into the proposed AEC to prevent the adaptive filter from diverging in double-talk situations. A de-correlation filter is also used alongside sub-band processing in order to enhance the performance and efficiency of AEC. All algorithms are implemented and verified on the platform of a fixed-point digital signal processor (DSP). The AECs are evaluated in terms of cancellation performance and computation complexity. In addition, listening tests are conducted to assess the subjective performance of the AECs. From the results, the proposed hybrid sub-band AEC was found to be the most effective among all methods in terms of echo reduction and timbral quality.

The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; Michael A. Pope; Harold F. McFarlane

2012-11-01

The creation of a large database of integral fast reactor physics experiments advanced nuclear science and technology in ways that were unachievable by less capital intensive and operationally challenging approaches. They enabled the compilation of integral physics benchmark data, validated (or not) analytical methods, and provided assurance of future rector designs The integral experiments performed at Argonne National Laboratory (ANL) represent decades of research performed to support fast reactor design and our understanding of neutronics behavior and reactor physics measurements. Experiments began in 1955 with the Zero Power Reactor No. 3 (ZPR-3) and terminated with the Zero Power Physics Reactormore » (ZPPR, originally the Zero Power Plutonium Reactor) in 1990 at the former ANL-West site in Idaho, which is now part of the Idaho National Laboratory (INL). Two additional critical assemblies, ZPR-6 and ZPR-9, operated at the ANL-East site in Illinois. A total of 128 fast reactor assemblies were constructed with these facilities [1]. The infrastructure and measurement capabilities are too expensive to be replicated in the modern era, making the integral database invaluable as the world pushes ahead with development of liquid metal cooled reactors.« less

Fuel supply of nuclear power industry with the introduction of fast reactors

NASA Astrophysics Data System (ADS)

Muraviev, E. V.

2014-12-01

The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

The prototype fast reactor at Dounreay, Scotland. Process and engineering development for sodium removal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mann, A.; Herrick, R.; Gunn, J.

2007-07-01

Dounreay was home to commercial fast reactor development in the UK. Following the construction and operation of the Dounreay Fast Reactor, a sodium-cooled Prototype Fast Reactor (PFR), was constructed. PFR started operating in 1974, closed in 1994 and is presently being decommissioned. To date the bulk of the sodium has been removed and treated. Due to the design of the existing extraction system however, a sodium pool will remain in the heel of the reactor. To remove this sodium, a pump/camera system was developed, tested and deployed. The Water Vapour Nitrogen (WVN) process has been selected to allow removal ofmore » the final sodium residues from the reactor. Due to the design of the reactor and potential for structural damage should Normal WVN (which produces hydrated sodium hydroxide) be used, Low Concentration WVN (LC WVN) has been developed. Pilot scale testing has shown that it is possible treat the reactor within 18 months at a WVN concentration of up to 4% v/v and temperature of 120 deg. C. At present the equipment that will be used to apply LC WVN to the reactor is being developed at the detail design stage. and is expected to be deployed within the next few years. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grabaskas, David; Bucknor, Matthew; Jerden, James

A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gapsmore » in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.« less

The benefits of a fast reactor closed fuel cycle in the UK

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gregg, R.; Hesketh, K.

2013-07-01

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size,more » so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)« less

Transient Effects in Turbulence Modelling.

DTIC Science & Technology

1979-12-01

plenum region of a liquid-metal- cooled fast breeder reactor (LMFBR). The efficient heat transfer characteristics of liquid metal coolant, combined...Transients in Generalized Liquid-Metal Fast Breeder Reactor Outlet Plenums," Nuclear Technology, Vol. 44, July 1979, p. 210. 135 15. Lorenz, J. J., "MIX... Sodium Coolant in the Outlet Plenum of a Fast Nuclear Reactor ," Int. J. Heat Mass Transfer, Vol. 21, 1978, pp. 1565-1579. 19. Chen, Y. B., Golay, M. W

Japan’s Nuclear Future: Policy Debate, Prospects, and U.S. Interests

DTIC Science & Technology

2008-05-09

raised in particular over the construction of an industrial- scale reprocessing facility in Japan,. Additionally, fast breeder reactors also produce more...Nuclear Fuel Cycle Engineering Laboratories. 10 A fast breeder reactor is a fast neutron reactor that produces more plutonium than it consumes, which can...Japan Nuclear Fuel Limited (JNFL) has built and is currently running active testing on a large - scale commercial reprocessing plant at Rokkasho-mura

Significance of breeding in fast nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Raza, S.M.; Abidi, S.B.M.

1983-12-01

Only breeder reactors--nuclear power plants that produce more fuel than they consume--are capable in principle of extracting the maximum amount of fission energy contained in uranium ore, thus offering a practical long-term solution to uranium supply problems. Uranium would then constitute a virtually inexhaustible fuel reserve for the world's future energy needs. The ultimate argument for breeding is to conserve the energy resources available to mankind. A long-term role for nuclear power with fast reactors is proven to be economically viable, environmentally acceptable and capable of wide scale exploitation in many countries. In this paper, various suggestions pertaining to themore » fuel fabrication route, fuel cycle economics, studies of the physics of fast nuclear reactors and of engineering design simplifications are presented. Fast reactors contain no moderator and inherently require enriched fuel. In general, the main aim is to suggest an improvement in the understanding of the safety and control characteristics of fast breeder power reactors. Development work is also being devoted to new carbide and nitride fuels, which are likely to exhibit breeding characteristics superior to those of the oxides of plutonium and uranium.« less

Thermomechanical analysis of fast-burst reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, J.D.

1994-08-01

Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor.

Enhanced rifampicin delivery to alveolar macrophages by solid lipid nanoparticles

NASA Astrophysics Data System (ADS)

Chuan, Junlan; Li, Yanzhen; Yang, Likai; Sun, Xun; Zhang, Qiang; Gong, Tao; Zhang, Zhirong

2013-05-01

The present study aimed at developing a drug delivery system targeting the densest site of tuberculosis infection, the alveolar macrophages (AMs). Rifampicin (RFP)-loaded solid lipid nanoparticles (RFP-SLNs) with an average size of 829.6 ± 16.1 nm were prepared by a modified lipid film hydration method. The cytotoxicity of RFP-SLNs to AMs and alveolar epithelial type II cells (AECs) was examined using MTT assays. The viability of AMs and AECs was above 80 % after treatment with RFP-SLNs, which showed low toxicity to both AMs and AECs. Confocal Laser Scanning Microscopy was employed to observe the interaction between RFP-SLNs and both AMs and AECs. After incubating the cells with RFP-SLNs for 2 h, the fluorescent intensity in AMs was more and remained longer (from 0.5 to 12 h) when compared with that in AECs (from 0.5 to 8 h). In vitro uptake characteristics of RFP-SLNs in AMs and AECs were also investigated by detection of intracellular RFP by High performance liquid chromatography. Results showed that RFP-SLNs delivered markedly higher RFP into AMs (691.7 ng/mg in cultured AMs, 662.6 ng/mg in primary AMs) than that into AECs (319.2 ng/mg in cultured AECs, 287.2 ng/mg in primary AECs). Subsequently, in vivo delivery efficiency and the selectivity of RFP-SLNs were further verified in Sprague-Dawley rats. Under pulmonary administration of RFP-SLNs, the amount of RFP in AMs was significantly higher than that in AECs at each time point. Our results demonstrated that solid lipid nanoparticles are a promising strategy for the delivery of rifampicin to alveolar macrophages selectively.

Expression pattern of salt tolerance-related genes in Aegilops cylindrica.

PubMed

Arabbeigi, Mahbube; Arzani, Ahmad; Majidi, Mohammad Mahdi; Sayed-Tabatabaei, Badraldin Ebrahim; Saha, Prasenjit

2018-02-01

Aegilops cylindrica , a salt-tolerant gene pool of wheat, is a useful plant model for understanding mechanism of salt tolerance. A salt-tolerant USL26 and a salt-sensitive K44 genotypes of A. cylindrica , originating from Uremia Salt Lake shores in Northwest Iran and a non-saline Kurdestan province in West Iran, respectively, were identified based on screening evaluation and used for this work. The objective of the current study was to investigate the expression patterns of four genes related to ion homeostasis in this species. Under treatment of 400 mM NaCl, USL26 showed significantly higher root and shoot dry matter levels and K + concentrations, together with lower Na + concentrations than K44 genotype. A. cylindrica HKT1;5 ( AecHKT1;5 ), SOS1 ( AecSOS1 ), NHX1 ( AecNHX1 ) and VP1 ( AecVP1 ) were partially sequenced to design each gene specific primer. Quantitative real-time PCR showed a differential expression pattern of these genes between the two genotypes and between the root and shoot tissues. Expressions of AecHKT1;5 and AecSOS1 was greater in the roots than in the shoots of USL26 while AecNHX1 and AecVP1 were equally expressed in both tissues of USL26 and K44. The higher transcripts of AecHKT1;5 in the roots versus the shoots could explain both the lower Na + in the shoots and the much lower Na + and higher K + concentrations in the roots/shoots of USL26 compared to K44. Therefore, the involvement of AecHKT1;5 in shoot-to-root handover of Na + in possible combination with the exclusion of excessive Na + from the root in the salt-tolerant genotype are suggested.

MOUND LABORATORY PROGRESS REPORT FOR DECEMBER 1960

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

Activities are reported in a program to investigate formulations and procedures which may lead to superior plastics and adhesives. In other work, processes for separating and purifying radioelements are being developed and supply sources are being evaluated. Research was initiated to determine the density, viscosity, thermal capacity, and thermal conductivity of Pu and Pu alloys for use in fast breeder reactors. (J.R.D.)

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

NASA Astrophysics Data System (ADS)

Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

Fast quench reactor and method

DOEpatents

Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

1998-05-12

A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-10-17

safeguards-irrelevant.” The following facilities and activities were not on the separation list: ! 8 indigenous Indian power reactors ! Fast Breeder ...test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction ! Enrichment facilities ! Spent fuel reprocessing facilities (except...potential use in a bomb. In addition, safeguards on enrichment, reprocessing plants, and breeder reactors would support the 2002 U.S. National Strategy to

LOGIC CIRCUIT

DOEpatents

Strong, G.H.; Faught, M.L.

1963-12-24

A device for safety rod counting in a nuclear reactor is described. A Wheatstone bridge circuit is adapted to prevent de-energizing the hopper coils of a ball backup system if safety rods, sufficient in total control effect, properly enter the reactor core to effect shut down. A plurality of resistances form one arm of the bridge, each resistance being associated with a particular safety rod and weighted in value according to the control effect of the particular safety rod. Switching means are used to switch each of the resistances in and out of the bridge circuit responsive to the presence of a particular safety rod in its effective position in the reactor core and responsive to the attainment of a predetermined velocity by a particular safety rod enroute to its effective position. The bridge is unbalanced in one direction during normal reactor operation prior to the generation of a scram signal and the switching means and resistances are adapted to unbalance the bridge in the opposite direction if the safety rods produce a predetermined amount of control effect in response to the scram signal. The bridge unbalance reversal is then utilized to prevent the actuation of the ball backup system, or, conversely, a failure of the safety rods to produce the predetermined effect produces no unbalance reversal and the ball backup system is actuated. (AEC)

FUEL ELEMENT

DOEpatents

Fortescue, P.; Zumwalt, L.R.

1961-11-28

A fuel element was developed for a gas cooled nuclear reactor. The element is constructed in the form of a compacted fuel slug including carbides of fissionable material in some cases with a breeder material carbide and a moderator which slug is disposed in a canning jacket of relatively impermeable moderator material. Such canned fuel slugs are disposed in an elongated shell of moderator having greater gas permeability than the canning material wherefore application of reduced pressure to the space therebetween causes gas diffusing through the exterior shell to sweep fission products from the system. Integral fission product traps and/or exterior traps as well as a fission product monitoring system may be employed therewith. (AEC)

METHOD OF CHEMICAL DECONTAMINATION OF STAINLESS STEEL NUCLEAR FACILITIES

DOEpatents

Pancer, G.P.; Zegger, J.L.

1961-12-19

A chemical method is given for removing activated corrosion products on the primary system surfaces of a pressurized water reactor. The corrosion product deposits are composed chiefly of magnetite (Fe/sub 3/O/sub 4/) with small amounts of nickel and chromium oxides. The corroded surfaces are first flushed with a caustic permanganate primary solution consisting of sodium hydroxide and potassium permanganate followed by a secondary rinse solution of ammonium citrate and citric acid containing the complexing agent Versene in small amounts. Demineralized water is used to clean out the primary and secondary solutions and a 60-minute drying period precedes the rinse solution. (AEC)

METHOD OF MAKING FUEL BODIES

DOEpatents

Goeddel, W.V.; Simnad, M.T.

1962-04-24

An improved method of making a fuel body containing carbon for reactors is described. Carbides of uranium and thorium having a particle size of from 100 to 500 microns are mixed with carbon having a particle size that will pass a 200 mesh screen but be retained by a 325 mesh screen, and 10 per cent by weight pitch. The mixture is heated to a temperature of about 700 to 900 deg C, at which point bonding is effected while maintaining it under mechanical pressure of over 3,000 pounds per square inch. The entire compact is heated to a uniform temperature during the process, preferably by electrical resistance of the compact itself. (AEC)

In situ TEM and synchrotron characterization of U–10Mo thin specimen annealed at the fast reactor temperature regime

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yun, Di, E-mail: diyun1979@xjtu.edu.cn; Xi'an Jiao Tong University, 28 Xian Ning West Road, Xi'an 710049; Mo, Kun

2015-12-15

U–Mo metallic alloys have been extensively used for the Reduced Enrichment for Research and Test Reactors (RERTR) program, which is now known as the Office of Material Management and Minimization under the Conversion Program. This fuel form has also recently been proposed as fast reactor metallic fuels in the recent DOE Ultra-high Burnup Fast Reactor project. In order to better understand the behavior of U–10Mo fuels within the fast reactor temperature regime, a series of annealing and characterization experiments have been performed. Annealing experiments were performed in situ at the Intermediate Voltage Electron Microscope (IVEM-Tandem) facility at Argonne National Laboratorymore » (ANL). An electro-polished U–10Mo alloy fuel specimen was annealed in situ up to 700 °C. At an elevated temperature of about 540 °C, the U–10Mo specimen underwent a relatively slow microstructure transition. Nano-sized grains were observed to emerge near the surface. At the end temperature of 700 °C, the near-surface microstructure had evolved to a nano-crystalline state. In order to clarify the nature of the observed microstructure, Laue diffraction and powder diffraction experiments were carried out at beam line 34-ID of the Advanced Photon Source (APS) at ANL. Phases present in the as-annealed specimen were identified with both Laue diffraction and powder diffraction techniques. The U–10Mo was found to recrystallize due to thermally-induced recrystallization driven by a high density of pre-existing dislocations. A separate in situ annealing experiment was carried out with a Focused Ion Beam processed (FIB) specimen. A similar microstructure transition occurred at a lower temperature of about 460 °C with a much faster transition rate compared to the electro-polished specimen. - Highlights: • TEM annealing experiments were performed in situ at the IVEM facility up to fast reactor temperature. • At 540 °C, the U-10Mo specimen underwent a slow microstructure transition where nano-sized grains were observed to emerge. • UO{sub 2} phase exists at the thin area of the as-annealed specimen whereas U-10Mo γ phase dominated at the thicker part. • Bcc γ U-10Mo recrystallized to become nano-meter sized crystallites near the specimen surface. • A separateannealing experiment was conducted with a FIB processed specimen where similar transition occurred at a lower temperature of 460 °C with a faster rate.« less

γ{\\hbox-}Flash suppression using a gated photomultiplier assembled with an LaBr3(Ce) detector to measure fast neutron capture reactions

NASA Astrophysics Data System (ADS)

Hara, K. Y.; Harada, H.; Toh, Y.; Hori, J.

2013-09-01

A gated photomultiplier tube (PMT) assembled with an LaBr3(Ce) detector was applied toward the prompt γ{hbox-}ray measurement of fast neutron capture reactions. Time-of-flight measurements of the neutron capture reactions of Cl and Al were performed using the 46-MeV electron linear accelerator at the Kyoto University Research Reactor Institute (KURRI) as a pulsed neutron source. The photomultiplier gating technique effectively suppressed the saturation of the PMT output and extended the energy region of the TOF measurement.

Conceptual design of laser fusion reactor KOYO-fast Concepts of reactor system and laser driver

NASA Astrophysics Data System (ADS)

Kozaki, Y.; Miyanaga, N.; Norimatsu, T.; Soman, Y.; Hayashi, T.; Furukawa, H.; Nakatsuka, M.; Yoshida, K.; Nakano, H.; Kubomura, H.; Kawashima, T.; Nishimae, J.; Suzuki, Y.; Tsuchiya, N.; Kanabe, T.; Jitsuno, T.; Fujita, H.; Kawanaka, J.; Tsubakimoto, K.; Fujimoto, Y.; Lu, J.; Matsuoka, S.; Ikegawa, T.; Owadano, Y.; Ueda, K.; Tomabechi, K.; Reactor Design Committee in Ife Forum, Members Of

2006-06-01

We have carried out the design studies of KOYO-Fast laser fusion power plant, using fast ignition cone targets, DPSSL lasers, and LiPb liquid wall chambers. Using fast ignition targets, we could design a middle sized 300 MWe reactor module, with 200 MJ fusion pulse energy and 4 Hz rep-rates, and 1200MWe modular power plants with 4 reactor modules and a 16 Hz laser driver. The liquid wall chambers with free surface cascade flows are proposed for cooling surface quickly enough to a 4 Hz pulse operation. We examined the potential of Yb-YAG ceramic lasers operated at 150˜ 225 K for both implosion and heating laser systems required for a 16-Hz repetition and 8 % total efficiency.

BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, Kyle Allan Lawrence; Williamson, Richard L.; Schwen, Daniel

2015-09-01

BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on themore » formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.« less

Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

DOE Office of Scientific and Technical Information (OSTI.GOV)

Widiawati, Nina, E-mail: nina-widiawati28@yahoo.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id

Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uraniummore » fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from −0.6695443 % at BOC to −0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.« less

FAST CHOPPER BUILDING, TRA665. CONTEXTUAL VIEW: CHOPPER BUILDING IN CENTER. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FAST CHOPPER BUILDING, TRA-665. CONTEXTUAL VIEW: CHOPPER BUILDING IN CENTER. MTR REACTOR SERVICES BUILDING,TRA-635, TO LEFT; MTR BUILDING TO RIGHT. CAMERA FACING WEST. INL NEGATIVE NO. HD42-1. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Can access to spirometry in asthma education centres influence the referral rate by primary physicians for education?

PubMed

Labrecque, M; Lavallée, M; Beauchesne, M F; Cartier, A; Boulet, L P

2006-01-01

Asthma remains uncontrolled in a large number of asthmatic patients. Recent surveys have shown that a minority of asthmatic patients are referred to asthma educators. The objective of the present study was to assess the influence of increased access to spirometry in asthma education centres (AECs) on the rate of patient referrals to these centres by general practitioners. A one-year, prospective, randomized, multicentric, parallel group study was conducted over two consecutive periods of six months each, with added spirometry being offered in the second six-month period to the experimental group. Ten AECs were enrolled in the project. An advertisement describing the AECs' services was sent by mail to a total of 303 general practitioners at the start of each period, inviting them to refer their patients. Measures of the frequency of medical referrals to the AECs were assessed for each period. The group of AECs randomly selected for spirometry in the second six-month period received 48 medical referrals during the first period and 32 during the second one, following proposed spirometry. AECs that had not offered spirometry received five referrals during the first period and seven during the second period. One AEC withdrew a few weeks after the study began and others encountered administrative problems, reducing their ability to provide interventions. Referral to AECs is not yet integrated into the primary care of asthma and offering more rapid access to spirometry in the AECs does not seem to be a significant incentive for such referrals.

Exosomes derived from human amniotic epithelial cells accelerate wound healing and inhibit scar formation.

PubMed

Zhao, Bin; Zhang, Yijie; Han, Shichao; Zhang, Wei; Zhou, Qin; Guan, Hao; Liu, Jiaqi; Shi, Jihong; Su, Linlin; Hu, Dahai

2017-04-01

Wound healing is a highly orchestrated physiological process consisting of a complex events, and scarless wound healing is highly desired for the development and application in clinical medicine. Recently, we have demonstrated that human amniotic epithelial cells (hAECs) promoted wound healing and inhibited scar formation through a paracrine mechanism. However, exosomes (Exo) are one of the most important paracrine factors. Whether exosomes derived from human amniotic epithelial cells (hAECs-Exo) have positive effects on scarless wound healing have not been reported yet. In this study, we examined the role of hAECs-Exo on wound healing in a rat model. We found that hAECs, which exhibit characteristics of both embryonic and mesenchymal stem cells, have the potential to differentiate into all three germ layers. hAECs-Exo ranged from 50 to 150 nm in diameter, and positive for exosomal markers CD9, CD63, CD81, Alix, TSG101 and HLA-G. Internalization of hAECs-Exo promoted the migration and proliferation of fibroblasts. Moreover, the deposition of extracellular matrix (ECM) were partly abolished by the treatment of high concentration of hAECs-Exo (100 μg/mL), which may be through stimulating the expression of matrix metalloproteinase-1 (MMP-1). In vivo animal experiments showed that hAECs-Exo improved the skin wound healing with well-organized collagen fibers. Taken together, These findings represent that hAECs-Exo can be used as a novel hope in cell-free therapy for scarless wound healing.

COUPLED FAST-THERMAL POWER BREEDER REACTOR

DOEpatents

Avery, R.

1961-07-18

A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vanderhaegen, M.; Laboratory of Waves and Acoustic, Institut Langevin, ESPCI ParisTech, 10 rue Vauquelin, 75005 Paris; Paumel, K.

2011-07-01

In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boilingmore » intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)« less

Quick-start of full-scale anaerobic digestion (AD) using aeration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lagerkvist, Anders, E-mail: al@ltu.se; Pelkonen, Markku; Wikström, Tommy

Highlights: • A fast, and original, start up procedure for anaerobic digestors has been applied at full scale. • The development of a methanogenic culture has been documented using fluorescent in situ hybridization. • The technique can be widely applied. - Abstract: A conventional 1300 m{sup 3} continuously stirred anaerobic tank reactor at the city of Boden, north Sweden, which was receiving a feed of both sewage sludge and food waste, was put out of operation due to the build-up of a float phase. The reactor was emptied and cleaned. At start-up there was no methanogenic sludge available, so anmore » unconventional start-up procedure was applied: The reactor was rapidly (8 days with 1200 kg of total solids (TS) added daily) filled with thickened, and slightly acidic sewage sludge, showing only slight methane generation, which was subsequently heated to 55 °C. Then compressed air was blown into the digester and within a month a fully functional methanogenic culture was established. The transfer from acidogenic to methanogenic conditions happened in about one week. As a start-up technique this is fast and cost efficient, it only requires the access of a compressor, electricity and a source of air. In total, about 16 tonnes of oxygen were used. It is proposed that this method may also be used as an operational amendment technique, should a reactor tend to acidify.« less

PULSE SYNTHESIZING GENERATOR

DOEpatents

Kerns, Q.A.

1963-08-01

>An electronlc circuit for synthesizing electrical current pulses having very fast rise times includes several sinewave generators tuned to progressively higher harmonic frequencies with signal amplitudes and phases selectable according to the Fourier series of the waveform that is to be synthesized. Phase control is provided by periodically triggering the generators at precisely controlled times. The outputs of the generators are combined in a coaxial transmission line. Any frequency-dependent delays that occur in the transmission line can be readily compensated for so that the desired signal wave shape is obtained at the output of the line. (AEC)

Converting Maturing Nuclear Sites to Integrated Power Production Islands

DOE PAGES

Solbrig, Charles W.

2011-01-01

Nuclear islands, which are integrated power production sites, could effectively sequester and safeguard the US stockpile of plutonium. A nuclear island, an evolution of the integral fast reactor, utilizes all the Transuranics (Pu plus minor actinides) produced in power production, and it eliminates all spent fuel shipments to and from the site. This latter attribute requires that fuel reprocessing occur on each site and that fast reactors be built on-site to utilize the TRU. All commercial spent fuel shipments could be eliminated by converting all LWR nuclear power sites to nuclear islands. Existing LWR sites have the added advantage ofmore » already possessing a license to produce nuclear power. Each could contribute to an increase in the nuclear power production by adding one or more fast reactors. Both the TRU and the depleted uranium obtained in reprocessing would be used on-site for fast fuel manufacture. Only fission products would be shipped to a repository for storage. The nuclear island concept could be used to alleviate the strain of LWR plant sites currently approaching or exceeding their spent fuel pool storage capacity. Fast reactor breeding ratio could be designed to convert existing sites to all fast reactors, or keep the majority thermal.« less

Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yacout, A. M.; Billone, M. C.

2016-09-16

The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of datamore » were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.« less

Development of Cross Section Library and Application Programming Interface (API)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C. H.; Marin-Lafleche, A.; Smith, M. A.

2014-04-09

The goal of NEAMS neutronics is to develop a high-fidelity deterministic neutron transport code termed PROTEUS for use on all reactor types of interest, but focused primarily on sodium-cooled fast reactors. While PROTEUS-SN has demonstrated good accuracy for homogeneous fast reactor problems and partially heterogeneous fast reactor problems, the simulation results were not satisfactory when applied on fully heterogeneous thermal problems like the Advanced Test Reactor (ATR). This is mainly attributed to the quality of cross section data for heterogeneous geometries since the conventional cross section generation approach does not work accurately for such irregular and complex geometries. Therefore, onemore » of the NEAMS neutronics tasks since FY12 has been the development of a procedure to generate appropriate cross sections for a heterogeneous geometry core.« less

The role of alveolar epithelial cells in initiating and shaping pulmonary immune responses: communication between innate and adaptive immune systems.

PubMed

Chuquimia, Olga D; Petursdottir, Dagbjort H; Rahman, Muhammad J; Hartl, Katharina; Singh, Mahavir; Fernández, Carmen

2012-01-01

Macrophages and dendritic cells have been recognized as key players in the defense against mycobacterial infection. However, more recently, other cells in the lungs such as alveolar epithelial cells (AEC) have been found to play important roles in the defense and pathogenesis of infection. In the present study we first compared AEC with pulmonary macrophages (PuM) isolated from mice in their ability to internalize and control Bacillus Calmette-Guérin (BCG) growth and their capacity as APCs. AEC were able to internalize and control bacterial growth as well as present antigen to primed T cells. Secondly, we compared both cell types in their capacity to secrete cytokines and chemokines upon stimulation with various molecules including mycobacterial products. Activated PuM and AEC displayed different patterns of secretion. Finally, we analyzed the profile of response of AEC to diverse stimuli. AEC responded to both microbial and internal stimuli exemplified by TLR ligands and IFNs, respectively. The response included synthesis by AEC of several factors, known to have various effects in other cells. Interestingly, TNF could stimulate the production of CCL2/MCP-1. Since MCP-1 plays a role in the recruitment of monocytes and macrophages to sites of infection and macrophages are the main producers of TNF, we speculate that both cell types can stimulate each other. Also, another cell-cell interaction was suggested when IFNs (produced mainly by lymphocytes) were able to induce expression of chemokines (IP-10 and RANTES) by AEC involved in the recruitment of circulating lymphocytes to areas of injury, inflammation, or viral infection. In the current paper we confirm previous data on the capacity of AEC regarding internalization of mycobacteria and their role as APC, and extend the knowledge of AEC as a multifunctional cell type by assessing the secretion of a broad array of factors in response to several different types of stimuli.

Is Weight-Based Adjustment of Automatic Exposure Control Necessary for the Reduction of Chest CT Radiation Dose?

PubMed Central

Prakash, Priyanka; Gilman, Matthew D.; Shepard, Jo-Anne O.; Digumarthy, Subba R.

2010-01-01

Objective To assess the effects of radiation dose reduction in the chest CT using a weight-based adjustment of the automatic exposure control (AEC) technique. Materials and Methods With Institutional Review Board Approval, 60 patients (mean age, 59.1 years; M:F = 35:25) and 57 weight-matched patients (mean age, 52.3 years, M:F = 25:32) were scanned using a weight-adjusted AEC and non-weight-adjusted AEC, respectively on a 64-slice multidetector CT with a 0.984:1 pitch, 0.5 second rotation time, 40 mm table feed/rotation, and 2.5 mm section thickness. Patients were categorized into 3 weight categories; < 60 kg (n = 17), 60-90 kg (n = 52), and > 90 kg (n = 48). Patient weights, scanning parameters, CT dose index volumes (CTDIvol) and dose length product (DLP) were recorded, while effective dose (ED) was estimated. Image noise was measured in the descending thoracic aorta. Data were analyzed using a standard statistical package (SAS/STAT) (Version 9.1, SAS institute Inc, Cary, NC). Results Compared to the non-weight-adjusted AEC, the weight-adjusted AEC technique resulted in an average decrease of 29% in CTDIvol and a 27% effective dose reduction (p < 0.0001). With weight-adjusted AEC, the CTDIvol decreased to 15.8, 15.9, and 27.3 mGy for the < 60, 60-90 and > 91 kg weight groups, respectively, compared to 20.3, 27.9 and 32.8 mGy, with non-weight-adjusted AEC. No significant difference was observed for objective image noise between the chest CT acquired with the non-weight-adjusted (15.0 ± 3.1) and weight-adjusted (16.1 ± 5.6) AEC techniques (p > 0.05). Conclusion The results of this study suggest that AEC should be tailored according to patient weight. Without weight-based adjustment of AEC, patients are exposed to a 17 - 43% higher radiation-dose from a chest CT. PMID:20046494

Is weight-based adjustment of automatic exposure control necessary for the reduction of chest CT radiation dose?

PubMed

Prakash, Priyanka; Kalra, Mannudeep K; Gilman, Matthew D; Shepard, Jo-Anne O; Digumarthy, Subba R

2010-01-01

To assess the effects of radiation dose reduction in the chest CT using a weight-based adjustment of the automatic exposure control (AEC) technique. With Institutional Review Board Approval, 60 patients (mean age, 59.1 years; M:F = 35:25) and 57 weight-matched patients (mean age, 52.3 years, M:F = 25:32) were scanned using a weight-adjusted AEC and non-weight-adjusted AEC, respectively on a 64-slice multidetector CT with a 0.984:1 pitch, 0.5 second rotation time, 40 mm table feed/rotation, and 2.5 mm section thickness. Patients were categorized into 3 weight categories; < 60 kg (n = 17), 60-90 kg (n = 52), and > 90 kg (n = 48). Patient weights, scanning parameters, CT dose index volumes (CTDIvol) and dose length product (DLP) were recorded, while effective dose (ED) was estimated. Image noise was measured in the descending thoracic aorta. Data were analyzed using a standard statistical package (SAS/STAT) (Version 9.1, SAS institute Inc, Cary, NC). Compared to the non-weight-adjusted AEC, the weight-adjusted AEC technique resulted in an average decrease of 29% in CTDIvol and a 27% effective dose reduction (p < 0.0001). With weight-adjusted AEC, the CTDIvol decreased to 15.8, 15.9, and 27.3 mGy for the < 60, 60-90 and > 91 kg weight groups, respectively, compared to 20.3, 27.9 and 32.8 mGy, with non-weight-adjusted AEC. No significant difference was observed for objective image noise between the chest CT acquired with the non-weight-adjusted (15.0 +/- 3.1) and weight-adjusted (16.1 +/- 5.6) AEC techniques (p > 0.05). The results of this study suggest that AEC should be tailored according to patient weight. Without weight-based adjustment of AEC, patients are exposed to a 17 - 43% higher radiation-dose from a chest CT.

Homogeneous fast-flux isotope-production reactor

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a liquid metal fast breeder reactor. Lithium target material is dissolved in the liquid metal coolant in order to facilitate the production and removal of tritium.

AEC to Referee, Not Promote, Industry.

PubMed

1971-10-29

A major turnabout in the attitude of the Atomic Energy Commission toward the nuclear power industry was signaled last week by the ntew AEC chairman James R. Schlesinger. With patrician froideur, Schlesinger informed a mass gathering of the nuclear power industry at Bal Harbour, Florida, that from henceforth the AEC woLuld act as the referee of nuclear power, not its promoter. Saying he would dispense with the "anecdotes and clumsy jests" customary on such occasions, Schlesinger served notice on the nuclear banqueters that their cozy relationship with the AEC was at an end. The industry should not expect the AEC to fight its battles: it should take its own case to the public-as the Sierra Club does. Nor did the AEC intend to bend the rules in industry's favor. "We have had a fair amount of advice on how to evade the clear mandate of the federal courts. It is advice we did not think proper to accept," Schlesinger said. Even on matters of engineering quality, the diners were told they knew full well they had "reason to blush." Roused out of any postprandial euphoria by this glacial disdain, the industry representatives heard the new chairman announce the following radical upheavals in official AEC philosophy.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Venkataraman, M.; Natarajan, R.; Raj, Baldev

The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR)more » spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)« less

FAST ACTING CURRENT SWITCH

DOEpatents

Batzer, T.H.; Cummings, D.B.; Ryan, J.F.

1962-05-22

A high-current, fast-acting switch is designed for utilization as a crowbar switch in a high-current circuit such as used to generate the magnetic confinement field of a plasma-confining and heat device, e.g., Pyrotron. The device particularly comprises a cylindrical housing containing two stationary, cylindrical contacts between which a movable contact is bridged to close the switch. The movable contact is actuated by a differential-pressure, airdriven piston assembly also within the housing. To absorb the acceleration (and the shock imparted to the device by the rapidly driven, movable contact), an adjustable air buffer assembly is provided, integrally connected to the movable contact and piston assembly. Various safety locks and circuit-synchronizing means are also provided to permit proper cooperation of the invention and the high-current circuit in which it is installed. (AEC)

Source Book on Plutonium and Its Decontamination

DTIC Science & Technology

1973-09-24

Energy Commisjsion Policy The AEC is the foremoacst regulator of the use of radioactive materials in this country. It derives this power from the...Congress and furthers Its responsibilities in two ways; first, through its licensin~g power and second, by regulation of the activities of its contractors...biological hazard. Plutonium- 239 is of interest because of its abundance in weapons and Pu-238 because of its use in power sources. Half lives for the two

Alarmin S100A8 Activates Alveolar Epithelial Cells in the Context of Acute Lung Injury in a TLR4-Dependent Manner.

PubMed

Chakraborty, Deblina; Zenker, Stefanie; Rossaint, Jan; Hölscher, Anna; Pohlen, Michele; Zarbock, Alexander; Roth, Johannes; Vogl, Thomas

2017-01-01

Alveolar epithelial cells (AECs) are an essential part of the respiratory barrier in lungs for gas exchange and protection against pathogens. Damage to AECs occurs during lung injury and PAMPs/DAMPs have been shown to activate AECs. However, their interplay as well as the mechanism of AECs' activation especially by the alarmin S100A8/A9 is unknown. Thus, our aim was to study the mechanism of activation of AECs (type I and type II) by S100A8 and/or lipopolysaccharide (LPS) and to understand the role of endogenous S100A8/A9 in neutrophil recruitment in the lung. For our studies, we modified a previous protocol for isolation and culturing of murine AECs. Next, we stimulated the cells with S100A8 and/or LPS and analyzed cytokine/chemokine release. We also analyzed the contribution of the known S100-receptors TLR4 and RAGE in AEC activation. In a murine model of lung injury, we investigated the role of S100A8/A9 in neutrophil recruitment to lungs. S100A8 activates type I and type II cells in a dose- and time-dependent manner which could be quantified by the release of IL-6, KC, and MCP-1. We here clearly demonstrate that AEC s are activated by S100A8 via a TLR4-dependent pathway. Surprisingly, RAGE, albeit mainly expressed in lung tissue, plays no role. Additionally, we show that S100A8/A9 is an essential factor for neutrophil recruitment to lungs. We, therefore, conclude that S100A8 promotes acute lung injury via Toll-like receptor 4-dependent activation of AECs.

Decreased Fibronectin Production Significantly Contributes to Dysregulated Repair of Asthmatic Epithelium

PubMed Central

Kicic, Anthony; Hallstrand, Teal S.; Sutanto, Erika N.; Stevens, Paul T.; Kobor, Michael S.; Taplin, Christopher; Paré, Peter D.; Beyer, Richard P.; Stick, Stephen M.; Knight, Darryl A.

2010-01-01

Rationale: Damage to airway epithelium is followed by deposition of extracellular matrix (ECM) and migration of adjacent epithelial cells. We have shown that epithelial cells from children with asthma fail to heal a wound in vitro. Objectives: To determine whether dysregulated ECM production by the epithelium plays a role in aberrant repair in asthma. Methods: Airway epithelial cells (AEC) from children with asthma (n = 36), healthy atopic control subjects (n = 23), and healthy nonatopic control subjects (n = 53) were investigated by microarray, gene expression and silencing, transcript regulation analysis, and ability to close mechanical wounds. Measurements and Main Results: Time to repair a mechanical wound in vitro by AEC from healthy and atopic children was not significantly different and both were faster than AEC from children with asthma. Microarray analysis revealed differential expression of multiple gene sets associated with repair and remodeling in asthmatic AEC. Fibronectin (FN) was the only ECM component whose expression was significantly lower in asthmatic AEC. Expression differences were verified by quantitative polymerase chain reaction and ELISA, and reduced FN expression persisted in asthmatic cells over passage. Silencing of FN expression in nonasthmatic AEC inhibited wound repair, whereas addition of FN to asthmatic AEC restored reparative capacity. Asthmatic AEC failed to synthesize FN in response to wounding or cytokine/growth factor stimulation. Exposure to 5′, 2′deoxyazacytidine had no effect on FN expression and subsequent analysis of the FN promoter did not show evidence of DNA methylation. Conclusions: These data show that the reduced capacity of asthmatic epithelial cells to secrete FN is an important contributor to the dysregulated AEC repair observed in these cells. PMID:20110557

Radioactive waste from decommissioning of fast reactors (through the example of BN-800)

NASA Astrophysics Data System (ADS)

Rybin, A. A.; Momot, O. A.

2017-01-01

Estimation of volume of radioactive waste from operating and decommissioning of fast reactors is introduced. Preliminary estimation has shown that the volume of RW from decommissioning of BN-800 is amounted to 63,000 cu. m. Comparison of the amount of liquid radioactive waste derived from operation of different reactor types is performed. Approximate costs of all wastes disposal for complete decommissioning of BN-800 reactor are estimated amounting up to approx. 145 million.

Should the automatic exposure control system of CT be disabled when scanning patients with endoaortic stents or mechanical heart valves? A phantom study.

PubMed

Di Leo, Giovanni; Spadavecchia, Chiara; Zanardo, Moreno; Secchi, Francesco; Veronese, Ivan; Cantone, Marie Claire; Sardanelli, Francesco

2017-07-01

To estimate the impact of endoaortic stents/mechanical heart valves on the output of an automatic exposure control (AEC) system and CT radiation dose. In this phantom study, seven stents and two valves were scanned with varying tube voltage (80/100/120 kVp), AEC activation (enabled/disabled) and prosthesis (present/absent), for a total of 540 scans. For each prosthesis, the dose-length product (DLP) was compared between scans with the AEC enabled and disabled. Percentage confidence levels for differences due to the prosthesis were calculated. Differences between results with the AEC enabled and disabled were not statistically significant (p ≥ 0.059). In the comparison with and without the prosthesis, DLP was unchanged at 80 kVp and 100 kVp, while a slight increase was observed at 120 kVp. The radiation dose varied from 1.8 mGy to 2.4 mGy without the prosthesis and from 1.8 mGy to 2.5 mGy with the prosthesis (confidence level 37-100%). The effect of the prosthesis on the AEC system was negligible and not clinically relevant. Therefore, disabling the AEC system when scanning these patients is not likely to provide a benefit. • CT-AEC system is not impaired in patients with endoaortic prostheses/heart valves. • Negligible differences may be observed only at 120 kVp. • Disabling the AEC system in these patients is not recommended.

Human amniotic epithelial cells cultured in substitute serum medium maintain their stem cell characteristics for up to four passages.

PubMed

Evron, Ayelet; Goldman, Shlomit; Shalev, Eliezer

2011-11-01

The common applied culture medium in which human amniotic epithelial cells (hAECs) maintain their stem cell characteristics contains fetal calf serum (FCS) and thus is not compatible with possible future clinical applications due to the danger of animal derived pathogens. To overcome this problem, we replaced FCS with serum substitute supplement, a serum substitute used in the in vitro fertilization for embryo development, in the common applied culture medium and cultured hAECs in this substitute serum medium (SSM). Purity validation and characterization of freshly isolated and cultured hAECs was assessed through the expression of stem cell specific markers by RT-PCR (gene expression), by immunofluorescence staining and FACS (protein expression). Furthermore, karyotype was performed at passage four in order to exclude possible chromosome anomalies in hAECs cultured in SSM. The differentiation potential of hAECs into the cardiomyogenic lineage was tested through cardiac Troponin T expression by immunohistochemistry. hAECs cultured in SSM maintained expression of all the major pluripotent genes Sox-2, Oct-4 and Nanog as well as the expression of the embryonic stem cell specific surface antigens SSEA-4, SSEA-3 and TRA-1-60 over four passages. Using cardiac differentiation medium containing 10% serum substitute supplement, hAECs differentiated into cardiac troponin T expressing cells. We can conclude that, hAECs maintain their stem cell characteristics when cultured in SSM for up to 4 passages. This makes possible future clinical applications of these cells more feasible.

Fast-acting nuclear reactor control device

DOEpatents

Kotlyar, Oleg M.; West, Phillip B.

1993-01-01

A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-07-01

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

An underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-basedmore » description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.« less

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

2002-01-01

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

1998-01-01

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

2002-09-24

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Performance of compact fast pyrolysis reactor with Auger-type modules for the continuous liquid biofuel production

NASA Astrophysics Data System (ADS)

Nishimura, Shun; Ebitani, Kohki

2018-01-01

Development of a compact fast pyrolysis reactor constructed using Auger-type technology to afford liquid biofuel with high yield has been an interesting concept in support of local production for local consumption. To establish a widely useable module package, details of the performance of the developing compact module reactor were investigated. This study surveyed the properties of as-produced pyrolysis oil as a function of operation time, and clarified the recent performance of the developing compact fast pyrolysis reactor. Results show that after condensation in the scrubber collector, e.g. approx. 10 h for a 25 kg/h feedstock rate, static performance of pyrolysis oil with approximately 20 MJ/kg (4.8 kcal/g) calorific values were constantly obtained after an additional 14 h. The feeding speed of cedar chips strongly influenced the time for oil condensation process: i.e. 1.6 times higher feeding speed decreased the condensation period by half (approx. 5 h in the case of 40 kg/h). Increasing the reactor throughput capacity is an important goal for the next stage in the development of a compact fast pyrolysis reactor with Auger-type modules.

One-dimensional biomass fast pyrolysis model with reaction kinetics integrated in an Aspen Plus Biorefinery Process Model

DOE PAGES

Humbird, David; Trendewicz, Anna; Braun, Robert; ...

2017-01-12

A biomass fast pyrolysis reactor model with detailed reaction kinetics and one-dimensional fluid dynamics was implemented in an equation-oriented modeling environment (Aspen Custom Modeler). Portions of this work were detailed in previous publications; further modifications have been made here to improve stability and reduce execution time of the model to make it compatible for use in large process flowsheets. The detailed reactor model was integrated into a larger process simulation in Aspen Plus and was stable for different feedstocks over a range of reactor temperatures. Sample results are presented that indicate general agreement with experimental results, but with higher gasmore » losses caused by stripping of the bio-oil by the fluidizing gas in the simulated absorber/condenser. Lastly, this integrated modeling approach can be extended to other well-defined, predictive reactor models for fast pyrolysis, catalytic fast pyrolysis, as well as other processes.« less

One-dimensional biomass fast pyrolysis model with reaction kinetics integrated in an Aspen Plus Biorefinery Process Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Humbird, David; Trendewicz, Anna; Braun, Robert

A biomass fast pyrolysis reactor model with detailed reaction kinetics and one-dimensional fluid dynamics was implemented in an equation-oriented modeling environment (Aspen Custom Modeler). Portions of this work were detailed in previous publications; further modifications have been made here to improve stability and reduce execution time of the model to make it compatible for use in large process flowsheets. The detailed reactor model was integrated into a larger process simulation in Aspen Plus and was stable for different feedstocks over a range of reactor temperatures. Sample results are presented that indicate general agreement with experimental results, but with higher gasmore » losses caused by stripping of the bio-oil by the fluidizing gas in the simulated absorber/condenser. Lastly, this integrated modeling approach can be extended to other well-defined, predictive reactor models for fast pyrolysis, catalytic fast pyrolysis, as well as other processes.« less

Safety and control of accelerator-driven subcritical systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rief, H.; Takahashi, H.

1995-10-01

To study control and safety of accelertor driven nuclear systems, a one point kinetic model was developed and programed. It deals with fast transients as a function of reactivity insertion. Doppler feedback, and the intensity of an external neutron source. The model allows for a simultaneous calculation of an equivalent critical reactor. It was validated by a comparison with a benchmark specified by the Nuclear Energy Agency Committee of Reactor Physics. Additional features are the possibility of inserting a linear or quadratic time dependent reactivity ramp which may account for gravity induced accidents like earthquakes, the possibility to shut downmore » the external neutron source by an exponential decay law of the form exp({minus}t/{tau}), and a graphical display of the power and reactivity changes. The calculations revealed that such boosters behave quite benignly even if they are only slightly subcritical.« less

Agro-ecological class stability decreases in response to climate change projections for the Pacific Northwest, USA

USDA-ARS?s Scientific Manuscript database

Climate change will impact bioclimatic drivers that regulate the geospatial distribution of dryland agro-ecological classes (AECs). Characterizing the geospatial relationship between present AECs and their bioclimatic controls will provide insights into potential future shifts in AECs as climate cha...

Denitrification of groundwater using PHBV blends in packed bed reactors and the microbial diversity.

PubMed

Chu, Libing; Wang, Jianlong

2016-07-01

In the present study, three kinds of biopolymers, PHBV, PHBV/starch and PHBV/bamboo powder (BP) blends were used as carbon source and biofilm carriers for denitrification in packed bed reactors to remove nitrate from groundwater. Results showed that a fast start-up was obtained in bioreactors filled with both PHBV/Starch and PHBV/BP blends without external inocula and it took more than 3 month for PHBV reactor to reach the same loading rate. The PHBV/BP packed reactor exhibited a better nitrate removal efficiency (87.4 ± 7.0%) and less adverse effects in nitrite accumulation and DOC release (below 0.5 mg NO2N L(-1) and 10.5 mg DOC L(-1) in the effluent) during stable operation. Pyrosequencing analysis demonstrated that bacteria belonging to genus Clostridium in phylum Firmicus, which play the primary role in degrading the biopolymers, are the most dominant (33-15% of the sequences). The predominant species in all samples is related to Clostridium crotonatovorans. All the identified 11 genera of denitrifying bacteria affiliated with phylum Proteobacteria and constituted 30-55% in the representative sequences. The PHBV/BP blend is economically attractive carbon source with good denitrification performance. Copyright © 2016 Elsevier Ltd. All rights reserved.

Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, J. D.; Briggs, J. B.; Gulliford, J.

Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energymore » Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning is the critical experiments with fast reactor fuel rods in water, interesting in terms of justification of nuclear safety during transportation and storage of fresh and spent fuel. These reports provide a detailed review of the experiment, designate the area of their application and include results of calculations on modern systems of constants in comparison with the estimated experimental data.« less

Accelerator-driven transmutation of spent fuel elements

DOEpatents

Venneri, Francesco; Williamson, Mark A.; Li, Ning

2002-01-01

An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

Apparatus and method for simulating material damage from a fusion reactor

DOEpatents

Smith, Dale L.; Greenwood, Lawrence R.; Loomis, Benny A.

1989-01-01

An apparatus and method for simulating a fusion environment on a first wall or blanket structure. A material test specimen is contained in a capsule made of a material having a low hydrogen solubility and permeability. The capsule is partially filled with a lithium solution, such that the test specimen is encapsulated by the lithium. The capsule is irradiated by a fast fission neutron source.

Apparatus and method for simulating material damage from a fusion reactor

DOEpatents

Smith, D.L.; Greenwood, L.R.; Loomis, B.A.

1988-05-20

This paper discusses an apparatus and method for simulating a fusion environment on a first wall or blanket structure. A material test specimen is contained in a capsule made of a material having a low hydrogen solubility and permeability. The capsule is partially filled with a lithium solution, such that the test specimen is encapsulated by the lithium. The capsule is irradiated by a fast fission neutron source.

Apparatus and method for simulating material damage from a fusion reactor

DOEpatents

Smith, Dale L.; Greenwood, Lawrence R.; Loomis, Benny A.

1989-03-07

An apparatus and method for simulating a fusion environment on a first wall or blanket structure. A material test specimen is contained in a capsule made of a material having a low hydrogen solubility and permeability. The capsule is partially filled with a lithium solution, such that the test specimen is encapsulated by the lithium. The capsule is irradiated by a fast fission neutron source.

Neutron Imaging Development at China Academy of Engineering Physics (CAEP)

NASA Astrophysics Data System (ADS)

Li, Hang; Wang, Sheng; Cao, Chao; Huo, Heyong; Tang, Bin

Based the China Mianyang Research Reactor (CMRR) and D-T accelerator neutron source, thermal neutron, cold neutron and fast neutron imaging facilities are all installed at China Academy of Engineering Physics (CAEP). Various samples have been imaged by different energy neutrons and shown the neutron imaging application in industry, aerospace and so on. The facilities parameters and recent neutron imaging development will be shown in this paper.

METHOD OF DISSOLVING REFRACTORY ALLOYS

DOEpatents

Helton, D.M.; Savolainen, J.K.

1963-04-23

This patent relates to the dissolution of alloys of uranium with zirconium, thorium, molybdenum, or niobium. The alloy is contacted with an anhydrous solution of mercuric chloride in a low-molecular-weight monohydric alcohol to produce a mercury-containing alcohol slurry. The slurry is then converted to an aqueous system by adding water and driving off the alcohol. The resulting aqueous slurry is electrolyzed in the presence of a mercury cathode to remove the mercury and produce a uranium-bearing aqueous solution. This process is useful for dissolving irradiated nuclear reactor fuels for radiochemical reprocessing by solvent extraction. In addition, zirconium-alloy cladding is selectively removed from uranium dioxide fuel compacts by this means. (AEC)

METHOD OF COMBINING HYDROGEN AND OXYGEN

DOEpatents

McBride, J.P.

1962-02-27

A method is given for the catalytic recombination of radiolytic hydrogen and/or deulerium and oxygen resulting from the subjection or an aqueous thorium oxide or thorium oxide-uranium oxide slurry to ionizing radiation. An improved catalyst is prepared by providing paliadium nitrate in an aqueous thorium oxide sol at a concentration of at least 0.05 grams per gram of thorium oxide and contacting the sol with gaseous hydrogen to form flocculated solids. The solids are then recovered and added to the slurry to provide a palladium concentration of 100 to 1000 parts per million. Recombination is effected by the calalyst at a rate sufficient to support high nuclear reactor power densities. (AEC)

PREPARATION OF HIGH-DENSITY THORIUM OXIDE SPHERES

DOEpatents

McNees, R.A. Jr.; Taylor, A.J.

1963-12-31

A method of preparing high-density thorium oxide spheres for use in pellet beds in nuclear reactors is presented. Sinterable thorium oxide is first converted to free-flowing granules by means such as compression into a compact and comminution of the compact. The granules are then compressed into cubes having a density of 5.0 to 5.3 grams per cubic centimeter. The cubes are tumbled to form spheres by attrition, and the spheres are then fired at 1250 to 1350 deg C. The fired spheres are then polished and fired at a temperature above 1650 deg C to obtain high density. Spherical pellets produced by this method are highly resistant to mechanical attrition hy water. (AEC)

CRUCIFORM CONTROL ROD JOINT

DOEpatents

Thorp, A.G. II

1962-08-01

An invention is described which relates to nuclear reactor control rod components and more particularly to a joint between cruciform control rod members and cruciform control rod follower members. In one embodiment this invention provides interfitting crossed arms at adjacent ends of a control rod and its follower in abutting relation. This holds the members against relative opposite longitudinal movement while a compression member keys the arms against relative opposite rotation around a common axis. Means are also provided for centering the control rod and its follower on a common axis and for selectively releasing the control rod from its follower for the insertion of a replacement of the control rod and reuse of the follower. (AEC)

METHOD OF PREPARING RADIOACTIVE CESIUM SOURCES

DOEpatents

Quinby, T.C.

1963-12-17

A method of preparing a cesium-containing radiation source with physical and chemical properties suitable for high-level use is presented. Finely divided silica is suspended in a solution containing cesium, normally the fission-product isotope cesium 137. Sodium tetraphenyl boron is then added to quantitatively precipitate the cesium. The cesium-containing precipitate is converted to borosilicate glass by heating to the melting point and cooling. Up to 60 weight percent cesium, with a resulting source activity of up to 21 curies per gram, is incorporated in the glass. (AEC)

FAST CHOPPER BUILDING, TRA665, INTERIOR. UPPER LEVEL. CONCRETE WALLS. INL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FAST CHOPPER BUILDING, TRA-665, INTERIOR. UPPER LEVEL. CONCRETE WALLS. INL NEGATIVE NO. HD42-2. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Thermionic fast spectrum reactor-converter on the basis of multi-cell TFE

NASA Astrophysics Data System (ADS)

Ponomarev-Stepnoi, N. N.; Kompaniets, G. V.; Poliakov, D. N.; Stepennov, B. S.; Andreev, P. V.; Zhabotinsky, E. E.; Nikolaev, Yu. V.; Lapochkin, N. V.

2001-02-01

Today Russian experts have technological experience in development of in-core thermionic converters for reactors of space nuclear power plants. Such a converter contains nuclear fuel inside and really represents a fuel element of a reactor. Two types of reactors can be considered on the basis of these thermionic fuel elements: with thermal or intermediate neutron spectrum, and with fast neutron spectrum. The first type is characterized by the presence of moderator in core that ensures most economical usage of nuclear fuel. The estimation shows that moderated system is the most effective in the power range of about 5 ... 100 kWe. The power systems of higher level are characterized by larger dimensions due to the presence of moderator. The second type of reactor is considered for higher power levels. This power range is about hundreds kWe. Dimensions of the fast reactor and core configuration are determined by the necessity to ensure the required net output power, on the one hand, and the necessity to ensure critical state on the other hand. In the case of using in-core thermionic fuel elements of the specified design, minimal reactor output power is determined by reactor criticality condition, and maximum reactor power output is determined by specifications and launcher capabilities. In the present paper the effective multiplication factor of a fast spectrum reactor on the basis of a multi-cell TFE developed by ``Lutch'' is considered a function of the total number of TFEs in the reactor. The MCU Monte-Carlo code, developed in Russia (Alekseev, et al., 1991), was used for computations. TFE computational models are placed in the nodes of a uniform triangular lattice and surrounded with pressure vessel and a side reflector. Ordinary fuel pins without thermionic converters were used instead of some TFEs to optimize criticality parameters, dimensions and output power of the reactor. General weight parameters of the reactor are presented in the paper. .

The clinical course of asymptomatic esophageal candidiasis incidentally diagnosed in general health inspection.

PubMed

Lee, Sang Pyo; Sung, In-Kyung; Kim, Jeong Hwan; Lee, Sun-Young; Park, Hyung Seok; Shim, Chan Sup

2015-01-01

Esophageal candidiasis mostly occurs in the immunocompromised host. However, it may also affect healthy people and is frequently asymptomatic. The clinical significance of asymptomatic esophageal candidiasis (AEC) is still unclear. The aims of the study were to investigate the prevalence of AEC during health inspection and to identify its predisposing factors and clinical significance. A total of 49,497 subjects who underwent a health inspection that included upper endoscopy were enrolled. We retrospectively reviewed the subject's self-reporting questionnaires, medical records and endoscopic findings. We considered "long-term" follow-up to be >6 months with at least one more follow-up endoscopy. One hundred and seventy (0.4%) subjects were endoscopically diagnosed as esophageal candidiasis and 141 subjects were AEC. Multivariate analysis revealed that old age (≥60 years) was an independent risk factor for AEC (OR, 1.862, p = 0.005). The number of subjects with long-term follow-up was 79 (195.3 person-years). Among these, AEC of 64 subjects (81.0%) had disappeared on the follow-up endoscopy and was not recurrent. The other 15 subjects had AEC diagnosed more than once on the follow-up endoscopy, and 5 of them were spontaneously healed during the follow-up period. The remaining 10 subjects whose candidiasis was sustained up to the last endoscopy did not complain of symptoms during the follow-up period, and their endoscopic findings did not worsen. AEC is rare and old age is the only risk factor. AEC does not require medical care because it is a self-limited disease.

Can access to spirometry in asthma education centres influence the referral rate by primary physicians for education?

PubMed Central

Labrecque, M; Lavallée, M; Beauchesne, MF; Cartier, A; Boulet, LP

2006-01-01

BACKGROUND AND OBJECTIVES: Asthma remains uncontrolled in a large number of asthmatic patients. Recent surveys have shown that a minority of asthmatic patients are referred to asthma educators. The objective of the present study was to assess the influence of increased access to spirometry in asthma education centres (AECs) on the rate of patient referrals to these centres by general practitioners. METHODS: A one-year, prospective, randomized, multicentric, parallel group study was conducted over two consecutive periods of six months each, with added spirometry being offered in the second six-month period to the experimental group. Ten AECs were enrolled in the project. An advertisement describing the AECs’ services was sent by mail to a total of 303 general practitioners at the start of each period, inviting them to refer their patients. Measures of the frequency of medical referrals to the AECs were assessed for each period. RESULTS: The group of AECs randomly selected for spirometry in the second six-month period received 48 medical referrals during the first period and 32 during the second one, following proposed spirometry. AECs that had not offered spirometry received five referrals during the first period and seven during the second period. One AEC withdrew a few weeks after the study began and others encountered administrative problems, reducing their ability to provide interventions. CONCLUSIONS: Referral to AECs is not yet integrated into the primary care of asthma and offering more rapid access to spirometry in the AECs does not seem to be a significant incentive for such referrals. PMID:17149461

CT dose modulation using automatic exposure control in whole-body PET/CT: effects of scout imaging direction and arm positioning.

PubMed

Inoue, Yusuke; Nagahara, Kazunori; Kudo, Hiroko; Itoh, Hiroyasu

2018-01-01

Automatic exposure control (AEC) modulates tube current and consequently X-ray exposure in CT. We investigated the behavior of AEC systems in whole-body PET/CT. CT images of a whole-body phantom were acquired using AEC on two scanners from different manufactures. The effects of scout imaging direction and arm positioning on dose modulation were evaluated. Image noise was assessed in the chest and upper abdomen. On one scanner, AEC using two scout images in the posteroanterior (PA) and lateral (Lat) directions provided relatively constant image noise along the z-axis with the arms at the sides. Raising the arms increased tube current in the head and neck and decreased it in the body trunk. Image noise increased in the upper abdomen, suggesting excessive reduction in radiation exposure. AEC using the PA scout alone strikingly increased tube current and reduced image noise in the shoulder. Raising the arms did not substantially influence dose modulation and decreased noise in the abdomen. On the other scanner, AEC using the PA scout alone or Lat scout alone resulted in similar dose modulation. Raising the arms increased tube current in the head and neck and decreased it in the trunk. Image noise was higher in the upper abdomen than in the middle and lower chest, and was not influenced by arm positioning. CT dose modulation using AEC may vary greatly depending on scout direction. Raising the arms tended to decrease radiation exposure; however, the effect depends on scout direction and the AEC system.

The adenylate energy charge as a new and useful indicator of capture stress in chondrichthyans.

PubMed

Guida, Leonardo; Walker, Terence I; Reina, Richard D

2016-02-01

Quantifying the physiological stress response of chondrichthyans to capture has assisted the development of fishing practices conducive to their survival. However, currently used indicators of stress show significant interspecific and intraspecific variation in species' physiological responses and tolerances to capture. To improve our understanding of chondrichthyan stress physiology and potentially reduce variation when quantifying the stress response, we investigated the use of the adenylate energy charge (AEC); a measure of available metabolic energy. To determine tissues sensitive to metabolic stress, we extracted samples of the brain, heart, liver, white muscle and blood from gummy sharks (Mustelus antarcticus) immediately following gillnet capture and after 3 h recovery under laboratory conditions. Capture caused significant declines in liver, white muscle and blood AEC, whereas no decline was detected in the heart and brain AEC. Following 3 h of recovery from capture, the AEC of the liver and blood returned to "unstressed" levels (control values) whereas white muscle AEC was not significantly different to that immediately after capture. Our results show that the liver is most sensitive to metabolic stress and white muscle offers a practical method to sample animals non-lethally for determination of the AEC. The AEC is a highly informative indicator of stress and unlike current indicators, it can directly measure the change in available energy and thus the metabolic stress experienced by a given tissue. Cellular metabolism is highly conserved across organisms and, therefore, we think the AEC can also provide a standardised form of measuring capture stress in many chondrichthyan species.

JPRS Report, Science & Technology, China: Energy.

DTIC Science & Technology

1992-03-30

breeder reactors should become...the primary type of reactors . In developing breeder reactors , we should follow the path of using metal fuel. Breeder reactors give us more time to...first reactor used for power generation was a fast reactor : the " Breeder 1" reactor at the Idaho National Reactor Test Center which was used to

A Comparison of Fast-Spectrum and Moderated Space Fission Reactors

NASA Astrophysics Data System (ADS)

Poston, David I.

2005-02-01

The reactor neutron spectrum is one of the fundamental design choices for any fission reactor, but the implications of using a moderated spectrum are vastly different for space reactors as opposed to terrestrial reactors. In addition, the pros and cons of neutron spectra are significantly different among many of the envisioned space power applications. This paper begins with a discussion of the neutronic differences between fast-spectrum and moderated space reactors. This is followed by a discussion of the pros and cons of fast-spectrum and moderated space reactors separated into three areas—technical risk, performance, and safety/safeguards. A mix of quantitative and qualitative arguments is presented, and some conclusions generally can be made regarding neutron spectrum and space power application. In most cases, a fast-spectrum system appears to be the better alternative (mostly because of simplicity and higher potential operating temperatures); however, in some cases, such as a low-power (<100-kWt) surface reactor, a moderated spectrum could provide a better approach. In all cases, the determination of which spectrum is preferred is a strong function of the metrics provided by the "customer"— i.e., if a certain level of performance is required, it could provide a different solution than if a certain level of safeguards is required (which in some cases could produce a null solution). The views expressed in this document are those of the author and do not necessarily reflect agreement by the Government.

Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)

NASA Astrophysics Data System (ADS)

Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

2014-06-01

Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated.

On fundamental quality of fission chain reaction to oppose rapid runaways of nuclear reactors

NASA Astrophysics Data System (ADS)

Kulikov, G. G.; Shmelev, A. N.; Apse, V. A.; Kulikov, E. G.

2017-01-01

It has been shown that the in-hour equation characterizes the barriers and resistibility of fission chain reaction (FCR) against rapid runaways in nuclear reactors. Traditionally, nuclear reactors are characterized by the presence of barriers based on delayed and prompt neutrons. A new barrier based on the reflector neutrons that can occur when the fast reactor core is surrounded by a weakly absorbing neutron reflector with heavy atomic weight was proposed. It has been shown that the safety of this fast reactor is substantially improved, and considerable elongation of prompt neutron lifetime "devalues" the role of delayed neutron fraction as the maximum permissible reactivity for the reactor safety.

Method of locating a leaking fuel element in a fast breeder power reactor

DOEpatents

Honekamp, John R.; Fryer, Richard M.

1978-01-01

Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

Preliminary Options Assessment of Versatile Irradiation Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sen, Ramazan Sonat

The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 xmore » 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.« less

Slow clean-up for fast reactor

NASA Astrophysics Data System (ADS)

Banks, Michael

2008-05-01

The year 2300 is so distant that one may be forgiven for thinking of it only in terms of science fiction. But this is the year that workers at the Dounreay power station in Northern Scotland - the UK's only centre for research into "fast" nuclear reactors - term as the "end point" by which time the site will be completely clear of radioactive material. More than 180 facilities - including the iconic dome that housed the Dounreay Fast Reactor (DFR) - were built at at the site since it opened in 1959, with almost 50 having been used to handle radioactive material.

ION SOURCE WITH SPACE CHARGE NEUTRALIZATION

DOEpatents

Flowers, J.W.; Luce, J.S.; Stirling, W.L.

1963-01-22

This patent relates to a space charge neutralized ion source in which a refluxing gas-fed arc discharge is provided between a cathode and a gas-fed anode to provide ions. An electron gun directs a controlled, monoenergetic electron beam through the discharge. A space charge neutralization is effected in the ion source and accelerating gap by oscillating low energy electrons, and a space charge neutralization of the source exit beam is effected by the monoenergetic electron beam beyond the source exit end. The neutralized beam may be accelerated to any desired energy at densities well above the limitation imposed by Langmuir-Child' s law. (AEC)

miR-34 miRNAs Regulate Cellular Senescence in Type II Alveolar Epithelial Cells of Patients with Idiopathic Pulmonary Fibrosis

PubMed Central

Disayabutr, Supparerk; Kim, Eun Kyung; Cha, Seung-Ick; Green, Gary; Naikawadi, Ram P.; Jones, Kirk D.; Golden, Jeffrey A.; Schroeder, Aaron; Matthay, Michael A.; Kukreja, Jasleen; Erle, David J.; Collard, Harold R.; Wolters, Paul J.

2016-01-01

Pathologic features of idiopathic pulmonary fibrosis (IPF) include genetic predisposition, activation of the unfolded protein response, telomere attrition, and cellular senescence. The mechanisms leading to alveolar epithelial cell (AEC) senescence are poorly understood. MicroRNAs (miRNAs) have been reported as regulators of cellular senescence. Senescence markers including p16, p21, p53, and senescence-associated β-galactosidase (SA-βgal) activity were measured in type II AECs from IPF lungs and unused donor lungs. miRNAs were quantified in type II AECs using gene expression arrays and quantitative RT-PCR. Molecular markers of senescence (p16, p21, and p53) were elevated in IPF type II AECs. SA-βgal activity was detected in a greater percentage in type II AECs isolated from IPF patients (23.1%) compared to patients with other interstitial lung diseases (1.2%) or normal controls (0.8%). The relative levels of senescence-associated miRNAs miR-34a, miR-34b, and miR-34c, but not miR-20a, miR-29c, or miR-let-7f were significantly higher in type II AECs from IPF patients. Overexpression of miR-34a, miR-34b, or miR-34c in lung epithelial cells was associated with higher SA-βgal activity (27.8%, 35.1%, and 38.2%, respectively) relative to control treated cells (8.8%). Targets of miR-34 miRNAs, including E2F1, c-Myc, and cyclin E2, were lower in IPF type II AECs. These results show that markers of senescence are uniquely elevated in IPF type II AECs and suggest that the miR-34 family of miRNAs regulate senescence in IPF type II AECs. PMID:27362652

Characteristics comparison between a cyclotron-based neutron source and KUR-HWNIF for boron neutron capture therapy

NASA Astrophysics Data System (ADS)

Tanaka, H.; Sakurai, Y.; Suzuki, M.; Masunaga, S.; Kinashi, Y.; Kashino, G.; Liu, Y.; Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Maruhashi, A.; Ono, K.

2009-06-01

At Kyoto University Research Reactor Institute (KURRI), 275 clinical trials of boron neutron capture therapy (BNCT) have been performed as of March 2006, and the effectiveness of BNCT has been revealed. In order to further develop BNCT, it is desirable to supply accelerator-based epithermal-neutron sources that can be installed near the hospital. We proposed the method of filtering and moderating fast neutrons, which are emitted from the reaction between a beryllium target and 30-MeV protons accelerated by a cyclotron accelerator, using an optimum moderator system composed of iron, lead, aluminum and calcium fluoride. At present, an epithermal-neutron source is under construction from June 2008. This system consists of a cyclotron accelerator, beam transport system, neutron-yielding target, filter, moderator and irradiation bed. In this article, an overview of this system and the properties of the treatment neutron beam optimized by the MCNPX Monte Carlo neutron transport code are presented. The distribution of biological effect weighted dose in a head phantom compared with that of Kyoto University Research Reactor (KUR) is shown. It is confirmed that for the accelerator, the biological effect weighted dose for a deeply situated tumor in the phantom is 18% larger than that for KUR, when the limit dose of the normal brain is 10 Gy-eq. The therapeutic time of the cyclotron-based neutron sources are nearly one-quarter of that of KUR. The cyclotron-based epithermal-neutron source is a promising alternative to reactor-based neutron sources for treatments by BNCT.

A Methodology for the Integration of a Mechanistic Source Term Analysis in a Probabilistic Framework for Advanced Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grabaskas, Dave; Brunett, Acacia J.; Bucknor, Matthew

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies to conduct a mechanistic source term (MST) analysis for event sequences that could result in the release ofmore » radionuclides. The MST analysis seeks to realistically model and assess the transport, retention, and release of radionuclides from the reactor to the environment. The MST methods developed during this project seek to satisfy the requirements of the Mechanistic Source Term element of the ASME/ANS Non-LWR PRA standard. The MST methodology consists of separate analysis approaches for risk-significant and non-risk significant event sequences that may result in the release of radionuclides from the reactor. For risk-significant event sequences, the methodology focuses on a detailed assessment, using mechanistic models, of radionuclide release from the fuel, transport through and release from the primary system, transport in the containment, and finally release to the environment. The analysis approach for non-risk significant event sequences examines the possibility of large radionuclide releases due to events such as re-criticality or the complete loss of radionuclide barriers. This paper provides details on the MST methodology, including the interface between the MST analysis and other elements of the PRA, and provides a simplified example MST calculation for a sodium fast reactor.« less

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

Deflection Measurements of a Thermally Simulated Nuclear Core Using a High-Resolution CCD-Camera

NASA Technical Reports Server (NTRS)

Stanojev, B. J.; Houts, M.

2004-01-01

Space fission systems under consideration for near-term missions all use compact. fast-spectrum reactor cores. Reactor dimensional change with increasing temperature, which affects neutron leakage. is the dominant source of reactivity feedback in these systems. Accurately measuring core dimensional changes during realistic non-nuclear testing is therefore necessary in predicting the system nuclear equivalent behavior. This paper discusses one key technique being evaluated for measuring such changes. The proposed technique is to use a Charged Couple Device (CCD) sensor to obtain deformation readings of electrically heated prototypic reactor core geometry. This paper introduces a technique by which a single high spatial resolution CCD camera is used to measure core deformation in Real-Time (RT). Initial system checkout results are presented along with a discussion on how additional cameras could be used to achieve a three- dimensional deformation profile of the core during test.

Telomere dysfunction in alveolar epithelial cells causes lung remodeling and fibrosis

PubMed Central

Naikawadi, Ram P.; Disayabutr, Supparerk; Mallavia, Benat; Donne, Matthew L.; Green, Gary; La, Janet L.; Rock, Jason R.; Looney, Mark R.; Wolters, Paul J.

2016-01-01

Telomeres are short in type II alveolar epithelial cells (AECs) of patients with idiopathic pulmonary fibrosis (IPF). Whether dysfunctional telomeres contribute directly to development of lung fibrosis remains unknown. The objective of this study was to investigate whether telomere dysfunction in type II AECs, mediated by deletion of the telomere shelterin protein TRF1, leads to pulmonary fibrosis in mice (SPC-Cre TRF1fl/fl mice). Deletion of TRF1 in type II AECs for 2 weeks increased γH2AX DNA damage foci, but not histopathologic changes in the lung. Deletion of TRF1 in type II AECs for up to 9 months resulted in short telomeres and lung remodeling characterized by increased numbers of type II AECs, α-smooth muscle actin+ mesenchymal cells, collagen deposition, and accumulation of senescence-associated β-galactosidase+ lung epithelial cells. Deletion of TRF1 in collagen-expressing cells caused pulmonary edema, but not fibrosis. These results demonstrate that prolonged telomere dysfunction in type II AECs, but not collagen-expressing cells, leads to age-dependent lung remodeling and fibrosis. We conclude that telomere dysfunction in type II AECs is sufficient to cause lung fibrosis, and may be a dominant molecular defect causing IPF. SPC-Cre TRF1fl/fl mice will be useful for assessing cellular and molecular mechanisms of lung fibrosis mediated by telomere dysfunction. PMID:27699234

U.S.-Russian Civilian Nuclear Cooperation Agreement: Issues for Congress

DTIC Science & Technology

2010-07-09

for nuclear cooperation in 1973 to allow for cooperation in controlled thermonuclear fusion, fast breeder reactors , and fundamental research. The...that a 123 agreement is needed to implement this action plan—for example, full scale technical cooperation on fast reactors and demonstration of...superpowers convened a Joint Coordinating Committee for Civilian Reactor Safety starting in 1988.10 After the fall of the Soviet Union and prior to July

Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

MH Lane

2006-02-15

This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

Decay heat of sodium fast reactor: Comparison of experimental measurements on the PHENIX reactor with calculations performed with the French DARWIN package

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benoit, J. C.; Bourdot, P.; Eschbach, R.

2012-07-01

A Decay Heat (DH) experiment on the whole core of the French Sodium-Cooled Fast Reactor PHENIX has been conducted in May 2008. The measurements began an hour and a half after the shutdown of the reactor and lasted twelve days. It is one of the experiments used for the experimental validation of the depletion code DARWIN thereby confirming the excellent performance of the aforementioned code. Discrepancies between measured and calculated decay heat do not exceed 8%. (authors)

The effect of the composition of plutonium loaded on the reactivity change and the isotopic composition of fuel produced in a fast reactor

NASA Astrophysics Data System (ADS)

Blandinskiy, V. Yu.

2014-12-01

This paper presents the results of a numerical investigation into burnup and breeding of nuclides in metallic fuel consisting of a mixture of plutonium and depleted uranium in a fast reactor with sodium coolant. The feasibility of using plutonium contained in spent nuclear fuel from domestic thermal reactors and weapons-grade plutonium is discussed. It is shown that the largest production of secondary fuel and the least change in the reactivity over the reactor lifetime can be achieved when employing plutonium contained in spent nuclear fuel from a reactor of the RBMK-1000 type.

History of the United States Atomic Energy Commission. Volume II. 1947 / 1952, Atomic Shield

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hewlett, Richard G.; Duncan, Francis

1972-01-01

Sponsored by the Historical Advisory Committee of the Atomic Energy Commission (AEC), this 2-volume series provides an unclassified history of the AEC. Volume I is subtitled ''The New World'' and covers the AEC from 1939 through 1946. This volume, Volume II, is subtitled ''Atomic Shield'' and covers the years 1947 through 1952.

History of the United States Atomic Energy Commission. Volume I. 1939 / 1946, The New World

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hewlett, Richard G.; Anderson, Jr., Oscar E.

1962-01-01

Sponsored by the Historical Advisory Committee of the Atomic Energy Commission (AEC), this 2-volume series provides an unclassified history of the AEC. This volume is subtitled ''The New World'' and covers the AEC from 1939 through 1946. Volume II, is subtitled ''Atomic Shield'' and covers the years 1947 through 1952.

Risks Associated with Federal Construction Projects

DTIC Science & Technology

2011-06-01

awarding contracts for construction projects (USACE, 2010). BIM offers a method to effectively design a facility while maximizing work performance during...includes Requirements, Programming, Funding, Solicitation, AEC Evaluation, Award , Project Validation, Design and Construction, and Project Management...includes the Solicitation, AEC Evaluation, and Award Steps. In this Phase, BIM is only used in the Solicitation and the AEC Evaluation steps

Prediction of the thermophysical properties of molten salt fast reactor fuel from first-principles

NASA Astrophysics Data System (ADS)

Gheribi, A. E.; Corradini, D.; Dewan, L.; Chartrand, P.; Simon, C.; Madden, P. A.; Salanne, M.

2014-05-01

Molten fluorides are known to show favourable thermophysical properties which make them good candidate coolants for nuclear fission reactors. Here we investigate the special case of mixtures of lithium fluoride and thorium fluoride, which act both as coolant and as fuel in the molten salt fast reactor concept. By using ab initio parameterised polarisable force fields, we show that it is possible to calculate the whole set of properties (density, thermal expansion, heat capacity, viscosity and thermal conductivity) which are necessary for assessing the heat transfer performance of the melt over the whole range of compositions and temperatures. We then deduce from our calculations several figures of merit which are important in helping the optimisation of the design of molten salt fast reactors.

Alarmin S100A8 Activates Alveolar Epithelial Cells in the Context of Acute Lung Injury in a TLR4-Dependent Manner

PubMed Central

Chakraborty, Deblina; Zenker, Stefanie; Rossaint, Jan; Hölscher, Anna; Pohlen, Michele; Zarbock, Alexander; Roth, Johannes; Vogl, Thomas

2017-01-01

Alveolar epithelial cells (AECs) are an essential part of the respiratory barrier in lungs for gas exchange and protection against pathogens. Damage to AECs occurs during lung injury and PAMPs/DAMPs have been shown to activate AECs. However, their interplay as well as the mechanism of AECs’ activation especially by the alarmin S100A8/A9 is unknown. Thus, our aim was to study the mechanism of activation of AECs (type I and type II) by S100A8 and/or lipopolysaccharide (LPS) and to understand the role of endogenous S100A8/A9 in neutrophil recruitment in the lung. For our studies, we modified a previous protocol for isolation and culturing of murine AECs. Next, we stimulated the cells with S100A8 and/or LPS and analyzed cytokine/chemokine release. We also analyzed the contribution of the known S100-receptors TLR4 and RAGE in AEC activation. In a murine model of lung injury, we investigated the role of S100A8/A9 in neutrophil recruitment to lungs. S100A8 activates type I and type II cells in a dose- and time-dependent manner which could be quantified by the release of IL-6, KC, and MCP-1. We here clearly demonstrate that AEC s are activated by S100A8 via a TLR4-dependent pathway. Surprisingly, RAGE, albeit mainly expressed in lung tissue, plays no role. Additionally, we show that S100A8/A9 is an essential factor for neutrophil recruitment to lungs. We, therefore, conclude that S100A8 promotes acute lung injury via Toll-like receptor 4-dependent activation of AECs. PMID:29180999

Alveolar Macrophages Prevent Lethal Influenza Pneumonia By Inhibiting Infection Of Type-1 Alveolar Epithelial Cells

PubMed Central

Cardani, Amber; Boulton, Adam; Kim, Taeg S.; Braciale, Thomas J.

2017-01-01

The Influenza A virus (IAV) is a major human pathogen that produces significant morbidity and mortality. To explore the contribution of alveolar macrophages (AlvMΦs) in regulating the severity of IAV infection we employed a murine model in which the Core Binding Factor Beta gene is conditionally disrupted in myeloid cells. These mice exhibit a selective deficiency in AlvMΦs. Following IAV infection these AlvMΦ deficient mice developed severe diffuse alveolar damage, lethal respiratory compromise, and consequent lethality. Lethal injury in these mice resulted from increased infection of their Type-1 Alveolar Epithelial Cells (T1AECs) and the subsequent elimination of the infected T1AECs by the adaptive immune T cell response. Further analysis indicated AlvMΦ-mediated suppression of the cysteinyl leukotriene (cysLT) pathway genes in T1AECs in vivo and in vitro. Inhibition of the cysLT pathway enzymes in a T1AECs cell line reduced the susceptibility of T1AECs to IAV infection, suggesting that AlvMΦ-mediated suppression of this pathway contributes to the resistance of T1AECs to IAV infection. Furthermore, inhibition of the cysLT pathway enzymes, as well as blockade of the cysteinyl leukotriene receptors in the AlvMΦ deficient mice reduced the susceptibility of their T1AECs to IAV infection and protected these mice from lethal infection. These results suggest that AlvMΦs may utilize a previously unappreciated mechanism to protect T1AECs against IAV infection, and thereby reduce the severity of infection. The findings further suggest that the cysLT pathway and the receptors for cysLT metabolites represent potential therapeutic targets in severe IAV infection. PMID:28085958

Primary Airway Epithelial Cell Gene Editing Using CRISPR-Cas9.

PubMed

Everman, Jamie L; Rios, Cydney; Seibold, Max A

2018-01-01

The adaptation of the clustered regularly interspaced short palindromic repeats (CRISPR) and CRISPR associated endonuclease 9 (CRISPR-Cas9) machinery from prokaryotic organisms has resulted in a gene editing system that is highly versatile, easily constructed, and can be leveraged to generate human cells knocked out (KO) for a specific gene. While standard transfection techniques can be used for the introduction of CRISPR-Cas9 expression cassettes to many cell types, delivery by this method is not efficient in many primary cell types, including primary human airway epithelial cells (AECs). More efficient delivery in AECs can be achieved through lentiviral-mediated transduction, allowing the CRISPR-Cas9 system to be integrated into the genome of the cell, resulting in stable expression of the nuclease machinery and increasing editing rates. In parallel, advancements have been made in the culture, expansion, selection, and differentiation of AECs, which allow the robust generation of a bulk edited AEC population from transduced cells. Applying these methods, we detail here our latest protocol to generate mucociliary epithelial cultures knocked out for a specific gene from donor-isolated primary human basal airway epithelial cells. This protocol includes methods to: (1) design and generate lentivirus which targets a specific gene for KO with CRISPR-Cas9 machinery, (2) efficiently transduce AECs, (3) culture and select for a bulk edited AEC population, (4) molecularly screen AECs for Cas9 cutting and specific sequence edits, and (5) further expand and differentiate edited cells to a mucociliary airway epithelial culture. The AEC knockouts generated using this protocol provide an excellent primary cell model system with which to characterize the function of genes involved in airway dysfunction and disease.

Use of automatic exposure control in multislice computed tomography of the coronaries: comparison of 16‐slice and 64‐slice scanner data with conventional coronary angiography

PubMed Central

Deetjen, Anja; Möllmann, Susanne; Conradi, Guido; Rolf, Andreas; Schmermund, Axel; Hamm, Christian W; Dill, Thorsten

2007-01-01

Objective To evaluate the radiation‐dose‐reduction potential of automatic exposure control (AEC) in 16‐slice and 64‐slice multislice computed tomography (MSCT) of the coronary arteries (computed tomography angiography, CTA) in patients. The rapid growth in MSCT CTA emphasises the necessity of adjusting technique factors to reduce radiation dose exposure. Design A retrospective data analysis was performed for 154 patients who had undergone MSCT CTA. Group 1 (n = 56) had undergone 16‐slice MSCT without AEC, and group 2 (n = 51), with AEC. In group 1, invasive coronary angiography (ICA) had been performed in addition. Group 3 (n = 47) had been examined using a 64‐slice scanner (with AEC, without ECG‐triggered tube current modulation). Results In group 1, the mean (SD) effective dose (ED) for MSCT CTA was 9.76 (1.84) mSv and for ICA it was 2.6 (1.27) mSv. In group 2, the mean ED for MSCT CTA was 5.83 (1.73) mSv, which signifies a 42.8% dose reduction for CTA by the use of AEC. In comparison to ICA, MSCT CTA without AEC shows a 3.8‐fold increase in radiation dose, and the radiation dose of CTA with AEC was increased by a factor of 1.9. In group 3, the mean ED for MSCT CTA was 13.58 (2.80) mSV. Conclusions This is the first study to show the significant dose‐reduction potential (42.8%) of AEC in MSCT CTA in patients. This relatively new technique can be used to optimise the radiation dose levels in MSCT CTA. PMID:17395667

Results of the Radiological Survey of the Iowa Army Ammunition Plant, Middletown, Iowa

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murray, M.E.

2001-07-17

At the request of the U.S. Department of Energy (DOE), a team from Oak Ridge National Laboratory conducted an indoor radiological survey of property at the Iowa Army Ammunition Plant (IAAAP), Middletown, Iowa in June 2000. The purpose of the survey was to determine if radioactive residuals resulting from previous Atomic Energy Commission (AEC) activities were present inside selected Line 1 buildings at the IAAAP and conduct sampling in those areas of previous AEC operations that utilized radioactive components at some point during the manufacturing process, in order to evaluate any possible immediate health hazards and to collect sufficient informationmore » to determine the next type of survey. The AEC occupied portions of IAAAP from 1947 to 1975 to assemble nuclear weapons. The surveyed areas were identified through interviews with current and former IAAAP employees who had worked at the plant during AEC's tenure, and from AEC records.« less

The New Status of Argon-37 Artificial Neutrino Source Project

NASA Astrophysics Data System (ADS)

Abdurashitov, J. N.; Gavrin, V. N.; Mirmov, I. N.; Veretenkin, E. P.; Yants, V. E.; Oshkanov, N. N.; Karpenko, A. I.; Maltsev, V. V.; Barsanov, V. I.; Trubin, K. S.; Zlokazov, S. B.; Khomyakov, Yu. S.; Poplavsky, V. M.; Saraeva, T. O.; Vasiliev, B. A.; Mishin, O. V.; Bowles, T. J.; Teasdale, W. A.; Lande, K.; Wildenhain, P.; Cleveland, B. T.; Elliott, S. R.; Haxton, W.; Wilkerson, J. F.; Suzuki, A.; Suzuki, Y.; Nakahata, M.

2002-07-01

Solution of the solar neutrino problem is significantly depends on the next generation of detectors that can measure the neutrino radiation from the Sun in intermediate energies. An intense (˜ 1 MCi) 37Ar source would be an ideal tool for the calibration of new solar neutrino detectors. The technology of the production of such a source is based on the irradiation of a large mass of a Ca-containing target in a high-flux fast-neutron reactor. Produced 37Ar extracted from this target, will be purified and encapsulated in a source holder. A joint scientific collaboration of Russian, US and Japanese institutions are researching and developing the initial steps of this work and are funded by ISTC and CRDF.

Space radiation studies at the White Sands Missile Range Fast Burst Reactor

NASA Technical Reports Server (NTRS)

Delapaz, A.

1972-01-01

The operation of the White Sands Missile Range Fast Burst Reactor is discussed. Space radiation studies in radiobiology, dosimetry, and transient radiation effects on electronic systems and components are described. Proposed modifications to increase the capability of the facility are discussed.

FAST CHOPPER BUILDING, TRA665. DETAIL OF STEEL DOOR ENTRY TO ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FAST CHOPPER BUILDING, TRA-665. DETAIL OF STEEL DOOR ENTRY TO LOWER LEVEL. CAMERA FACING NORTH. INL NEGATIVE NO. HD42-1. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

DOPPLER CALCULATIONS FOR LARGE FAST CERAMIC REACTORS--EFFECTS OF IMPROVED METHODS AND RECENT CROSS SECTION INFORMATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greebler, P.; Goldman, E.

1962-12-19

Doppler calculations for large fast ceramic reactors (FCR), using recent cross section information and improved methods, are described. Cross sections of U/sup 238/, Pu/sup 239/, and Pu/sup 210/ with fuel temperature variations needed for perturbation calculations of Doppler reactivity changes are tabulated as a function of potential scattering cross section per absorber isotope at energies below 400 kev. These may be used in Doppler calculations for anv fast reactor. Results of Doppler calculations on a large fast ceramic reactor are given to show the effects of the improved calculation methods and of recent cross secrion data on the calculated Dopplermore » coefficient. The updated methods and cross sections used yield a somewhat harder spectrum and accordingly a somewhat smaller Doppler coefficient for a given FCR core size and composition than calculated in earlier work, but they support the essential conclusion derived earlier that the Doppler effect provides an important safety advantage in a large FCR. 28 references. (auth)« less

Individually ventilated cages cause chronic low-grade hypoxia impacting mice hematologically and behaviorally

PubMed Central

York, Jason M.; McDaniel, Allison W.; Blevins, Neil A.; Guillet, Riley R.; Allison, Sarah O.; Cengel, Keith A.; Freund, Gregory G.

2012-01-01

Use of individually ventilated caging (IVC) systems for mouse-based laboratory investigation has dramatically increased. We found that without mice present, intra-cage oxygen concentration was comparable (21%) between IVC housing and ambient environment caging (AEC) that used wire top lids. However, when mice were housed 4-to-a-cage for 1 week, intra-cage oxygen dropped to 20.5% in IVC housing as compared to 21% for AEC housing. IVC intra-cage humidity was also elevated relative to AEC housing. Mice raised in IVC housing as compared to mice raised in AEC housing had higher RBC mass, hematocrit and hemoglobin concentrations. They also had elevated platelet counts but lower white blood cell counts. IVC mice relative to AEC mice had increased saccharin preference and increased fluid consumption but similar locomotion, food intake, social exploration and novel object recognition when tested in an AEC environment. Taken together, these data indicate that ventilated caging systems can have a 0.5% reduction from ambient oxygen concentration that is coupled to mouse red blood cell indices indicative of chronic exposure to a hypoxia. Importantly, IVC housing can impact behavioral testing for depressive-like behavior. PMID:22561683

Impact of Including Higher Actinides in Fast Reactor Transmutation Analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

B. Forget; M. Asgari; R. Ferrer

2007-09-01

Previous fast reactor transmutation studies generally disregarded higher mass minor actinides beyond Cm-246 due to various considerations including deficiencies in nuclear cross-section data. Although omission of these higher mass actinides does not significantly impact the neutronic calculations and fuel cycle performance parameters follow-on neutron dose calculations related to fuel recycling, transportation and handling are significantly impacted. This report shows that including the minor actinides in the equilibrium fast reactor calculations will increase the predicted neutron emission by about 30%. In addition a sensitivity study was initiated by comparing the impact of different cross-section evaluation file for representing these minor actinides.

Influence of fast alpha diffusion and thermal alpha buildup on tokamak reactor performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uckan, N.A.; Tolliver, J.S.; Houlberg, W.A.

1987-11-01

The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor (ETR) plasma (Tokamak Ignition/Burn Experimental Reactor (TIBER-II)) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-D transport models. Estimates are made of the threshold for radial diffusion of fast alphas and thermal alpha buildup. It is shown that a relatively low level of radial transport, when combined with large gradients in the fast alpha density, leads to a significant radial flow with a deleterious effect on plasma performance. Similarly, modest levels of thermal alphamore » concentration significantly influence the ignition and steady-state burn capability. 23 refs., 9 figs., 4 tabs.« less

Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

Front-end Design and Characterization for the ν-Angra Nuclear Reactor Monitoring Detector

NASA Astrophysics Data System (ADS)

Dornelas, T. I.; Araújo, F. T. H.; Cerqueira, A. S.; Costa, J. A.; Nóbrega, R. A.

2016-07-01

The Neutrinos Angra (ν-Angra) Experiment aims to construct an antineutrinos detection device capable of monitoring the Angra dos Reis nuclear reactor activity. Nuclear reactors are intense sources of antineutrinos, and the thermal power released in the fission process is directly related to the flow rate of these particles. The antineutrinos energy spectrum also provides valuable information on the nuclear source isotopic composition. The proposed detector will be equipped with photomultipliers tubes (PMT) which will be readout by a custom Amplifier-Shaper-Discriminator circuit designed to condition its output signals to the acquisition modules to be digitized and processed by an FPGA. The readout circuit should be sensitive to single photoelectron signals, process fast signals, with a full-width-half-amplitude of about 5 ns, have a narrow enough output pulse width to detect both particles coming out from the inverse beta decay (bar nue+p → n + e+), and its output amplitude should be linear to the number of photoelectrons generated inside the PMT, used for energy estimation. In this work, some of the main PMT characteristics are measured and a new readout circuit is proposed, described and characterized.

High mass resolution time of flight mass spectrometer for measuring products in heterogeneous catalysis in highly sensitive microreactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andersen, T.; Jensen, R.; Christensen, M. K.

2012-07-15

We demonstrate a combined microreactor and time of flight system for testing and characterization of heterogeneous catalysts with high resolution mass spectrometry and high sensitivity. Catalyst testing is performed in silicon-based microreactors which have high sensitivity and fast thermal response. Gas analysis is performed with a time of flight mass spectrometer with a modified nude Bayard-Alpert ionization gauge as gas ionization source. The mass resolution of the time of flight mass spectrometer using the ion gauge as ionization source is estimated to m/{Delta}m > 2500. The system design is superior to conventional batch and flow reactors with accompanying product detectionmore » by quadrupole mass spectrometry or gas chromatography not only due to the high sensitivity, fast temperature response, high mass resolution, and fast acquisition time of mass spectra but it also allows wide mass range (0-5000 amu in the current configuration). As a demonstration of the system performance we present data from ammonia oxidation on a Pt thin film showing resolved spectra of OH and NH{sub 3}.« less

High mass resolution time of flight mass spectrometer for measuring products in heterogeneous catalysis in highly sensitive microreactors

NASA Astrophysics Data System (ADS)

Andersen, T.; Jensen, R.; Christensen, M. K.; Pedersen, T.; Hansen, O.; Chorkendorff, I.

2012-07-01

We demonstrate a combined microreactor and time of flight system for testing and characterization of heterogeneous catalysts with high resolution mass spectrometry and high sensitivity. Catalyst testing is performed in silicon-based microreactors which have high sensitivity and fast thermal response. Gas analysis is performed with a time of flight mass spectrometer with a modified nude Bayard-Alpert ionization gauge as gas ionization source. The mass resolution of the time of flight mass spectrometer using the ion gauge as ionization source is estimated to m/Δm > 2500. The system design is superior to conventional batch and flow reactors with accompanying product detection by quadrupole mass spectrometry or gas chromatography not only due to the high sensitivity, fast temperature response, high mass resolution, and fast acquisition time of mass spectra but it also allows wide mass range (0-5000 amu in the current configuration). As a demonstration of the system performance we present data from ammonia oxidation on a Pt thin film showing resolved spectra of OH and NH3.

High mass resolution time of flight mass spectrometer for measuring products in heterogeneous catalysis in highly sensitive microreactors.

PubMed

Andersen, T; Jensen, R; Christensen, M K; Pedersen, T; Hansen, O; Chorkendorff, I

2012-07-01

We demonstrate a combined microreactor and time of flight system for testing and characterization of heterogeneous catalysts with high resolution mass spectrometry and high sensitivity. Catalyst testing is performed in silicon-based microreactors which have high sensitivity and fast thermal response. Gas analysis is performed with a time of flight mass spectrometer with a modified nude Bayard-Alpert ionization gauge as gas ionization source. The mass resolution of the time of flight mass spectrometer using the ion gauge as ionization source is estimated to m/Δm > 2500. The system design is superior to conventional batch and flow reactors with accompanying product detection by quadrupole mass spectrometry or gas chromatography not only due to the high sensitivity, fast temperature response, high mass resolution, and fast acquisition time of mass spectra but it also allows wide mass range (0-5000 amu in the current configuration). As a demonstration of the system performance we present data from ammonia oxidation on a Pt thin film showing resolved spectra of OH and NH(3).

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

The in vitro protection of human decay accelerating factor and hDAF/heme oxygenase-1 transgenes in porcine aortic endothelial cells against sera of Formosan macaques.

PubMed

Tu, C-F; Tai, H-C; Wu, C-P; Ho, L-L; Lin, Y-J; Hwang, C-S; Yang, T-S; Lee, J-M; Tseng, Y-L; Huang, C-C; Weng, C-N; Lee, P-H

2010-01-01

To mitigate hyperacute rejection, pigs have been generated with alpha-Gal transferase gene knockout and transgenic expression of human decay accelerating factor (hDAF), MCP, and CD59. Additionally, heme-oxygenase-1 (HO-1) has been suggested to defend endothelial cells. Sera (MS) (0%, 1%, 5%, 10%, and 15%) from Formosan macaques (Macaca cyclopis, MC), an Old World monkey wildly populated in Taiwan, was used to test the protective in vitro, effects of hDAF or hDAF/hHO-1 on porcine aortic endothelial cells (pAEC) derived from hDAF(+), hDAF(+)/hHO-1(+), and hDAF(+)/hHO-1(-) and 1 nontransgenic pAEC. Ten percent human serum (HS) served as a positive control. When MS addition increased to 10% or 15%, all transgenic pAEC exhibited a greater survival than nontransgenic pAEC. Noticeably, 15% MS reduced survived to <10% versus >40% in nontransgenic and transgenic pAEC, respectively. These results revealed that hDAF exerted protective effects against MC complement activation. However, comparing with 10% MS and HS in pAEC of nontransgenic pigs, the survivability was higher in HS, suggesting that complement activation by MS was more toxic than that by HS. Furthermore, hDAF(+)/hHO-1(+) showed no further protection against effects of MS on transgenic pAEC. Copyright 2010 Elsevier Inc. All rights reserved.

KINETICS OF TREAT USED AS A TEST REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickerman, C.E.; Johnson, R.D.; Gasidlo, J.

1962-05-01

An analysis is presented concerning the reactor kinetics of TREAT used as a pulsed, engineering test reactor for fast reactor fuel element studies. A description of the reactor performance is given for a wide range of conditions associated with its use as a test reactor. Supplemental information on meltdown experimentation is included. (J.R.D.)

Impact of Nuclear Data Uncertainties on Advanced Fuel Cycles and their Irradiated Fuel - a Comparison between Libraries

NASA Astrophysics Data System (ADS)

Díez, C. J.; Cabellos, O.; Martínez, J. S.

2014-04-01

The uncertainties on the isotopic composition throughout the burnup due to the nuclear data uncertainties are analysed. The different sources of uncertainties: decay data, fission yield and cross sections; are propagated individually, and their effect assessed. Two applications are studied: EFIT (an ADS-like reactor) and ESFR (Sodium Fast Reactor). The impact of the uncertainties on cross sections provided by the EAF-2010, SCALE6.1 and COMMARA-2.0 libraries are compared. These Uncertainty Quantification (UQ) studies have been carried out with a Monte Carlo sampling approach implemented in the depletion/activation code ACAB. Such implementation has been improved to overcome depletion/activation problems with variations of the neutron spectrum.

Ex-vessel neutron dosimetry analysis for westinghouse 4-loop XL pressurized water reactor plant using the RadTrack{sup TM} Code System with the 3D parallel discrete ordinates code RAPTOR-M3G

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, J.; Alpan, F. A.; Fischer, G.A.

2011-07-01

Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the belt-line region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locationsmore » and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure. (authors)« less

STABILIZED RARE EARTH OXIDES FOR A CONTROL ROD AND METHOD OF PREPARATION

DOEpatents

McNees, R.A.; Potter, R.A.

1964-01-14

A method is given for preparing mixed oxides of the formula MR/sub x/O/ sub 12/ wherein M is tungsten or molybdenum and R is a rare earth in the group consisting of samarium, europium, dysprosium, and gadolinium and x is 4 to 5. Oxides of this formula, and particularly the europiumcontaining species, are useful as control rod material for water-cooled nuclear reactors owing to their stability, favorable nuclear properties, and resistance to hydration. These oxides may be utilized as a dispersion in a stainlesssteel matrix. Preparation of these oxides is effected by blending tungsten oxide or molybdenum oxide with a rare earth oxide, compressing the mixture, and firing at an elevated temperature in an oxygen-containing atmosphere. (AEC)

Determination of efficiency of an aged HPGe detector for gaseous sources by self absorption correction and point source methods

NASA Astrophysics Data System (ADS)

Sarangapani, R.; Jose, M. T.; Srinivasan, T. K.; Venkatraman, B.

2017-07-01

Methods for the determination of efficiency of an aged high purity germanium (HPGe) detector for gaseous sources have been presented in the paper. X-ray radiography of the detector has been performed to get detector dimensions for computational purposes. The dead layer thickness of HPGe detector has been ascertained from experiments and Monte Carlo computations. Experimental work with standard point and liquid sources in several cylindrical geometries has been undertaken for obtaining energy dependant efficiency. Monte Carlo simulations have been performed for computing efficiencies for point, liquid and gaseous sources. Self absorption correction factors have been obtained using mathematical equations for volume sources and MCNP simulations. Self-absorption correction and point source methods have been used to estimate the efficiency for gaseous sources. The efficiencies determined from the present work have been used to estimate activity of cover gas sample of a fast reactor.

Production of ZrC Matrix for Use in Gas Fast Reactor Composite Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vasudevamurthy, Gokul; Knight, Travis W.; Roberts, Elwyn

2007-07-01

Zirconium carbide is being considered as a candidate for inert matrix material in composite nuclear fuel for Gas fast reactors due to its favorable characteristics. ZrC can be produced by the direct reaction of pure zirconium and graphite powders. Such a reaction is exothermic in nature. The reaction is self sustaining once initial ignition has been achieved. The heat released during the reaction is high enough to complete the reaction and achieve partial sintering without any external pressure applied. External heat source is required to achieve ignition of the reactants and maintain the temperature close to the adiabatic temperature tomore » achieve higher levels of sintering. External pressure is also a driving force for sintering. In the experiments described, cylindrical compacts of ZrC were produced by direct combustion reaction. External induction heating combined with varying amounts of external applied pressure was employed to achieve varying degrees of density/porosity. The effect of reactant particle size on the product characteristics was also studied. The samples were characterized for density/porosity, composition and microstructure. (authors)« less

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

Bio-oil production from palm fronds by fast pyrolysis process in fluidized bed reactor

NASA Astrophysics Data System (ADS)

Rinaldi, Nino; Simanungkalit, Sabar P.; Kiky Corneliasari, S.

2017-01-01

Fast pyrolysis process of palm fronds has been conducted in the fluidized bed reactor to yield bio-oil product (pyrolysis oil). The process employed sea sand as the heat transfer medium. The objective of this study is to design of the fluidized bed rector, to conduct fast pyrolysis process to product bio-oil from palm fronds, and to characterize the feed and bio-oil product. The fast pyrolysis process was conducted continuously with the feeding rate around 500 g/hr. It was found that the biomass conversion is about 35.5% to yield bio-oil, however this conversion is still minor. It is suggested due to the heating system inside the reactor was not enough to decompose the palm fronds as a feedstock. Moreover, the acids compounds ware mostly observed on the bio-oil product.

Effects of AEC chamber selection on patient dose and image quality.

PubMed

Hawking, Nancy; Elmore, Angie

2009-01-01

To determine whether manipulation of the standard automatic exposure control (AEC) chamber selections reduces the patient's entrance skin exposure (ESE) without compromising image quality. Data for density and radiation dose were gathered at 2 clinical locations by exposing abdomen and pelvis phantoms to radiation using 3 AEC chamber selection configurations. ESE (skin dose) was measured using a multipurpose dosimeter. The experiment included both film-screen and computed radiography (CR) systems. For both phantoms, using the 2 outside chambers resulted in the lowest dose on the film-screen and CR systems. In general, optical density (OD) and exposure indicator (EI) remained within acceptable ranges and image quality was maintained using this chamber configuration. Using only the center chamber resulted in the highest dose increases and lowest image quality for film-screen and CR systems. When performing anteroposterior (AP) abdomen and AP pelvis examinations, radiographers can reduce patients' ESE and maintain image quality by selecting the 2 outside AEC chambers. Further research on AEC chamber selection should be conducted for additional anatomical regions.

Novel variant in the TP63 gene associated to ankyloblepharon-ectodermal dysplasia-cleft lip/palate (AEC) syndrome.

PubMed

Gonzalez, Francisco; Loidi, Lourdes; Abalo-Lojo, Jose M

2017-01-01

Ankyloblepharon-ectodermal dysplasia-cleft lip/palate (AEC) syndrome is a disorder resulting from anomalous embryonic development of ectodermal tissues. There is evidence that AEC syndrome is caused by mutations in the TP63 gene, which encodes the p63 protein. This is an important regulatory protein involved in epidermal proliferation and differentiation. Genome sequencing was performed in DNA from peripheral blood leukocytes of a newborn with AEC syndrome and her parents. Variants were searched in all coding exons and intron-exon boundaries of the TP63 gene. A heterozygous missense variant (NM_003722.4:c.1063G>C (p.Asp355His) was found in the newborn patient. No variants were found in either of the parents. We identified a previously unreported variant in TP63 gene which seems to be involved in the somatic malformations found in the AEC syndrome. The absence of this variant in both parents suggests that the variant appeared de novo.

Human amniotic epithelial cell feeder layers maintain mouse embryonic stem cell pluripotency via epigenetic regulation of the c-Myc promoter.

PubMed

Liu, Te; Cheng, Weiwei; Liu, Tianjin; Guo, Lihe; Huang, Qin; Jiang, Lizhen; Du, Xiling; Xu, Fuhui; Liu, Zhixue; Lai, Dongmei

2010-02-01

Mouse embryonic stem cells (ESCs) are typically cultured on a feeder layer of mouse embryonic fibroblasts (MEFs), with leukemia inhibitory factor (LIF) added to maintain them in an undifferentiated state. We have previously shown that human amniotic epithelial cells (hAECs) can be used as feeder cells to maintain mouse ESC pluripotency, but the mechanism for this is unknown. In the present study, we found that CpG islands 5' of the c-Myc gene remain hypomethylated in mouse ESCs cultured on hAECs. In addition, levels of acetylation of histone H3 and trimethylation of histone H3K4 in the c-Myc gene promoter were higher in ES cells cultured on hAECs than those in ES cells cultured on MEFs. These data suggested that hAECs can alter mouse ESC gene expression via epigenetic modification of c-Myc, providing a possible mechanism for the hAEC-induced maintenance of ESCs in an undifferentiated state.

Fast-quench reactor for hydrogen and elemental carbon production from natural gas and other hydrocarbons

DOEpatents

Detering, Brent A.; Kong, Peter C.

2006-08-29

A fast-quench reactor for production of diatomic hydrogen and unsaturated carbons is provided. During the fast quench in the downstream diverging section of the nozzle, such as in a free expansion chamber, the unsaturated hydrocarbons are further decomposed by reheating the reactor gases. More diatomic hydrogen is produced, along with elemental carbon. Other gas may be added at different stages in the process to form a desired end product and prevent back reactions. The product is a substantially clean-burning hydrogen fuel that leaves no greenhouse gas emissions, and elemental carbon that may be used in powder form as a commodity for several processes.

Conference Report: International Research Symposium on Ankyloblepharon-Ectodermal Defects-Cleft Lip and/or Palate (AEC) Syndrome

PubMed Central

Fete, Mary; vanBokhoven, Hans; Clements, Suzanne; McKeon, Frank; Roop, Dennis R.; Koster, Maranke I.; Missero, Caterina; Attardi, Laura D.; Lombillo, Vivian A.; Ratovitski, Edward; Julapalli, Meena; Ruths, Derek; Sybert, Virginia P.; Siegfried, Elaine C.; Bree, Alanna F.

2009-01-01

Ankyloblepharon-Ectodermal Defects-Cleft Lip/Palate (AEC) Syndrome (Hay-Wells syndrome, MIM #106220) is a rare autosomal dominant ectodermal dysplasia syndrome. It is due to mutations in the p63 gene, known to be a regulatory gene with many downstream gene targets. TP63 is important in the differentiation and proliferation of the epidermis, as well as many other processes including limb and facial development. It is also known that mutations in p63 lead to skin erosions. These erosions, especially on the scalp, are defining features of AEC syndrome and cause significant morbidity and mortality in these patients. It was this fact that led to the 2003 AEC Skin Erosion Workshop. That conference laid the groundwork for the International Research Symposium for AEC Syndrome held at Texas Children's Hospital in 2006. The conference brought together the largest cohort of individuals with AEC syndrome, along with a multitude of physicians and scientists. The overarching goals were to define the clinical and pathologic findings for improved diagnostic criteria, to obtain tissue samples for further study and to define future research directions. The symposium was successful in accomplishing these aims as detailed in this conference report. Following our report, we also present eleven manuscripts within this special section that outline the collective clinical, pathologic and mutational data from eighteen individuals enrolled in the concurrent Baylor College of Medicine IRB-approved protocol: Characterization of AEC syndrome. These collaborative findings will hopefully provide a stepping stone to future translational projects of p63 and p63-related syndromes. PMID:19353643

Configuration of automatic exposure control on mammography units for computed radiography to match patient dose of screen film systems

NASA Astrophysics Data System (ADS)

Yang, Chang-Ying Joseph; Huang, Weidong

2009-02-01

Computed radiography (CR) is considered a drop-in addition or replacement for traditional screen-film (SF) systems in digital mammography. Unlike other technologies, CR has the advantage of being compatible with existing mammography units. One of the challenges, however, is to properly configure the automatic exposure control (AEC) on existing mammography units for CR use. Unlike analogue systems, the capture and display of digital CR images is decoupled. The function of AEC is changed from ensuring proper and consistent optical density of the captured image on film to balancing image quality with patient dose needed for CR. One of the preferences when acquiring CR images under AEC is to use the same patient dose as SF systems. The challenge is whether the existing AEC design and calibration process-most of them proprietary from the X-ray systems manufacturers and tailored specifically for SF response properties-can be adapted for CR cassettes, in order to compensate for their response and attenuation differences. This paper describes the methods for configuring the AEC of three different mammography units models to match the patient dose used for CR with those that are used for a KODAK MIN-R 2000 SF System. Based on phantom test results, these methods provide the dose level under AEC for the CR systems to match with the dose of SF systems. These methods can be used in clinical environments that require the acquisition of CR images under AEC at the same dose levels as those used for SF systems.

Involvement of Alveolar Epithelial Cell Necroptosis in IPF Pathogenesis.

PubMed

Lee, Ji-Min; Yoshida, Masahiro; Kim, Mi-So; Lee, June-Hyuk; Baek, Ae-Rin; Jang, An Soo; Kim, Do Jin; Minagawa, Shunsuke; Chin, Su Sie; Park, Choon-Sik; Araya, Jun; Kuwano, Kazuyoshi; Park, Sung Woo

2018-02-14

Alveolar epithelial cell (AEC) injury leading to cell death is involved in the process of fibrosis development during idiopathic pulmonary fibrosis (IPF). Among regulated/programmed cell death, the excessive apoptosis of AECs has been widely implicated in IPF pathogenesis. Necroptosis is a type of regulated/programmed necrosis. A multiprotein complex composed of receptor-interacting protein kinase-1 and -3 (RIPK1/3) plays a key regulatory role in initiating necroptosis. Although necroptosis participates in disease pathogeneses through the release of damage-associated molecular patterns (DAMPs), its association with IPF progression remains elusive. In this study, we attempted to illuminate the involvement of RIPK3-regulated necroptosis in IPF pathogenesis. IPF lung tissues were used to detect necroptosis, and the role of RIPK3 was determined using cell culturing models of AECs. Lung fibrosis models of bleomycin (BLM) treatment were also used. RIPK3 expression levels were increased in IPF lungs and both apoptosis and necroptosis were detected mainly in AECs. Necrostatin-1 and RIPK3 knockdown experiments in AECs revealed the participation of necroptosis in BLM and hydrogen peroxide-induced cell death. BLM treatment induced RIPK3 expression in AECs and increased High Mobility Group Box 1 (HMGB1) and interleukin 1β (IL-1β) levels in mouse lungs. The efficient attenuation of BLM-induced lung inflammation and fibrosis was determined in RIPK3 knockout mice and by necrostatin-1 with a concomitant reduction in HMGB1 and IL-1β. RIPK3-regulated necroptosis in AECs is involved in the mechanism of lung fibrosis development through the release of DAMPs as part of the pathogenic sequence of IPF.

A20 Deficiency in Lung Epithelial Cells Protects against Influenza A Virus Infection

PubMed Central

Vereecke, Lars; Mc Guire, Conor; Sze, Mozes; Schuijs, Martijn J.; Willart, Monique; Itati Ibañez, Lorena; Hammad, Hamida; Lambrecht, Bart N.; Beyaert, Rudi; Saelens, Xavier; van Loo, Geert

2016-01-01

A20 negatively regulates multiple inflammatory signalling pathways. We here addressed the role of A20 in club cells (also known as Clara cells) of the bronchial epithelium in their response to influenza A virus infection. Club cells provide a niche for influenza virus replication, but little is known about the functions of these cells in antiviral immunity. Using airway epithelial cell-specific A20 knockout (A20AEC-KO) mice, we show that A20 in club cells critically controls innate immune responses upon TNF or double stranded RNA stimulation. Surprisingly, A20AEC-KO mice are better protected against influenza A virus challenge than their wild type littermates. This phenotype is not due to decreased viral replication. Instead host innate and adaptive immune responses and lung damage are reduced in A20AEC-KO mice. These attenuated responses correlate with a dampened cytotoxic T cell (CTL) response at later stages during infection, indicating that A20AEC-KO mice are better equipped to tolerate Influenza A virus infection. Expression of the chemokine CCL2 (also named MCP-1) is particularly suppressed in the lungs of A20AEC-KO mice during later stages of infection. When A20AEC-KO mice were treated with recombinant CCL2 the protective effect was abrogated demonstrating the crucial contribution of this chemokine to the protection of A20AEC-KO mice to Influenza A virus infection. Taken together, we propose a mechanism of action by which A20 expression in club cells controls inflammation and antiviral CTL responses in response to influenza virus infection. PMID:26815999

Low-dose computed tomography scans with automatic exposure control for patients of different ages undergoing cardiac PET/CT and SPECT/CT.

PubMed

Yang, Ching-Ching; Yang, Bang-Hung; Tu, Chun-Yuan; Wu, Tung-Hsin; Liu, Shu-Hsin

2017-06-01

This study aimed to evaluate the efficacy of automatic exposure control (AEC) in order to optimize low-dose computed tomography (CT) protocols for patients of different ages undergoing cardiac PET/CT and single-photon emission computed tomography/computed tomography (SPECT/CT). One PET/CT and one SPECT/CT were used to acquire CT images for four anthropomorphic phantoms representative of 1-year-old, 5-year-old and 10-year-old children and an adult. For the hybrid systems investigated in this study, the radiation dose and image quality of cardiac CT scans performed with AEC activated depend mainly on the selection of a predefined image quality index. Multiple linear regression methods were used to analyse image data from anthropomorphic phantom studies to investigate the effects of body size and predefined image quality index on CT radiation dose in cardiac PET/CT and SPECT/CT scans. The regression relationships have a coefficient of determination larger than 0.9, indicating a good fit to the data. According to the regression models, low-dose protocols using the AEC technique were optimized for patients of different ages. In comparison with the standard protocol with AEC activated for adult cardiac examinations used in our clinical routine practice, the optimized paediatric protocols in PET/CT allow 32.2, 63.7 and 79.2% CT dose reductions for anthropomorphic phantoms simulating 10-year-old, 5-year-old and 1-year-old children, respectively. The corresponding results for cardiac SPECT/CT are 8.4, 51.5 and 72.7%. AEC is a practical way to reduce CT radiation dose in cardiac PET/CT and SPECT/CT, but the AEC settings should be determined properly for optimal effect. Our results show that AEC does not eliminate the need for paediatric protocols and CT examinations using the AEC technique should be optimized for paediatric patients to reduce the radiation dose as low as reasonably achievable.

Transversal inducing differentiation of human amniotic epithelial cells into hepatocyte-like cells.

PubMed

Luo, Hongwu; Huang, Xiangjun; Huang, Feizhou; Liu, Xunyang

2011-06-01

To evaluate the in vitro differentiation of human amniotic epithelial cells (hAECs ) into hepatocyte-like cells. Combined approach of dexamethasone, HGF, IGF and other cytokines were used to induce the differentiation of hAECs into hepatocyte-like cells. The induction lasted 2 weeks. During the induction, the expression of albumin ALB, CYP1A1, CYP1A2, IGFR, c-met and key functional genes related to liver cells as well as transcription factors HNF3, HNF4 and C/EBPa were monitored by RT-PCR. Time dependent changes of the surface marker colony ALB, AFP and CK18 were analyzed by cell flow cytometry. After the 2 week induction, the expressions of liver hepatocyte-like cell functional genes such as albumin, CYP1A1, CYP1A2, c-met, and transcription factors such as HNF3, HNF4, C/EBPa and HNF1 were observed. Six days after the induction, hAECs mainly were stained AFP+, and the positive rate was (15.1 ± 2.1)%. While 10 days after the induction, part of the hAECs showed AFP+/ALB+ (6.5 ± 1.4)%; and on 14th day, hAECs only showed ALB+, and the rate was (13.9 ± 2.3)%. ALB+ cell increase indicated a gradual functional maturation from the hAECs to hepatocyte-like cells. Similaritly, the number of CK18+ cells in the whole population was also increased: On 10th day, the rate was (16.1 ± 1.2)%; on 14th day, that was (21.3 ± 4.6)%, which proved the above hypothesis of the trandifferentiation. By extending the induction time, the expression of functional genes increased gradually, and a maturing process of hAECs was detected by cell surface markers. The differentiation of hAECs induced in vitro has the characteristics of hepatocyte-like cells.

Fast quench reactor method

DOEpatents

Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.; Berry, R.A.

1999-08-10

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

Fast quench reactor method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.; Berry, Ray A.

1999-01-01

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

CHARACTERISTICS OF A FAST RISE TIME POWER SUPPLY FOR A PULSED PLASMA REACTOR FOR CHEMICAL VAPOR DESTRUCTION

EPA Science Inventory

Rotating spark gap devices for switching high-voltage direct current (dc) into a corona plasma reactor can achieve pulse rise times in the range of tens of nanoseconds. The fast rise times lead to vigorous plasma generation without sparking at instantaneous applied voltages highe...

The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, R.R.

1986-01-01

This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Furuta, H.; Imura, A.; Furuta, Y.

Recently, technique of Gadolinium loaded liquid scintillator (Gd-LS) for reactor neutrino oscillation experiments has attracted attention as a monitor of reactor operation and 'nuclear Gain (GA)' for IAEA safeguards. For the practical use, R and D of the 1 ton class compact detector, which is measurable above ground, is necessary. Especially, it is important to reduce much amount of fast neutron background induced by cosmic muons with data analysis for the measurement above ground. We developed a prototype of the Gd-LS detector with 200 L of the target volume, which has Pulse Shape Discrimination (PSD) ability for the fast neutronmore » reduction with data analysis. Usually, it is well known that it is difficult to keep high fast neutron reduction power of PSD with the large volume size such as the neutrino reactor monitor. We evaluated the PSD ability of our prototype with real fast neutrons induced by the muons in our laboratory above ground, and we could confirm to keep the high fast neutron reduction power with even our large detector size. (authors)« less

DE-NE0008277_PROTEUS final technical report 2018

DOE Office of Scientific and Technical Information (OSTI.GOV)

Enqvist, Andreas

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce suchmore » a comprehensive suite of experimental data to support any future GCFR endeavors.« less

Non-destructive diagnostics of irradiated materials using neutron scattering from pulsed neutron sources

NASA Astrophysics Data System (ADS)

Korenev, Sergey; Sikolenko, Vadim

2004-09-01

The advantage of neutron-scattering studies as compared to the standard X-ray technique is the high penetration of neutrons that allow us to study volume effects. The high resolution of instrumentation on the basis neutron scattering allows measurement of the parameters of lattice structure with high precision. We suggest the use of neutron scattering from pulsed neutron sources for analysis of materials irradiated with pulsed high current electron and ion beams. The results of preliminary tests using this method for Ni foils that have been studied by neutron diffraction at the IBR-2 (Pulsed Fast Reactor at Joint Institute for Nuclear Research) are presented.

Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stoller, R.E.

1987-12-01

The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and alteredmore » mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs.« less

NEUTRONIC REACTOR DESIGN TO REDUCE NEUTRON LOSS

DOEpatents

Mills, F.T.

1961-05-01

A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall which is surrounded by successive layers of pure fertile material and fertile material having moderator. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. As the steel has a smaller capture cross-section for the fast neutrons, then greater numbers of the neutrons will pass into the blanket thereby increasing the over-all efficiency of the reactor.

Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

NASA Astrophysics Data System (ADS)

Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

2015-12-01

U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

Thermionic Energy

NASA Technical Reports Server (NTRS)

1985-01-01

Thermionic energy conversion is the production of energy from a nuclear source. It is a technology advanced by SNSO, a joint research and development organization formed by NASA and the AEC. SNSO contracted with Thermo Electron Corporation to develop high temperature applications, i.e., metals with high melting points. Thermo Electron Corporation's expertise resulted in contracts for products made from exotic metals such as bone implants, artificial hips, and heart pacemakers.

ELECTRIC DISCHARGE APPARATUS

DOEpatents

Ryan, T.M.

1962-04-01

A steel or aluminum small diameter (1/4 in.) tube-type neutron detector containing an inert atmosphere and having a coating of fissionable material on its inner circumference is described. A conducting wire, positioned along the axis of the tube by spaced insulators, is connected to a power source. The coating of fissionable material is brushed onto a nickel foil which is inserted into the tube and supported between the insulators. (AEC)

Fuel Cycle System Analysis Handbook

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steven J. Piet; Brent W. Dixon; Dirk Gombert

2009-06-01

This Handbook aims to improve understanding and communication regarding nuclear fuel cycle options. It is intended to assist DOE, Campaign Managers, and other presenters prepare presentations and reports. When looking for information, check here. The Handbook generally includes few details of how calculations were performed, which can be found by consulting references provided to the reader. The Handbook emphasizes results in the form of graphics and diagrams, with only enough text to explain the graphic, to ensure that the messages associated with the graphic is clear, and to explain key assumptions and methods that cause the graphed results. Some ofmore » the material is new and is not found in previous reports, for example: (1) Section 3 has system-level mass flow diagrams for 0-tier (once-through), 1-tier (UOX to CR=0.50 fast reactor), and 2-tier (UOX to MOX-Pu to CR=0.50 fast reactor) scenarios - at both static and dynamic equilibrium. (2) To help inform fast reactor transuranic (TRU) conversion ratio and uranium supply behavior, section 5 provides the sustainable fast reactor growth rate as a function of TRU conversion ratio. (3) To help clarify the difference in recycling Pu, NpPu, NpPuAm, and all-TRU, section 5 provides mass fraction, gamma, and neutron emission for those four cases for MOX, heterogeneous LWR IMF (assemblies mixing IMF and UOX pins), and a CR=0.50 fast reactor. There are data for the first 10 LWR recycle passes and equilibrium. (4) Section 6 provides information on the cycle length, planned and unplanned outages, and TRU enrichment as a function of fast reactor TRU conversion ratio, as well as the dilution of TRU feedstock by uranium in making fast reactor fuel. (The recovered uranium is considered to be more pure than recovered TRU.) The latter parameter impacts the required TRU impurity limits specified by the Fuels Campaign. (5) Section 7 provides flows for an 800-tonne UOX separation plant. (6) To complement 'tornado' economic uncertainty diagrams, which show at a glance combined uncertainty information, section 9.2 has a new set of simpler graphs that show the impact on fuel cycle costs for once through, 1-tier, and 2-tier scenarios as a function of key input parameters.« less

The slow and fast pyrolysis of cherry seed.

PubMed

Duman, Gozde; Okutucu, Cagdas; Ucar, Suat; Stahl, Ralph; Yanik, Jale

2011-01-01

The slow and fast pyrolysis of cherry seeds (CWS) and cherry seeds shells (CSS) was studied in fixed-bed and fluidized bed reactors at different pyrolysis temperatures. The effects of reactor type and temperature on the yields and composition of products were investigated. In the case of fast pyrolysis, the maximum bio-oil yield was found to be about 44 wt% at pyrolysis temperature of 500 °C for both CWS and CSS, whereas the bio yields were of 21 and 15 wt% obtained at 500 °C from slow pyrolysis of CWS and CSS, respectively. Both temperature and reactor type affected the composition of bio-oils. The results showed that bio-oils obtained from slow pyrolysis of CWS and CSS can be used as a fuel for combustion systems in industry and the bio-oil produced from fast pyrolysis can be evaluated as a chemical feedstock. Copyright © 2010 Elsevier Ltd. All rights reserved.

The ASEAN economic community and medical qualification

PubMed Central

Kittrakulrat, Jathurong; Jongjatuporn, Witthawin; Jurjai, Ravipol; Jarupanich, Nicha; Pongpirul, Krit

2014-01-01

Background In the regional movement toward ASEAN Economic Community (AEC), medical professions including physicians can be qualified to practice medicine in another country. Ensuring comparable, excellent medical qualification systems is crucial but the availability and analysis of relevant information has been lacking. Objective This study had the following aims: 1) to comparatively analyze information on Medical Licensing Examinations (MLE) across ASEAN countries and 2) to assess stakeholders’ view on potential consequences of AEC on the medical profession from a Thai perspective. Design To search for relevant information on MLE, we started with each country's national body as the primary data source. In case of lack of available data, secondary data sources including official websites of medical universities, colleagues in international and national medical student organizations, and some other appropriate Internet sources were used. Feasibility and concerns about validity and reliability of these sources were discussed among investigators. Experts in the region invited through HealthSpace.Asia conducted the final data validation. For the second objective, in-depth interviews were conducted with 13 Thai stakeholders, purposely selected based on a maximum variation sampling technique to represent the points of view of the medical licensing authority, the medical profession, ethicists and economists. Results MLE systems exist in all ASEAN countries except Brunei, but vary greatly. Although the majority has a national MLE system, Singapore, Indonesia, and Vietnam accept results of MLE conducted at universities. Thailand adopted the USA's 3-step approach that aims to check pre-clinical knowledge, clinical knowledge, and clinical skills. Most countries, however, require only one step. A multiple choice question (MCQ) is the most commonly used method of assessment; a modified essay question (MEQ) is the next most common. Although both tests assess candidate's knowledge, the Objective Structured Clinical Examination (OSCE) is used to verify clinical skills of the examinee. The validity of the medical license and that it reflects a consistent and high standard of medical knowledge is a sensitive issue because of potentially unfair movement of physicians and an embedded sense of domination, at least from a Thai perspective. Conclusions MLE systems differ across ASEAN countries in some important aspects that might be of concern from a fairness viewpoint and therefore should be addressed in the movement toward AEC. PMID:25215908

Importance of the (n,gamma) Cm-247 Evaluation on Neutron Emission in Fast Reactor Fuel Cycle Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benoit Forget; Mehdi Asgari; Rodolfo M. Ferrer

2007-11-01

As part of the GNEP program, it is envisioned to build a fast reactor for the transmutation of minor actinides. The spent nuclear fuel from the current fleet of light water reactors would be recycled, the current baseline is the UREX+1a process, and would act as a feed for the fast reactor. As the fuel is irradiated in a fast reactor a certain quantity of minor actinides would thus build up in the fuel stream creating possible concerns with the neutron emission of these minor actinides for fuel transportation, handling and fabrication. Past neutronic analyses had not tracked minor actinidesmore » above Cm-246 in the transmutation chain, because of the small influence on the overall reactor performance and cycle parameters. However, when trying to quantify the neutron emission from the recycled fuel with high minor actinide content, these higher isotopes play an essential role and should be included in the analysis. In this paper, the influence of tracking these minor actinides on the calculated neutron emission is presented. Also presented is the particular influence of choosing a different evaluated cross section data set to represent the minor actinides above Cm-246. The first representation uses the cross-sections provided by MC2-2 for all isotopes, while the second representation uses infinitely diluted ENDF/BVII.0 cross-sections for Cm-247 to Cf-252 and MC2-2 for all other isotopes.« less

Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

NASA Astrophysics Data System (ADS)

Dutta, N. G.

2012-11-01

Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

Radial blanket assembly orificing arrangement

DOEpatents

Patterson, J.F.

1975-07-01

A nuclear reactor core for a liquid metal cooled fast breeder reactor is described in which means are provided for increasing the coolant flow through the reactor fuel assemblies as the reactor ages by varying the coolant flow rate with the changing coolant requirements during the core operating lifetime. (auth)

Mitochondrial catalase overexpressed transgenic mice are protected against lung fibrosis in part via preventing alveolar epithelial cell mitochondrial DNA damage.

PubMed

Kim, Seok-Jo; Cheresh, Paul; Jablonski, Renea P; Morales-Nebreda, Luisa; Cheng, Yuan; Hogan, Erin; Yeldandi, Anjana; Chi, Monica; Piseaux, Raul; Ridge, Karen; Michael Hart, C; Chandel, Navdeep; Scott Budinger, G R; Kamp, David W

2016-12-01

Alveolar epithelial cell (AEC) injury and mitochondrial dysfunction are important in the development of lung fibrosis. Our group has shown that in the asbestos exposed lung, the generation of mitochondrial reactive oxygen species (ROS) in AEC mediate mitochondrial DNA (mtDNA) damage and apoptosis which are necessary for lung fibrosis. These data suggest that mitochondrial-targeted antioxidants should ameliorate asbestos-induced lung. To determine whether transgenic mice that express mitochondrial-targeted catalase (MCAT) have reduced lung fibrosis following exposure to asbestos or bleomycin and, if so, whether this occurs in association with reduced AEC mtDNA damage and apoptosis. Crocidolite asbestos (100µg/50µL), TiO 2 (negative control), bleomycin (0.025 units/50µL), or PBS was instilled intratracheally in 8-10 week-old wild-type (WT - C57Bl/6J) or MCAT mice. The lungs were harvested at 21d. Lung fibrosis was quantified by collagen levels (Sircol) and lung fibrosis scores. AEC apoptosis was assessed by cleaved caspase-3 (CC-3)/Surfactant protein C (SFTPC) immunohistochemistry (IHC) and semi-quantitative analysis. AEC (primary AT2 cells from WT and MCAT mice and MLE-12 cells) mtDNA damage was assessed by a quantitative PCR-based assay, apoptosis was assessed by DNA fragmentation, and ROS production was assessed by a Mito-Sox assay. Compared to WT, crocidolite-exposed MCAT mice exhibit reduced pulmonary fibrosis as measured by lung collagen levels and lung fibrosis score. The protective effects in MCAT mice were accompanied by reduced AEC mtDNA damage and apoptosis. Similar findings were noted following bleomycin exposure. Euk-134, a mitochondrial SOD/catalase mimetic, attenuated MLE-12 cell DNA damage and apoptosis. Finally, compared to WT, asbestos-induced MCAT AT2 cell ROS production was reduced. Our finding that MCAT mice have reduced pulmonary fibrosis, AEC mtDNA damage and apoptosis following exposure to asbestos or bleomycin suggests an important role for AEC mitochondrial H 2 O 2 -induced mtDNA damage in promoting lung fibrosis. We reason that strategies aimed at limiting AEC mtDNA damage arising from excess mitochondrial H 2 O 2 production may be a novel therapeutic target for mitigating pulmonary fibrosis. Published by Elsevier Inc.

Mitochondrial catalase overexpressed transgenic mice are protected against lung fibrosis in part via preventing alveolar epithelial cell mitochondrial DNA damage

PubMed Central

Kim, Seok-Jo; Cheresh, Paul; Jablonski, Renea P.; Morales-Nebreda, Luisa; Cheng, Yuan; Hogan, Erin; Yeldandi, Anjana; Chi, Monica; Piseaux, Raul; Ridge, Karen; Hart, C. Michael; Chandel, Navdeep; Budinger, G.R. Scott; Kamp, David W.

2018-01-01

Rationale Alveolar epithelial cell (AEC) injury and mitochondrial dysfunction are important in the development of lung fibrosis. Our group has shown that in the asbestos exposed lung, the generation of mitochondrial reactive oxygen species (ROS) in AEC mediate mitochondrial DNA (mtDNA) damage and apoptosis which are necessary for lung fibrosis. These data suggest that mitochondrial-targeted antioxidants should ameliorate asbestos-induced lung. Objective To determine whether transgenic mice that express mitochondrial-targeted catalase (MCAT) have reduced lung fibrosis following exposure to asbestos or bleomycin and, if so, whether this occurs in association with reduced AEC mtDNA damage and apoptosis. Methods Crocidolite asbestos (100 μg/50 μL), TiO2 (negative control), bleomycin (0.025 units/50 μL), or PBS was instilled intratracheally in 8–10 week-old wild-type (WT - C57Bl/6 J) or MCAT mice. The lungs were harvested at 21 d. Lung fibrosis was quantified by collagen levels (Sircol) and lung fibrosis scores. AEC apoptosis was assessed by cleaved caspase-3 (CC-3)/Surfactant protein C (SFTPC) immunohistochemistry (IHC) and semi-quantitative analysis. AEC (primary AT2 cells from WT and MCAT mice and MLE-12 cells) mtDNA damage was assessed by a quantitative PCR-based assay, apoptosis was assessed by DNA fragmentation, and ROS production was assessed by a Mito-Sox assay. Results Compared to WT, crocidolite-exposed MCAT mice exhibit reduced pulmonary fibrosis as measured by lung collagen levels and lung fibrosis score. The protective effects in MCAT mice were accompanied by reduced AEC mtDNA damage and apoptosis. Similar findings were noted following bleomycin exposure. Euk-134, a mitochondrial SOD/catalase mimetic, attenuated MLE-12 cell DNA damage and apoptosis. Finally, compared to WT, asbestos-induced MCAT AT2 cell ROS production was reduced. Conclusions Our finding that MCAT mice have reduced pulmonary fibrosis, AEC mtDNA damage and apoptosis following exposure to asbestos or bleomycin suggests an important role for AEC mitochondrial H2O2-induced mtDNA damage in promoting lung fibrosis. We reason that strategies aimed at limiting AEC mtDNA damage arising from excess mitochondrial H2O2 production may be a novel therapeutic target for mitigating pulmonary fibrosis. PMID:27840320

Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hess, J. D.; Gauld, I. C.; Gulliford, J.

2017-01-01

Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of various higher-actinide isotopes were irradiated for 492 effective full-power days and radiochemically assayed at Oak Ridge National Laboratory (ORNL) and Japan Atomic Energy Research Institute (JAERI). Limited data were available regarding the PFR irradiation; a six-group neutron spectra was available with some power history data to support a burnup depletion analysis validation study. Under the guidance of the Organisation for Economic Co-Operation and Development Nuclear Energy Agency (OECD NEA), the International Reactor Physics Experiment Evaluation Projectmore » (IRPhEP) and Spent Fuel Isotopic Composition (SFCOMPO) Project are collaborating to recover all measurement data pertaining to these measurements, including collaboration with the United Kingdom to obtain pertinent reactor physics design and operational history data. These activities will produce internationally peer-reviewed benchmark data to support validation of minor actinide cross section data and modern neutronic simulation of fast reactors with accompanying fuel cycle activities such as transportation, recycling, storage, and criticality safety.« less

Catalytic fast pyrolysis of white oak wood in-situ using a bubbling fluidized bed reactor

USDA-ARS?s Scientific Manuscript database

Catalytic fast pyrolysis was performed on white oak wood using two zeolite-type catalysts as bed material in a bubbling fluidized bed reactor. The two catalysts chosen, based on a previous screening study, were Ca2+ exchanged Y54 (Ca-Y54) and a proprietary ß-zeolite type catalyst (catalyst M) both ...

FAST CHOPPER BUILDING, TRA665. DETAIL SHOWS UPPER AND LOWER LEVEL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FAST CHOPPER BUILDING, TRA-665. DETAIL SHOWS UPPER AND LOWER LEVEL WALLS OF DIFFERING MATERIALS. NOTE DOORWAY TO MTR TO RIGHT OF CHOPPER BUILDING'S CLIPPED CORNER. CAMERA FACING WEST. INL NEGATIVE NO. HD42-1. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.

A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paolo Balestra; Carlo Parisi; Andrea Alfonsi

2016-02-01

The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution).more » Comparison between both solutions is briefly illustrated in this summary.« less

CT dose reduction using Automatic Exposure Control and iterative reconstruction: A chest paediatric phantoms study.

PubMed

Greffier, Joël; Pereira, Fabricio; Macri, Francesco; Beregi, Jean-Paul; Larbi, Ahmed

2016-04-01

To evaluate the impact of Automatic Exposure Control (AEC) on radiation dose and image quality in paediatric chest scans (MDCT), with or without iterative reconstruction (IR). Three anthropomorphic phantoms representing children aged one, five and 10-year-old were explored using AEC system (CARE Dose 4D) with five modulation strength options. For each phantom, six acquisitions were carried out: one with fixed mAs (without AEC) and five each with different modulation strength. Raw data were reconstructed with Filtered Back Projection (FBP) and with two distinct levels of IR using soft and strong kernels. Dose reduction and image quality indices (Noise, SNR, CNR) were measured in lung and soft tissues. Noise Power Spectrum (NPS) was evaluated with a Catphan 600 phantom. The use of AEC produced a significant dose reduction (p<0.01) for all anthropomorphic sizes employed. According to the modulation strength applied, dose delivered was reduced from 43% to 91%. This pattern led to significantly increased noise (p<0.01) and reduced SNR and CNR (p<0.01). However, IR was able to improve these indices. The use of AEC/IR preserved image quality indices with a lower dose delivered. Doses were reduced from 39% to 58% for the one-year-old phantom, from 46% to 63% for the five-year-old phantom, and from 58% to 74% for the 10-year-old phantom. In addition, AEC/IR changed the patterns of NPS curves in amplitude and in spatial frequency. In chest paediatric MDCT, the use of AEC with IR allows one to obtain a significant dose reduction while maintaining constant image quality indices. Copyright © 2016 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.

Macrophage Control of Phagocytosed Mycobacteria Is Increased by Factors Secreted by Alveolar Epithelial Cells through Nitric Oxide Independent Mechanisms

PubMed Central

Freidl, Raphaela; Fernández, Carmen

2014-01-01

Tissue-resident macrophages are heterogeneous with tissue-specific and niche-specific functions. Thus, simplified models of macrophage activation do not explain the extent of heterogeneity seen in vivo. We focus here on the respiratory tract and ask whether factors secreted by alveolar epithelial cells (AEC) can influence the functionality of resident pulmonary macrophages (PuM). We have previously reported that factors secreted by AEC increase control of intracellular growth of BCG in macrophages. In the current study, we also aimed to investigate possible mechanisms by which AEC-derived factors increase intracellular control of BCG in both primary murine interstitial macrophages, and bone marrow-derived macrophages and characterize further the effect of these factors on macrophage differentiation. We show that; a) in contrast to other macrophage types, IFN-γ did not increase intracellular growth control of Mycobacterium bovis, Bacillus Calmette-Guérin (BCG) by interstitial pulmonary macrophages although the same macrophages could be activated by factors secreted by AEC; b) the lack of response of pulmonary macrophages to IFN-γ was apparently regulated by suppressor of cytokine signaling (SOCS)1; c) AEC-derived factors did not induce pro-inflammatory pathways induced by IFN-γ e.g. expression of inducible nitric oxide synthase (iNOS), secretion of nitric oxide (NO), or IL-12, d) in contrast to IFN-γ, intracellular bacterial destruction induced by AEC-derived factors was not dependent on iNOS transcription and NO production. Collectively, our data show that PuM were restricted in inflammatory responses mediated by IFN-γ through SOCS1 and that factors secreted by AEC- enhanced the microbicidal capacities of macrophages by iNOS independent mechanisms. PMID:25089618

Hypoglycemic and anti-lipemic effects of the aqueous extract from Cissus sicyoides

PubMed Central

Viana, Glauce SB; Medeiros, Ana Carolina C; Lacerda, Ana Michelle R; Leal, L Kalyne AM; Vale, Tiago G; Matos, F José de Abreu

2004-01-01

Background Cissus sicyoides (Vitaceae) is a medicinal plant popularly known in Brazil as "cipó-pucá, anil-trepador, cortina, and insulina". The plant is used in several diseases, including rheumatism, epilepsy, stroke and also in the treatment of diabetes. In the present work, we studied the hypoglycemic and anti-lipemic effects of the aqueous extract prepared from fresh leaves of the plant (AECS), in the model of alloxan-induced diabetes in rats. In addition, hepatic enzyme levels were also determined. Results Results showed that the daily treatment of diabetic rats with AECS for 7 days (100 and 200 mg/kg, p.o.) significantly decreased blood glucose levels in 25 and 22% respectively, as compared to the same groups before AECS treatment. No significant changes were seen in control diabetic rats before (48 h after alloxan administration) and after distilled water treatment. While no changes were seen in total cholesterol levels, a significant decrease was observed in plasma triglyceride levels, in the alloxan-induced diabetic rats after AECS treatment with both doses, as compared to the same groups before treatment. Significant decreases in blood glucose (25%) and triglyceride levels (48%) were also observed in the alloxan-induced diabetic rats after 4 days treatment with AECS (200 mg/kg, p.o.). Aspartate (AST) and alanine (ALT) aminotransferases levels, in diabetic controls and AECS-treated rats, were in the range of reference values presented by normal rats. Conclusions The results justify the popular use of C. sicyoides, pointing out to the potential benefit of the plant aqueous extract (AECS) in alternative medicine, in the treatment of type 2 diabetes mellitus. PMID:15182373

Risk of high-grade cervical dysplasia and gynaecological malignancies following the cytologic diagnosis of atypical endocervical cells of undetermined significance: a retrospective study of a state-wide screening population in Western Australia.

PubMed

Munro, Aime; Williams, Vincent; Semmens, James; Leung, Yee; Stewart, Colin J R; Codde, Jim; Spilsbury, Katrina; Steel, Nerida; Cohen, Paul; O'leary, Peter

2015-06-01

In 2006, Australia adopted a revised cervical cytology terminology system, known as the Australian Modified Bethesda System (AMBS). One substantial change in the AMBS was the introduction of the diagnostic category of atypical endocervical cells (AEC) of undetermined significance. The aim of this study was to investigate the incidence of histologically confirmed high-grade cervical dysplasia (cervical intra-epithelial neoplasia (CIN) grades 2 and 3 and adenocarcinoma in situ (ACIS)), cervical carcinoma and endometrial carcinoma in women presenting with AEC on cervical cytology. A seven-year retrospective study examining clinical outcomes of women with AEC on a screening cervical smear. Cytology and histology results were extracted from the Western Australia Cervical Screening Registry, and time-to-event analysis was used to predict the odds of having or developing in situ and invasive neoplasia. AEC was reported in index smears from 0.093% (584/622754) women during the study period. No follow-up was available in 35 AEC cases. Sixty-five of the remaining 549 women (11.8%) had, or developed, high-grade cervical dysplasia within five years of their index AEC diagnosis. Endometrial cancer was diagnosed in 21 women and cervical cancer in four women during the follow-up period. Cytologic demonstration of AEC requires careful gynaecologic evaluation, particularly in younger women who may be found to have either high-grade squamous (CIN) or glandular (ACIS) lesions, while in older women, the possibility of endometrial neoplasia needs to be considered. © 2015 The Royal Australian and New Zealand College of Obstetricians and Gynaecologists.

Modeling AEC—New Approaches to Study Rare Genetic Disorders

PubMed Central

Koch, Peter J.; Dinella, Jason; Fete, Mary; Siegfried, Elaine C.; Koster, Maranke I.

2015-01-01

Ankyloblepharon-ectodermal defects-cleft lip/palate (AEC) syndrome is a rare monogenetic disorder that is characterized by severe abnormalities in ectoderm-derived tissues, such as skin and its appendages. A major cause of morbidity among affected infants is severe and chronic skin erosions. Currently, supportive care is the only available treatment option for AEC patients. Mutations in TP63, a gene that encodes key regulators of epidermal development, are the genetic cause of AEC. However, it is currently not clear how mutations in TP63 lead to the various defects seen in the patients’ skin. In this review, we will discuss current knowledge of the AEC disease mechanism obtained by studying patient tissue and genetically engineered mouse models designed to mimic aspects of the disorder. We will then focus on new approaches to model AEC, including the use of patient cells and stem cell technology to replicate the disease in a human tissue culture model. The latter approach will advance our understanding of the disease and will allow for the development of new in vitro systems to identify drugs for the treatment of skin erosions in AEC patients. Further, the use of stem cell technology, in particular induced pluripotent stem cells (iPSC), will enable researchers to develop new therapeutic approaches to treat the disease using the patient’s own cells (autologous keratinocyte transplantation) after correction of the disease-causing mutations. PMID:24665072

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

NASA Astrophysics Data System (ADS)

Alekseev, P. N.; Bobrov, E. A.; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A.

2015-12-01

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U-Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium-plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: 235U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or 233U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.

2015-12-15

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no usemore » of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.« less

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

NASA Astrophysics Data System (ADS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.

Long lifetime fast spectrum reactor for lunar surface power system

NASA Astrophysics Data System (ADS)

Kambe, Mitsuru

1993-01-01

In the framework of innovative reactor research activities, a conceptual design study of fast spectrum reactor and primary system for 800 kWe lunar surface power system to be combined with potassium Rankine cycle power conversion has been conducted to meet the power requirements of the lunar base activities in the next century. The reactor subsystem is characterized by RAPID (Refueling by All Pins Integrated Design) concept to enhance inherent safety and to enable quick and simplifed refueling in every 10 years. RAPID concept affords power plant design lifetime of up to 30 years. Integrity of the reactor structure and replacement of failed primary circuits are also discussed. Substantial reduction in per-kWh cost on considering launch, emplacement, and final disposition can be expected by a long system lifetime.

FAST NEUTRON REACTOR

DOEpatents

Soodak, H.; Wigner, E.P.

1961-07-25

A reactor comprising fissionable material in concentration sufficiently high so that the average neutron enengy within the reactor is at least 25,000 ev is described. A natural uranium blanket surrounds the reactor, and a moderating reflector surrounds the blanket. The blanket is thick enough to substantially eliminate flow of neutrons from the reflector.

Hydrogen and elemental carbon production from natural gas and other hydrocarbons

DOEpatents

Detering, Brent A.; Kong, Peter C.

2002-01-01

Diatomic hydrogen and unsaturated hydrocarbons are produced as reactor gases in a fast quench reactor. During the fast quench, the unsaturated hydrocarbons are further decomposed by reheating the reactor gases. More diatomic hydrogen is produced, along with elemental carbon. Other gas may be added at different stages in the process to form a desired end product and prevent back reactions. The product is a substantially clean-burning hydrogen fuel that leaves no greenhouse gas emissions, and elemental carbon that may be used in powder form as a commodity for several processes.

Recruited Monocytes and Type 2 Immunity Promote Lung Regeneration following Pneumonectomy.

PubMed

Lechner, Andrew J; Driver, Ian H; Lee, Jinwoo; Conroy, Carmen M; Nagle, Abigail; Locksley, Richard M; Rock, Jason R

2017-07-06

To investigate the role of immune cells in lung regeneration, we used a unilateral pneumonectomy model that promotes the formation of new alveoli in the remaining lobes. Immunofluorescence and single-cell RNA sequencing found CD115+ and CCR2+ monocytes and M2-like macrophages accumulating in the lung during the peak of type 2 alveolar epithelial stem cell (AEC2) proliferation. Genetic loss of function in mice and adoptive transfer studies revealed that bone marrow-derived macrophages (BMDMs) traffic to the lung through a CCL2-CCR2 chemokine axis and are required for optimal lung regeneration, along with Il4ra-expressing leukocytes. Our data suggest that these cells modulate AEC2 proliferation and differentiation. Finally, we provide evidence that group 2 innate lymphoid cells are a source of IL-13, which promotes lung regeneration. Together, our data highlight the potential for immunomodulatory therapies to stimulate alveologenesis in adults. Copyright © 2017 Elsevier Inc. All rights reserved.

Placental stem cell correction of murine intermediate maple syrup urine disease.

PubMed

Skvorak, Kristen J; Dorko, Kenneth; Marongiu, Fabio; Tahan, Veysel; Hansel, Marc C; Gramignoli, Roberto; Gibson, K Michael; Strom, Stephen C

2013-03-01

There is improved survival and partial metabolic correction of a mouse intermediate maple syrup urine disease (iMSUD) model after allogenic hepatocyte transplantation, confirming that a small number of enzyme-proficient liver-engrafted cells can improve phenotype. However, clinical shortages of suitable livers for hepatocyte isolation indicate a need for alternative cell sources. Human amnion epithelial cells (hAECs) share stem cell characteristics without the latter's safety and ethical concerns and differentiate to hepatocyte-like cells. Eight direct hepatic hAEC transplantations were performed in iMSUD mice over the first 35 days beginning at birth; animals were provided a normal protein diet and sacrificed at 35 and 100 days. Treatment at the neonatal stage is clinically relevant for MSUD and may offer a donor cell engraftment advantage. Survival was significantly extended and body weight was normalized in iMSUD mice receiving hAEC transplantations compared with untreated iMSUD mice, which were severely cachectic and died ≤28 days after birth. Branched chain α-keto acid dehydrogenase enzyme activity was significantly increased in transplanted livers. The branched chain amino acids leucine, isoleucine, valine, and alloisoleucine were significantly improved in serum and brain, as were other large neutral amino acids. Placental-derived stem cell transplantation lengthened survival and corrected many amino acid imbalances in a mouse model of iMSUD. This highlights the potential for their use as a viable alternative clinical therapy for MSUD and other liver-based metabolic diseases. Copyright © 2012 American Association for the Study of Liver Diseases.

Placental Stem Cell Correction of Murine Intermediate Maple Syrup Urine Disease

PubMed Central

Skvorak, Kristen J.; Dorko, Kenneth; Marongiu, Fabio; Tahan, Veysel; Hansel, Marc C.; Gramignoli, Roberto; Gibson, K. Michael; Strom, Stephen C.

2012-01-01

We previously reported improved survival and partial metabolic correction of a mouse intermediate maple syrup urine disease (iMSUD) model post allogenic hepatocyte transplant, confirming that a small number of enzyme proficient liver-engrafted cells can improve phenotype. However, clinical shortages of suitable livers for hepatocyte isolation indicate a need for alternative cell sources. Human amnion epithelial cells (hAEC) share stem cell characteristics while lacking many safety and ethical concerns, and differentiate to hepatocyte-like cells. Eight direct hepatic hAEC transplants were administered to iMSUD mice over the first 35 days beginning at birth; animals were provided a normal protein diet and sacrificed at days 35 and 100. Treatment at the neonatal stage is clinically relevant for MSUD, and may offer a donor cell engraftment advantage. Survival was significantly extended and body weight was normalized in iMSUD mice receiving hAEC transplants compared to iMSUD (severely cachectic; dead ≤28 days). Branched chain α-keto acid dehydrogenase enzyme activity was significantly increased in transplanted livers. Branched chain amino acids leucine, isoleucine, valine, and alloisoleucine were significantly improved in the sera and brain, as were other large neutral amino acids. Conclusion: Placental-derived stem cell transplantation lengthened survival and corrected many amino acid imbalances in a mouse model of iMSUD. This highlights the potential for their use as a viable alternative clinical therapy for MSUD and other liver-based metabolic diseases. PMID:23175463

Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.

2008-06-23

This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been mademore » at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.« less

DISCHARGE VALVE FOR GRANULAR MATERIAL

DOEpatents

Stoughton, L.D.; Robinson, S.T.

1962-05-15

A gravity-red dispenser or valve is designed for discharging the fueled spherical elements used in a pebble bed reactor. The dispenser consists of an axially movable tube terminating under a hood having side walls with openings. When the tube is moved so that its top edge is above the tops of the side openings the elements will not flow. As the tube is moved downwardly, the elements flow into the hood through the side openings and over the top edge into the tube at an increasing rate as the tube is lowered further. The tube is spaced at all times from the hood and side walls a distance greater than the diameter of the largest element to prevent damaging of the elements when the dispenser is closed to flow. (AEC)

A high voltage power supply for the AE-C and D low energy electron experiment

NASA Technical Reports Server (NTRS)

Gillis, J. A.

1974-01-01

A description is given of the electrical and mechanical design and operation of high voltage power supplies for space flight use. The supply was used to generate the spiraltron high voltage for low energy electron experiment on AE-C and D. Two versions of the supply were designed and built; one design is referred to as the low power version (AE-C) and the other as the high power version (AE-D). Performance is discussed under all operating conditions.

Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

NASA Astrophysics Data System (ADS)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

Preparing for radiological assessments in the event of a tornado strike at Argonne National Lab. -East

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goodkind, M.E.; Klimczak, C.A.; Munyon, W.J.

1993-01-01

Argonne National Laboratory-East (ANL) is a Department of Energy (DOE)-owned, contractor-operated national laboratory located 22 miles southwest of downtown Chicago on a wooded, 1700-acre site. The principal nuclear facilities at ANL include a large fast neutron source (Intense Pulse Neutron Source) in which high-energy protons strike a uranium target to produce neutrons for research studies; [sup 60]Co irradiation sources; chemical and metallurgical plutonium laboratories, some of which are currently being decommissioned; several large hot cell facilities designed for work with multicurie quantities of actinide elements and irradiated reactor fuel materials; a few small research reactors currently in different phases ofmore » being decommissioned; and a variety of research laboratories handling many different sources in various chemical and physical forms. The hazards analysis for the ANL site shows that tornado strikes are a serious threat. The site has been struck twice in the past 20 yr, receiving only minor building damage and no release of radioactivity to the environment. Although radioactive materials in general are handled in areas that provide good tornado protection, ANL is prepared to address the problems that would occur should there be a loss of control of radioactive materials due to severe building damage.« less

Fission neutron source in Rome

NASA Astrophysics Data System (ADS)

Coppola, Mario; Di Majo, V.; Ingrao, G.; Rebessi, S.; Testa, A.

1997-02-01

A fission neutron source is operating in Rome at the ENEA Casaccia Research Center since 1971, consisting of a low power fast reactor named RSV-Tapiro. it is employed for a variety of experiments, including dosimetry, material testing, radiation protection and biology. In particular, application to experimental radiobiology includes studies of the biological action of neutrons in the whole-body irradiated animal, or in specialized systems in vivo or in vitro. For his purpose a vertical irradiation facility was originally constructed. Recently, a new horizontal irradiation facility has been designed to allow the exposure of larger samples or larger sample batches at one time. Dosimetry at the sample irradiation positions is routinely carried out by the conventional method of using two ion chambers. This physical dosimetry has recently been compared with the results of biological dosimetry based on the detection of chromosomal aberrations in peripheral blood human lymphocytes irradiated in vitro. A characterization of the radiation quality in the two configurations has been carried out by tissue equivalent proportional counter microdosimetry measurements. Information about the main characteristics of the reactor and the two irradiation facilities is provided and relevant results of the various measurements are summarized. Radiobiological results obtained using this neutron source are also briefly outlined.

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-2: Liquid Metal Fast Breeder Reactors.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical liquid metal fast breeder reactor (LMFBR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating the use with a simplified model. The heart of the module is…

Packed rod neutron shield for fast nuclear reactors

DOEpatents

Eck, John E.; Kasberg, Alvin H.

1978-01-01

A fast neutron nuclear reactor including a core and a plurality of vertically oriented neutron shield assemblies surrounding the core. Each assembly includes closely packed cylindrical rods within a polygonal metallic duct. The shield assemblies are less susceptible to thermal stresses and are less massive than solid shield assemblies, and are cooled by liquid coolant flow through interstices among the rods and duct.

FAST CHOPPER DETECTOR HOUSE, TRA665. SECOND FLOOR ADDITION: PLAN, SECTIONS ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FAST CHOPPER DETECTOR HOUSE, TRA-665. SECOND FLOOR ADDITION: PLAN, SECTIONS AND DETAILS AS ADDED TO THE EXISTING CHOPPER HOUSE IN 1962. F.C. TORKELSON 842-MTR-665-S-3, 4/1962. INL INDEX NO. 531-0665-60-851-150997, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ADJUSTABLE DOUBLE PULSE GENERATOR

DOEpatents

Gratian, J.W.; Gratian, A.C.

1961-08-01

>A modulator pulse source having adjustable pulse width and adjustable pulse spacing is described. The generator consists of a cross coupled multivibrator having adjustable time constant circuitry in each leg, an adjustable differentiating circuit in the output of each leg, a mixing and rectifying circuit for combining the differentiated pulses and generating in its output a resultant sequence of negative pulses, and a final amplifying circuit for inverting and square-topping the pulses. (AEC)

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions.

PubMed

Geslot, B; Vermeeren, L; Filliatre, P; Lopez, A Legrand; Barbot, L; Jammes, C; Bréaud, S; Oriol, L; Villard, J-F

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 10(20) n∕cm(2). A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

NASA Astrophysics Data System (ADS)

Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

A summary of sodium-cooled fast reactor development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aoto, Kazumi; Dufour, Philippe; Hongyi, Yang

Much of the basic technology for the Sodium-cooled fast Reactor (SFR) has been established through long term development experience with former fast reactor programs, and is being confirmed by the Phénix end-of-life tests in France, the restart of Monju in Japan, the lifetime extension of BN-600 in Russia, and the startup of the China Experimental Fast Reactor in China. Planned startup in 2014 for new SFRs: BN-800 in Russia and PFBR in India, will further enhance the confirmation of the SFR basic technology. Nowadays, the SFR development has advanced to aiming at establishment of the Generation-IV system which is dedicatedmore » to sustainable energy generation and actinide management, and several advanced SFR concepts are under development such as PRISM, JSFR, ASTRID, PGSFR, BN-1200, and CFR-600. Generation-IV International Forum is an international collaboration framework where various R&D activities are progressing on design of system and component, safety and operation, advanced fuel, and actinide cycle for the Generation-IV SFR development, and will play a beneficial role of promoting them thorough providing an opportunity to share the past experience and the latest data of design and R&D among countries developing SFR.« less

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

NASA Astrophysics Data System (ADS)

Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

2012-08-01

CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

Quantitative Analysis of Proteome Modulations in Alveolar Epithelial Type II Cells in Response to Pulmonary Aspergillus fumigatus Infection.

PubMed

Seddigh, Pegah; Bracht, Thilo; Molinier-Frenkel, Válerie; Castellano, Flavia; Kniemeyer, Olaf; Schuster, Marc; Weski, Juliane; Hasenberg, Anja; Kraus, Andreas; Poschet, Gernot; Hager, Thomas; Theegarten, Dirk; Opitz, Christiane A; Brakhage, Axel A; Sitek, Barbara; Hasenberg, Mike; Gunzer, Matthias

2017-12-01

The ubiquitous mold Aspergillus fumigatus threatens immunosuppressed patients as inducer of lethal invasive aspergillosis. A. fumigatus conidia are airborne and reach the alveoli, where they encounter alveolar epithelial cells (AEC). Previous studies reported the importance of the surfactant-producing AEC II during A. fumigatus infection via in vitro experiments using cell lines. We established a negative isolation protocol yielding untouched primary murine AEC II with a purity >90%, allowing ex vivo analyses of the cells, which encountered the mold in vivo By label-free proteome analysis of AEC II isolated from mice 24h after A. fumigatus or mock infection we quantified 2256 proteins and found 154 proteins to be significantly differentially abundant between both groups (ANOVA p value ≤ 0.01, ratio of means ≥1.5 or ≤0.67, quantified with ≥2 peptides). Most of these proteins were higher abundant in the infected condition and reflected a comprehensive activation of AEC II on interaction with A. fumigatus This was especially represented by proteins related to oxidative phosphorylation, hence energy production. However, the most strongly induced protein was the l-amino acid oxidase (LAAO) Interleukin 4 induced 1 (IL4I1) with a 42.9 fold higher abundance (ANOVA p value 2.91 -10 ). IL4I1 has previously been found in B cells, macrophages, dendritic cells and rare neurons. Increased IL4I1 abundance in AEC II was confirmed by qPCR, Western blot and immunohistology. Furthermore, A. fumigatus infected lungs showed high levels of IL4I1 metabolic products. Importantly, higher IL4I1 abundance was also confirmed in lung tissue from human aspergilloma. Because LAAO are key enzymes for bactericidal product generation, AEC II might actively participate in pathogen defense. We provide insights into proteome changes of primary AEC II thereby opening new avenues to analyze the molecular changes of this central lung cell on infectious threats. Data are available via ProteomeXchange with identifier PXD005834. © 2017 by The American Society for Biochemistry and Molecular Biology, Inc.

Reduced transforming growth factor β1 (TGF-β1) in the repair of airway epithelial cells of children with asthma.

PubMed

Ling, Kak-Ming; Sutanto, Erika N; Iosifidis, Thomas; Kicic-Starcevich, Elizabeth; Looi, Kevin; Garratt, Luke W; Martinovich, Kelly M; Lannigan, Francis J; Knight, Darryl A; Stick, Stephen M; Kicic, Anthony

2016-10-01

Evidence into the role of TGF-β1 in airway epithelial repair in asthma is still controversial. This study tested the hypothesis that the reduced TGF-β1 levels previously observed in paediatric asthmatic airway epithelial cells directly contribute to the dysregulated repair seen in these cells. Primary airway epithelial cells (pAEC) from children with asthma (n = 16) and non-asthmatic subjects (n = 20) were isolated, and subcultured for investigation of TGF-β1 gene and protein via quantitative polymerase chain reaction (qPCR) and enzyme-linked immunosorbent assay (ELISA), respectively. Expression of other associated genes such as integrins αvβ6, αvβ8 and MT1-MMP were also tested. Small interfering RNA (siRNA) was employed to assess the role of TGF-β1 during wound repair. TGF-β1 gene and protein expression were significantly downregulated in asthmatic pAEC over the course of repair, compared with cells from non-asthmatic children. Messenger RNA (mRNA) expression of TGF-β1 was also directly implicated in non-asthmatic and asthmatic pAEC proliferation over their quiescent counterparts. Small interfering RNA-mediated knockdown of TGF-β1 compromised repair in non-asthmatic pAEC and exacerbated the dysregulated repair seen in asthmatic pAEC. Expression of major TGF-β1 activators of epithelial cells, integrin αvβ6 and αvβ8 was also measured and there was no difference in αvβ6 gene expression between the two cohorts. Although integrin αvβ8 gene expression was significantly higher in asthmatic pAEC, the expression of MT1-MMP (MMP14) which facilitates the αvβ8 mediated TGF-β1 activation was significantly downregulated. Our data has highlighted the importance of TGF-β1 in pAEC wound repair in vitro. The significantly lower levels seen in asthmatic pAEC subsequently contributes to the dysregulated repair observed in these cells. © 2016 Asian Pacific Society of Respirology.

Lessons Learned about Liquid Metal Reactors from FFTF Experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.

2016-09-20

The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens.more » In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports« less

Sensitivity Analysis and Optimization of the Nuclear Fuel Cycle: A Systematic Approach

NASA Astrophysics Data System (ADS)

Passerini, Stefano

For decades, nuclear energy development was based on the expectation that recycling of the fissionable materials in the used fuel from today's light water reactors into advanced (fast) reactors would be implemented as soon as technically feasible in order to extend the nuclear fuel resources. More recently, arguments have been made for deployment of fast reactors in order to reduce the amount of higher actinides, hence the longevity of radioactivity, in the materials destined to a geologic repository. The cost of the fast reactors, together with concerns about the proliferation of the technology of extraction of plutonium from used LWR fuel as well as the large investments in construction of reprocessing facilities have been the basis for arguments to defer the introduction of recycling technologies in many countries including the US. In this thesis, the impacts of alternative reactor technologies on the fuel cycle are assessed. Additionally, metrics to characterize the fuel cycles and systematic approaches to using them to optimize the fuel cycle are presented. The fuel cycle options of the 2010 MIT fuel cycle study are re-examined in light of the expected slower rate of growth in nuclear energy today, using the CAFCA (Code for Advanced Fuel Cycle Analysis). The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycle options available in the future. The options include limited recycling in L WRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. Additional fuel cycle scenarios presented for the first time in this work assume the deployment of innovative recycling reactor technologies such as the Reduced Moderation Boiling Water Reactors and Uranium-235 initiated Fast Reactors. A sensitivity study focused on system and technology parameters of interest has been conducted to test the robustness of the conclusions presented in the MIT Fuel Cycle Study. These conclusions are found to still hold, even when considering alternative technologies and different sets of simulation assumptions. Additionally, a first of a kind optimization scheme for the nuclear fuel cycle analysis is proposed and the applications of such an optimization are discussed. Optimization metrics of interest for different stakeholders in the fuel cycle (economics, fuel resource utilization, high level waste, transuranics/proliferation management, and environmental impact) are utilized for two different optimization techniques: a linear one and a stochastic one. Stakeholder elicitation provided sets of relative weights for the identified metrics appropriate to each stakeholder group, which were then successfully used to arrive at optimum fuel cycle configurations for recycling technologies. The stochastic optimization tool, based on a genetic algorithm, was used to identify non-inferior solutions according to Pareto's dominance approach to optimization. The main tradeoff for fuel cycle optimization was found to be between economics and most of the other identified metrics. (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs mit.edu)

Cross-Section Measurements in the Fast Neutron Energy Range

NASA Astrophysics Data System (ADS)

Plompen, Arjan

2006-04-01

Generation IV focuses research for advanced nuclear reactors on six concepts. Three of these concepts, the lead, gas and sodium fast reactors (LFR, GFR and SFR) have fast neutron spectra, whereas a fourth, the super-critical water reactor (SCWR), can be configured to have a fast spectrum. Such fast neutron spectra are essential to meet the sustainability objective of GenIV. Nuclear data requirements for GenIV concepts will therefore emphasize the energy region from about 1 keV to 10 MeV. Here, the potential is illustrated of the GELINA neutron time-of-flight facility and the Van de Graaff laboratory at IRMM to measure the relevant nuclear data in this energy range: the total, capture, fission and inelastic-scattering cross sections. In particular, measurement results will be shown for lead and bismuth inelastic scattering for which the need was recently expressed in a quantitative way by Aliberti et al. for Accelerator Driven Systems. Even without completion of the quantitative assessment of the data needs for GenIV concepts at ANL it is clear that this particular effort is of relevance to LFR system studies.

Tailoring automatic exposure control toward constant detectability in digital mammography.

PubMed

Salvagnini, Elena; Bosmans, Hilde; Struelens, Lara; Marshall, Nicholas W

2015-07-01

The automatic exposure control (AEC) modes of most full field digital mammography (FFDM) systems are set up to hold pixel value (PV) constant as breast thickness changes. This paper proposes an alternative AEC mode, set up to maintain some minimum detectability level, with the ultimate goal of improving object detectability at larger breast thicknesses. The default "opdose" AEC mode of a Siemens MAMMOMAT Inspiration FFDM system was assessed using poly(methyl methacrylate) (PMMA) of thickness 20, 30, 40, 50, 60, and 70 mm to find the tube voltage and anode/filter combination programmed for each thickness; these beam quality settings were used for the modified AEC mode. Detectability index (d'), in terms of a non-prewhitened model observer with eye filter, was then calculated as a function of tube current-time product (mAs) for each thickness. A modified AEC could then be designed in which detectability never fell below some minimum setting for any thickness in the operating range. In this study, the value was chosen such that the system met the achievable threshold gold thickness (Tt) in the European guidelines for the 0.1 mm diameter disc (i.e., Tt ≤ 1.10 μm gold). The default and modified AEC modes were compared in terms of contrast-detail performance (Tt), calculated detectability (d'), signal-difference-to-noise ratio (SDNR), and mean glandular dose (MGD). The influence of a structured background on object detectability for both AEC modes was examined using a CIRS BR3D phantom. Computer-based CDMAM reading was used for the homogeneous case, while the images with the BR3D background were scored by human observers. The default opdose AEC mode maintained PV constant as PMMA thickness increased, leading to a reduction in SDNR for the homogeneous background 39% and d' 37% in going from 20 to 70 mm; introduction of the structured BR3D plate changed these figures to 22% (SDNR) and 6% (d'), respectively. Threshold gold thickness (0.1 mm diameter disc) for the default AEC mode in the homogeneous background increased by 62% in going from 20 to 70 mm PMMA thickness; in the structured background, the increase was 39%. Implementation of the modified mode entailed an increase in mAs at PMMA thicknesses >40 mm; the modified AEC held threshold gold thickness constant above 40 mm PMMA with a maximum deviation of 5% in the homogeneous background and 3% in structured background. SDNR was also held constant with a maximum deviation of 4% and 2% for the homogeneous and the structured background, respectively. These results were obtained with an increase of MGD between 15% and 73% going from 40 to 70 mm PMMA thickness. This work has proposed and implemented a modified AEC mode, tailored toward constant detectability at larger breast thickness, i.e., above 40 mm PMMA equivalent. The desired improvement in object detectability could be obtained while maintaining MGD within the European guidelines achievable dose limit. (A study designed to verify the performance of the modified mode using more clinically realistic data is currently underway.).

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salvagnini, Elena, E-mail: elena.salvagnini@uzleuven.be; Bosmans, Hilde; Struelens, Lara

Purpose: The automatic exposure control (AEC) modes of most full field digital mammography (FFDM) systems are set up to hold pixel value (PV) constant as breast thickness changes. This paper proposes an alternative AEC mode, set up to maintain some minimum detectability level, with the ultimate goal of improving object detectability at larger breast thicknesses. Methods: The default “OPDOSE” AEC mode of a Siemens MAMMOMAT Inspiration FFDM system was assessed using poly(methyl methacrylate) (PMMA) of thickness 20, 30, 40, 50, 60, and 70 mm to find the tube voltage and anode/filter combination programmed for each thickness; these beam quality settingsmore » were used for the modified AEC mode. Detectability index (d′), in terms of a non-prewhitened model observer with eye filter, was then calculated as a function of tube current-time product (mAs) for each thickness. A modified AEC could then be designed in which detectability never fell below some minimum setting for any thickness in the operating range. In this study, the value was chosen such that the system met the achievable threshold gold thickness (T{sub t}) in the European guidelines for the 0.1 mm diameter disc (i.e., T{sub t} ≤ 1.10 μm gold). The default and modified AEC modes were compared in terms of contrast-detail performance (T{sub t}), calculated detectability (d′), signal-difference-to-noise ratio (SDNR), and mean glandular dose (MGD). The influence of a structured background on object detectability for both AEC modes was examined using a CIRS BR3D phantom. Computer-based CDMAM reading was used for the homogeneous case, while the images with the BR3D background were scored by human observers. Results: The default OPDOSE AEC mode maintained PV constant as PMMA thickness increased, leading to a reduction in SDNR for the homogeneous background 39% and d′ 37% in going from 20 to 70 mm; introduction of the structured BR3D plate changed these figures to 22% (SDNR) and 6% (d′), respectively. Threshold gold thickness (0.1 mm diameter disc) for the default AEC mode in the homogeneous background increased by 62% in going from 20 to 70 mm PMMA thickness; in the structured background, the increase was 39%. Implementation of the modified mode entailed an increase in mAs at PMMA thicknesses >40 mm; the modified AEC held threshold gold thickness constant above 40 mm PMMA with a maximum deviation of 5% in the homogeneous background and 3% in structured background. SDNR was also held constant with a maximum deviation of 4% and 2% for the homogeneous and the structured background, respectively. These results were obtained with an increase of MGD between 15% and 73% going from 40 to 70 mm PMMA thickness. Conclusions: This work has proposed and implemented a modified AEC mode, tailored toward constant detectability at larger breast thickness, i.e., above 40 mm PMMA equivalent. The desired improvement in object detectability could be obtained while maintaining MGD within the European guidelines achievable dose limit. (A study designed to verify the performance of the modified mode using more clinically realistic data is currently underway.)« less

Heterogeneous sodium fast reactor designed for transmuting minor actinide waste isotopes into plutonium fuel

NASA Astrophysics Data System (ADS)

Bays, Samuel Eugene

2008-10-01

In the past several years there has been a renewed interest in sodium fast reactor (SFR) technology for the purpose of destroying transuranic waste (TRU) produced by light water reactors (LWR). The utility of SFRs as waste burners is due to the fact that higher neutron energies allow all of the actinides, including the minor actinides (MA), to contribute to fission. It is well understood that many of the design issues of LWR spent nuclear fuel (SNF) disposal in a geologic repository are linked to MAs. Because the probability of fission for essentially all the "non-fissile" MAs is nearly zero at low neutron energies, these isotopes act as a neutron capture sink in most thermal reactor systems. Furthermore, because most of the isotopes produced by these capture reactions are also non-fissile, they too are neutron sinks in most thermal reactor systems. Conversely, with high neutron energies, the MAs can produce neutrons by fast fission. Additionally, capture reactions transmute the MAs into mostly plutonium isotopes, which can fission more readily at any energy. The transmutation of non-fissile into fissile atoms is the premise of the plutonium breeder reactor. In a breeder reactor, not only does the non-fissile "fertile" U-238 atom contribute fast fission neutrons, but also transmutes into fissile Pu-239. The fissile value of the plutonium produced by MA transmutation can only be realized in fast neutron spectra. This is due to the fact that the predominate isotope produced by MA transmutation, Pu-238, is itself not fissile. However, the Pu-238 fission cross section is significantly larger than the original transmutation parent, predominately: Np-237 and Am-241, in the fast energy range. Also, Pu-238's fission cross section and fission-to-capture ratio is almost as high as that of fissile Pu-239 in the fast neutron spectrum. It is also important to note that a neutron absorption in Pu-238, that does not cause fission, will instead produce fissile Pu-239. Given this fast fissile quality and also the fact that Pu-238 is transmuted from Np-237 and Am-241, these MAs are regarded as fertile material in the SFR design proposed by this dissertation. This dissertation demonstrates a SFR design which is dedicated to plutonium breeding by targeting Am-241 transmutation. This SFR design uses a moderated axial transmutation target that functions primarily as a pseudo-blanket fuel, which is reprocessed with the active driver fuel in an integrated recycling strategy. This work demonstrates the cost and feasibility advantages of plutonium breeding via MA transmutation by adopting reactor, reprocessing and fuel technologies previously demonstrated for traditional breeder reactors. The fuel cycle proposed seeks to find a harmony between the waste management advantages of transuranic burning SFRs and the resource sustainability of traditional plutonium breeder SFRs. As a result, the enhanced plutonium conversion from MAs decreases the burner SFR's fuel costs, by extracting more fissile value from the initial TRU purchased through SNF reprocessing.

FAST CHOPPER DETECTOR HOUSE, TRA665. FIRST FLOOR, PLAN AND SECTION, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FAST CHOPPER DETECTOR HOUSE, TRA-665. FIRST FLOOR, PLAN AND SECTION, AS PROPOSED FOR MODIFICATION IN 1962. CONCRETE WALLS THREE FEET THICK. EXISTING WINDOWS IN MTR AND DETECTOR HOUSE WALLS WERE TO BE FILLED IN WITH HIGH-DENSITY BRICK. NOTE 20-METER MARK, WHERE THE FAST CHOPPER DETECTOR HAD BEEN LOCATED. F.C. TORKELSON 842-MTR-665-S-2, 4/1962. INL INDEX NO. 531-0665-60-851-150996, REV. 5. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Thermal-hydraulics of internally heated molten salts and application to the Molten Salt Fast Reactor

NASA Astrophysics Data System (ADS)

Fiorina, Carlo; Cammi, Antonio; Luzzi, Lelio; Mikityuk, Konstantin; Ninokata, Hisashi; Ricotti, Marco E.

2014-04-01

The Molten Salt Reactors (MSR) are an innovative kind of nuclear reactors and are presently considered in the framework of the Generation IV International Forum (GIF-IV) for their promising performances in terms of low resource utilization, waste minimization and enhanced safety. A unique feature of MSRs is that molten fluoride salts play the distinctive role of both fuel (heat source) and coolant. The presence of an internal heat generation perturbs the temperature field and consequences are to be expected on the heat transfer characteristics of the molten salts. In this paper, the problem of heat transfer for internally heated fluids in a straight circular channel is first faced on a theoretical ground. The effect of internal heat generation is demonstrated to be described by a corrective factor applied to traditional correlations for the Nusselt number. It is shown that the corrective factor can be fully characterized by making explicit the dependency on Reynolds and Prandtl numbers. On this basis, a preliminary correlation is proposed for the case of molten fluoride salts by interpolating the results provided by an analytic approach previously developed at the Politecnico di Milano. The experimental facility and the related measuring procedure for testing the proposed correlation are then presented. Finally, the developed correlation is used to carry out a parametric investigation on the effect of internal heat generation on the main out-of-core components of the Molten Salt Fast Reactor (MSFR), the reference circulating-fuel MSR design in the GIF-IV. The volumetric power determines higher temperatures at the channel wall, but the effect is significant only in case of large diameters and/or low velocities.

Atmosphere Explorer control system software (version 2.0)

NASA Technical Reports Server (NTRS)

Mocarsky, W.; Villasenor, A.

1973-01-01

The Atmosphere Explorer Control System (AECS) was developed to provide automatic computer control of the Atmosphere Explorer spacecraft and experiments. The software performs several vital functions, such as issuing commands to the spacecraft and experiments, receiving and processing telemetry data, and allowing for extensive data processing by experiment analysis programs. The AECS was written for a 48K XEROX Data System Sigma 5 computer, and coexists in core with the XDS Real-time Batch Monitor (RBM) executive system. RBM is a flexible operating system designed for a real-time foreground/background environment, and hence is ideally suited for this application. Existing capabilities of RBM have been used as much as possible by AECS to minimize programming redundancy. The most important functions of the AECS are to send commands to the spacecraft and experiments, and to receive, process, and display telemetry data.

Mechanisms of transcriptional repression of cell-cycle G2/M promoters by p63

PubMed Central

Testoni, Barbara; Mantovani, Roberto

2006-01-01

p63 is a developmentally regulated transcription factor related to p53, which activates and represses specific genes. The human AEC (Ankyloblepharon–Ectodermal dysplasia-Clefting) and EEC (Ectrodactyly–Ectodermal dysplasia–Cleft lip/palate) syndromes are caused by missense mutations of p63, within the DNA-binding domain (EEC) or in the C-terminal sterile alpha motif domain (AEC). We show here that p63 represses transcription of cell-cycle G2/M genes by binding to multiple CCAAT core promoters in immortalized and primary keratinocytes. The CCAAT-activator NF-Y and ΔNp63α are associated in vivo and a conserved α-helix of the NF-YC histone fold is required. p63 AEC mutants, but not an EEC mutant, are incapable to bind NF-Y. ΔNp63α, but not the AEC mutants repress CCAAT-dependent transcription of G2/M genes. Chromatin immunoprecipitation recruitment assays establish that the AEC mutants are not recruited to G2/M promoters, while normally present on 14-3-3σ, which contains a sequence-specific binding site. Surprisingly, the EEC C306R mutant activates transcription. Upon keratinocytes differentiation, NF-Y and p63 remain bound to G2/M promoters, while HDACs are recruited, histones deacetylated, Pol II displaced and transcription repressed. Our data indicate that NF-Y is a molecular target of p63 and that inhibition of growth activating genes upon differentiation is compromised by AEC missense mutations. PMID:16473849

Amniotic Mesenchymal Stromal Cells Exhibit Preferential Osteogenic and Chondrogenic Differentiation and Enhanced Matrix Production Compared With Adipose Mesenchymal Stromal Cells.

PubMed

Topoluk, Natasha; Hawkins, Richard; Tokish, John; Mercuri, Jeremy

2017-09-01

Therapeutic efficacy of various mesenchymal stromal cell (MSC) types for orthopaedic applications is currently being investigated. While the concept of MSC therapy is well grounded in the basic science of healing and regeneration, little is known about individual MSC populations in terms of their propensity to promote the repair and/or regeneration of specific musculoskeletal tissues. Two promising MSC sources, adipose and amnion, have each demonstrated differentiation and extracellular matrix (ECM) production in the setting of musculoskeletal tissue regeneration. However, no study to date has directly compared the differentiation potential of these 2 MSC populations. To compare the ability of human adipose- and amnion-derived MSCs to undergo osteogenic and chondrogenic differentiation. Controlled laboratory study. MSC populations from the human term amnion were quantified and characterized via cell counting, histologic assessment, and flow cytometry. Differentiation of these cells in comparison to commercially purchased human adipose-derived mesenchymal stromal cells (hADSCs) in the presence and absence of differentiation media was evaluated via reverse transcription polymerase chain reaction (PCR) for bone and cartilage gene transcript markers and histology/immunohistochemistry to examine ECM production. Analysis of variance and paired t tests were performed to compare results across all cell groups investigated. The authors confirmed that the human term amnion contains 2 primary cell types demonstrating MSC characteristics-(1) human amniotic epithelial cells (hAECs) and (2) human amniotic mesenchymal stromal cells (hAMSCs)-and each exhibited more than 90% staining for MSC surface markers (CD90, CD105, CD73). Average viable hAEC and hAMSC yields at harvest were 2.3 × 10 6 ± 3.7 × 10 5 and 1.6 × 10 6 ± 4.7 × 10 5 per milliliter of amnion, respectively. As well, hAECs and hAMSCs demonstrated significantly greater osteocalcin ( P = .025), aggrecan ( P < .0001), and collagen type 2 ( P = .044) gene expression compared with hADSCs, respectively, after culture in differentiation medium. Moreover, both hAECs and hAMSCs produced significantly greater quantities of mineralized ( P < .0001) and cartilaginous ( P = .0004) matrix at earlier time points compared with hADSCs when cultured under identical osteogenic and chondrogenic differentiation conditions, respectively. Amnion-derived MSCs demonstrate a greater differentiation potential toward bone and cartilage compared with hADSCs. Amniotic MSCs may be the source of choice in the regenerative treatment of bone or osteochondral musculoskeletal disease. They show significantly higher yields and better differentiation toward these tissues than MSCs derived from adipose.

Thermodynamic properties of a high pressure subcritical UF6/He gas volume (irradiated by an external source)

NASA Technical Reports Server (NTRS)

Sterritt, D. E.; Lalos, G. T.; Schneider, R. T.

1976-01-01

A computer simulation study concerning a compressed fissioning UF6 gas is presented. The compression is to be achieved by a ballistic piston compressor. Data on UF6 obtained with this compressor were incorporated in the simulation study. As a neutron source to create the fission events in the compressed gas, a fast burst reactor was considered. The conclusion is that it takes a neutron flux in excess of 10 to the 15th power n/sec sq cm to produce measurable increases in pressure and temperature, while a flux in excess of 10 to 19th power n/sq cm sec would probably damage the compressor.

Thermodynamic properties of a high pressure subcritical UF/sub 6/He gas volume (irradiated by an external source)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterritt, D.E.; Lalos, G.T.; Schneider, R.T.

1976-12-01

A computer simulation study concerning a compressed fissioning UF/sub 6/ gas is presented. The compression is to be achieved by a ballistic piston compressor. Data on UF/sub 6/ obtained with this compressor were incorporated in the simulation study. As a neutron source to create the fission events in the compressed gas, a fast burst reactor was considered. The conclusion is that it takes a neutron flux in excess of 10/sup 15/ n/cm/sup 2/-s to produce measurable increases in pressure and temperature, while a flux in excess of 10/sup 19/ n/cm/sup 2/-s would probably damage the compressor.

SHARP pre-release v1.0 - Current Status and Documentation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mahadevan, Vijay S.; Rahaman, Ronald O.

The NEAMS Reactor Product Line effort aims to develop an integrated multiphysics simulation capability for the design and analysis of future generations of nuclear power plants. The Reactor Product Line code suite’s multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. In this report, building on a several previous report issued in September 2014, we describe our continued efforts to integrate thermal/hydraulics, neutronics, and structural mechanics modeling codes to perform coupled analysis of a representativemore » fast sodium-cooled reactor core in preparation for a unified release of the toolkit. The work reported in the current document covers the software engineering aspects of managing the entire stack of components in the SHARP toolkit and the continuous integration efforts ongoing to prepare a release candidate for interested reactor analysis users. Here we report on the continued integration effort of PROTEUS/Nek5000 and Diablo into the NEAMS framework and the software processes that enable users to utilize the capabilities without losing scientific productivity. Due to the complexity of the individual modules and their necessary/optional dependency library chain, we focus on the configuration and build aspects for the SHARP toolkit, which includes capability to autodownload dependencies and configure/install with optimal flags in an architecture-aware fashion. Such complexity is untenable without strong software engineering processes such as source management, source control, change reviews, unit tests, integration tests and continuous test suites. Details on these processes are provided in the report as a building step for a SHARP user guide that will accompany the first release, expected by Mar 2016.« less

Exercise training regulates SOD-1 and oxidative stress in porcine aortic endothelium.

PubMed

Rush, James W E; Turk, James R; Laughlin, M Harold

2003-04-01

Vascular oxidative stress contributes to endothelial dysfunction. Aerobic exercise training improves vascular function. The purpose of this study was to test the hypothesis that exercise training would improve the balance of antioxidant to prooxidant enzymes and reduce markers of oxidative stress in aortic endothelial cells (AEC). Female Yucatan miniature pigs either remained sedentary (SED) or were exercise trained (EX) for 16-19 wk. EX pigs had increased AEC SOD-1 protein levels and Cu/Zn SOD activity of the whole aorta compared with SED pigs. Protein levels of other antioxidant enzymes (SOD-2, catalase) were not affected by exercise training. Protein levels of p67(phox), a subunit of the prooxidant enzyme NAD(P)H oxidase, were reduced in EX vs. SED AEC. These EX adaptations were associated with lower AEC malondialdehyde levels and decreased phosphorylation of ERK-1/2. Endothelial nitric oxide synthase protein, protein nitrotyrosine content, and heme oxygenase-1 protein were not different in EX vs. SED pigs. We conclude that chronic aerobic exercise training influenced both antioxidant and prooxidant enzymes and decreased indexes of oxidative stress in AEC. These adaptations may contribute to improved endothelial function with exercise training.

Interactions of Aspergillus fumigatus Conidia with Airway Epithelial Cells: A Critical Review

PubMed Central

Croft, Carys A.; Culibrk, Luka; Moore, Margo M.; Tebbutt, Scott J.

2016-01-01

Aspergillus fumigatus is an environmental filamentous fungus that also acts as an opportunistic pathogen able to cause a variety of symptoms, from an allergic response to a life-threatening disseminated fungal infection. The infectious agents are inhaled conidia whose first point of contact is most likely to be an airway epithelial cell (AEC). The interaction between epithelial cells and conidia is multifaceted and complex, and has implications for later steps in pathogenesis. Increasing evidence has demonstrated a key role for the airway epithelium in the response to respiratory pathogens, particularly at early stages of infection; therefore, elucidating the early stages of interaction of conidia with AECs is essential to understand the establishment of infection in cohorts of at-risk patients. Here, we present a comprehensive review of the early interactions between A. fumigatus and AECs, including bronchial and alveolar epithelial cells. We describe mechanisms of adhesion, internalization of conidia by AECs, the immune response of AECs, as well as the role of fungal virulence factors, and patterns of fungal gene expression characteristic of early infection. A clear understanding of the mechanisms involved in the early establishment of infection by A. fumigatus could point to novel targets for therapy and prophylaxis. PMID:27092126

Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

DOEpatents

Brehm, Jr., William F.; Colburn, Richard P.

1982-01-01

An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.

Status of liquid metal fast breeder reactor fuel development in Japan

NASA Astrophysics Data System (ADS)

Katsuragawa, M.; Kashihara, H.; Akebi, M.

1993-09-01

The mixed-oxide fuel technology for a liquid metal fast breeder reactor (LMFBR) in Japan is progressing toward commercial deployment of LMFBR. Based on accumulated experience in Joyo and Monju fuel development, efforts for large scale LMFBR fuel development are devoted to improved irradiation performance, reliability and economy. This paper summarizes accomplishments, current activities and future plans for LMFBR fuel development in Japan.

Assessment of the high temperature fission chamber technology for the French fast reactor program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jammes, C.; Filliatre, P.; Geslot, B.

2011-07-01

High temperature fission chambers are key instruments for the control and protection of the sodium-cooled fast reactor. First, the developments of those neutron detectors, which are carried out either in France or abroad are reviewed. Second, the French realizations are assessed with the use of the technology readiness levels in order to identify tracks of improvement. (authors)

New Technological Platform for the National Nuclear Energy Strategy Development

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Rachkov, V. I.

2017-12-01

The paper considers the need to update the development strategy of Russia's nuclear power industry and various approaches to the large-scale nuclear power development. Problems of making decisions on fast neutron reactors and closed nuclear fuel cycle (NFC) arrangement are discussed. The current state of the development of fast neutron reactors and closed NFC technologies in Russia is considered and major problems are highlighted.

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

NASA Astrophysics Data System (ADS)

Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

2010-06-01

In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hetrick, D.L.; Sowers, G.W.

1978-06-01

This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. Amore » list of variable names and a listing for BRENDA are included as appendices.« less

Continuous production of tritium in an isotope-production reactor with a separate circulation system

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium is allowed to flow through the reactor in separate loops in order to facilitate the production and removal of tritium.

Nuclear Rocket Technology Conference

NASA Technical Reports Server (NTRS)

1966-01-01

The Lewis Research Center has a strong interest in nuclear rocket propulsion and provides active support of the graphite reactor program in such nonnuclear areas as cryogenics, two-phase flow, propellant heating, fluid systems, heat transfer, nozzle cooling, nozzle design, pumps, turbines, and startup and control problems. A parallel effort has also been expended to evaluate the engineering feasibility of a nuclear rocket reactor using tungsten-matrix fuel elements and water as the moderator. Both of these efforts have resulted in significant contributions to nuclear rocket technology. Many successful static firings of nuclear rockets have been made with graphite-core reactors. Sufficient information has also been accumulated to permit a reasonable Judgment as to the feasibility of the tungsten water-moderated reactor concept. We therefore consider that this technoIogy conference on the nuclear rocket work that has been sponsored by the Lewis Research Center is timely. The conference has been prepared by NASA personnel, but the information presented includes substantial contributions from both NASA and AEC contractors. The conference excludes from consideration the many possible mission requirements for nuclear rockets. Also excluded is the direct comparison of nuclear rocket types with each other or with other modes of propulsion. The graphite reactor support work presented on the first day of the conference was partly inspired through a close cooperative effort between the Cleveland extension of the Space Nuclear Propulsion Office (headed by Robert W. Schroeder) and the Lewis Research Center. Much of this effort was supervised by Mr. John C. Sanders, chairman for the first day of the conference, and by Mr. Hugh M. Henneberry. The tungsten water-moderated reactor concept was initiated at Lewis by Mr. Frank E. Rom and his coworkers. The supervision of the recent engineering studies has been shared by Mr. Samuel J. Kaufman, chairman for the second day of the conference, and Mr. Roy V. Humble. Dr. John C. Eward served as general chairman for the conference.

Detection and Sizing of Fatigue Cracks in Steel Welds with Advanced Eddy Current Techniques

NASA Astrophysics Data System (ADS)

Todorov, E. I.; Mohr, W. C.; Lozev, M. G.

2008-02-01

Butt-welded specimens were fatigued to produce cracks in the weld heat-affected zone. Advanced eddy current (AEC) techniques were used to detect and size the cracks through a coating. AEC results were compared with magnetic particle and phased-array ultrasonic techniques. Validation through destructive crack measurements was also conducted. Factors such as geometry, surface treatment, and crack tightness interfered with depth sizing. AEC inspection techniques have the potential of providing more accurate and complete sizing flaw data for manufacturing and in-service inspections.

Improved vortex reactor system

DOEpatents

Diebold, James P.; Scahill, John W.

1995-01-01

An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

SU-E-I-48: The Behavior of AEC in Scan Regions Outside the Localizer Radiograph FOV: An In Phantom Study of CT Systems From Four Vendors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Supanich, M; Bevins, N

Purpose: This review of scanners from 4 major manufacturers examines the clinical impact of performing CT scans that extend into areas of the body that were not acquired in the CT localizer radiograph. Methods: Anthropomorphic chest and abdomen phantoms were positioned together on the tables of CT scanners from 4 different vendors. All of the scanners offered an Automatic Exposure Control (AEC) option with both lateral and axial tube current modulation. A localizer radiograph was taken covering the entire extent of both phantoms and then the scanner's Chest-Abdomen-Pelvis (CAP) study was performed with the clinical AEC settings employed and themore » scan and reconstruction range extending from the superior portion of the chest phantom through the inferior portion of the abdomen phantom. A new study was then initiated with a localizer radiograph extending the length of the chest phantom (not covering the abdomen phantom). The same CAP protocol and AEC settings were then used to scan and reconstruct the entire length of both phantoms. Scan parameters at specific locations in the abdomen phantom from both studies were investigated using the information contained in the DICOM metadata of the reconstructed images. Results: The AEC systems on all scanners utilized different tube current settings in the abdomen phantom for the scan completed without the full localizer radiograph. The AEC system behavior was also scanner dependent with the default manual tube current, the maximum tube current and the tube current at the last known position observed as outcomes. Conclusion: The behavior of the AEC systems of CT scanners in regions not covered by the localizer radiograph is vendor dependent. To ensure optimal image quality and radiation exposure it is important to include the entire planned scan region in the localizer radiograph.« less

HEAT TRANSFER AND TRITIUM PRODUCING SYSTEM

DOEpatents

Johnson, E.F.

1962-06-01

This invention related to a circulating lithium-containing blanket system in a neution source hav'ing a magnetic field associated therewith. The blanket serves simultaneously and efficiently as a heat transfer mediunm and as a source of tritium. The blanket is composed of a lithium-6-enriched fused salt selected from the group consisting of lithium nitrite, lithium nitrate, a mixture of said salts, a mixture of each of said salts with lithium oxide, and a mixture of said salts with each other and with lithium oxide. The moderator, which is contained within the blanket in a separate conduit, can be water. A stellarator is one of the neutron sources which can be used in this invention. (AEC)

Research Needs: Prime-Power for High Energy Space Systems.

DTIC Science & Technology

1982-06-01

nuclear sources included both developments from earlier NASA/AEC efforts, such as the present SP-100 program , and also advanced concepts in the form of...and, in some cases, set aside). major programs of applied-research and exploratory- development exist at AFWAL and at NASA labora- tories. The results...decades. In such a research program , the particular I problems of a specific device presently under development are less important than the creation of

Pretreatment with alcoholic extract of Crataegus oxycantha (AEC) activates mitochondrial protection during isoproterenol - induced myocardial infarction in rats.

PubMed

Jayalakshmi, R; Thirupurasundari, C J; Devaraj, S Niranjali

2006-11-01

Crataegus oxycantha (hawthorn) is used in herbal and homeopathic medicine as a cardiotonic. The present study was done to investigate the effect of the alcoholic extract of Crataegus oxycantha (AEC) on mitochondrial function during experimentally induced myocardial infarction in rat. AEC was administered orally to male albino rats (150-200 g), at a dosage of 0.5 ml/100 g body weight/day, for 30 days. At the end of the experimental period, the animals were administered isoproterenol (85 mg/kg body weight, s.c) for 2 days at an interval of 24 h. After 48 h, the rats were anaesthetized and sacrificed. The hearts were homogenized for biochemical and electron microscopic analysis. AEC pretreatment maintained mitochondrial antioxidant status, prevented mitochondrial lipid peroxidative damage and decrease in Kreb's cycle enzymes induced by isoproterenol in rat heart.

Carbon mass balance and microbial ecology in a laboratory scale reactor achieving simultaneous sludge reduction and nutrient removal.

PubMed

Huang, Pei; Li, Liang; Kotay, Shireen Meher; Goel, Ramesh

2014-04-15

Solids reduction in activated sludge processes (ASP) at source using process manipulation has been researched widely over the last two-decades. However, the absence of nutrient removal component, lack of understanding on the organic carbon, and limited information on key microbial community in solids minimizing ASP preclude the widespread acceptance of sludge minimizing processes. In this manuscript, we report simultaneous solids reduction through anaerobiosis along with nitrogen and phosphorus removals. The manuscript also reports carbon mass balance using stable isotope of carbon, microbial ecology of nitrifiers and polyphosphate accumulating organisms (PAOs). Two laboratory scale reactors were operated in anaerobic-aerobic-anoxic (A(2)O) mode. One reactor was run in the standard mode (hereafter called the control-SBR) simulating conventional A(2)O type of activated sludge process and the second reactor was run in the sludge minimizing mode (called the modified-SBR). Unlike other research efforts where the sludge minimizing reactor was maintained at nearly infinite solids retention time (SRT). To sustain the efficient nutrient removal, the modified-SBR in this research was operated at a very small solids yield rather than at infinite SRT. Both reactors showed consistent NH3-N, phosphorus and COD removals over a period of 263 days. Both reactors also showed active denitrification during the anoxic phase even if there was no organic carbon source available during this phase, suggesting the presence of denitrifying PAOs (DNPAOs). The observed solids yield in the modified-SBR was 60% less than the observed solids yield in the control-SBR. Specific oxygen uptake rate (SOUR) for the modified-SBR was almost 44% more than the control-SBR under identical feeding conditions, but was nearly the same for both reactors under fasting conditions. The modified-SBR showed greater diversity of ammonia oxidizing bacteria and PAOs compared to the control-SBR. The diversity of PAOs in the modified-SBR was even more interesting in which case novel clades of Candidatus Accumulibacter phosphatis (CAP), an uncultured but widely found PAOs, were found. Copyright © 2014 Elsevier Ltd. All rights reserved.

Design and testing of a self-actuated shut down system for the protection of liquid metal fast breeder reactors (LMFBRs)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Josephson, J.; Sowa, E.S.

1977-04-01

The design and testing of a simple and reliable Self-Actuated Shutdown System (SASS) for the protection of Liquid Metal Fast Breeder Reactors (LMFBRs) is described. A ferromagnetic Curie temperature permanent magnet holding device has been selected for the design of the Self-Actuated Shutdown System in order to enhance the safety of liquid metal cooled fast reactors (LMFBRs). The self-actuated, self-contained device operates such that accident conditions, resulting in increased coolant temperature or neutron flux reduce the magnetic holding force suspending a neutron absorber above the core by raising the temperature of the trigger mechanism above the Curie point. Neutron absorbermore » material is then inserted into the core, under gravity, terminating the accident. Two possible design variations of the selected concept are presented.« less

The Sustainable Nuclear Future: Fission and Fusion E.M. Campbell Logos Technologies

NASA Astrophysics Data System (ADS)

Campbell, E. Michael

2010-02-01

Global industrialization, the concern over rising CO2 levels in the atmosphere and other negative environmental effects due to the burning of hydrocarbon fuels and the need to insulate the cost of energy from fuel price volatility have led to a renewed interest in nuclear power. Many of the plants under construction are similar to the existing light water reactors but incorporate modern engineering and enhanced safety features. These reactors, while mature, safe and reliable sources of electrical power have limited efficiency in converting fission power to useful work, require significant amounts of water, and must deal with the issues of nuclear waste (spent fuel), safety, and weapons proliferation. If nuclear power is to sustain its present share of the world's growing energy needs let alone displace carbon based fuels, more than 1000 reactors will be needed by mid century. For this to occur new reactors that are more efficient, versatile in their energy markets, require minimal or no water, produce less waste and more robust waste forms, are inherently safe and minimize proliferation concerns will be necessary. Graphite moderated, ceramic coated fuel, and He cooled designs are reactors that can satisfy these requirements. Along with other generation IV fast reactors that can further reduce the amounts of spent fuel and extend fuel resources, such a nuclear expansion is possible. Furthermore, facilities either in early operations or under construction should demonstrate the next step in fusion energy development in which energy gain is produced. This demonstration will catalyze fusion energy development and lead to the ultimate development of the next generation of nuclear reactors. In this presentation the role of advanced fission reactors and future fusion reactors in the expansion of nuclear power will be discussed including synergies with the existing worldwide nuclear fleet. )

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-04-01

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

NASA Astrophysics Data System (ADS)

Guidez, Joel; Saturnin, Anne

2017-11-01

During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

FRAUD/SABOTAGE Killing Nuclear-Reactors!!! ``Super"alloys GENERIC ENDEMIC Wigner's-Disease IN-stability!!!

NASA Astrophysics Data System (ADS)

Asphahani, Aziz; Siegel, Sidney; Siegel, Edward

2010-03-01

Siegel [[J.Mag.Mag.Mtls.7,312(78); PSS(a)11,45(72); Semis.& Insuls.5(79)] (at: ORNL, ANS, Westin``KL"ouse, PSEG, IAEA, ABB) warning of old/new nuclear-reactors/spent-fuel-casks/refineries/ jet/missile/rocket-engines austenitic/FCC Ni/Fe-based (so MIS- called)``super"alloys(182/82;Hastelloy-X; 600;304/304L-SSs; 690 !!!) GENERIC ENDEMIC EXTANT detrimental(synonyms): Wigner's- diseas(WD)[J.Appl.Phys.17,857(46)]; Ostwald-ripening; spinodal- decomposition; overageing-embrittlement; thermomechanical- INstability: Mayo[Google: ``If Leaks Could Kill"; at flickr.com search on ``Giant-Magnotoresistance"; find: [Siegel<<<``Fert"(88) 2007-Nobel/Wolf/Japan-prizes]necessitating NRC inspections on 40+25=65 Westin``KL"ouse PWRs(12/06)]; Lai[Met.Trans.AIME,9A,827 (78)]-Sabol-Stickler[PSS(70)]; Ashpahani[Intl.Conf. H in Metals (77)]; Russell[Prog. Mtls.Sci.(83)]; Pollard[last UCS rept. (9/95)]; Lofaro[BNL/DOE/NRC Repts.]; Pringle[Nuclear-Power:From Physics to Politics(79)]; Hoffman[animatedsoftware.com],...what DOE/NRC MISlabels as ``butt-welds" ``stress-corrosion cracking" endpoint's ROOT-CAUSE ULTIMATE-ORIGIN is WD overageing-embrit- tlement caused brittle-fracture cracking from early/ongoing AEC/DOE-n``u''tional-la``v''atories sabotage!!!

FRAUD/SABOTAGE Killing Nuclear-Reactors Need Modeling!!!: "Super"alloys GENERIC ENDEMIC Wigner's-Disease/.../IN-stability: Ethics? SHMETHICS!!!

NASA Astrophysics Data System (ADS)

Asphahani, Aziz; Siegel, Sidney; Siegel, Edward

2010-03-01

Carbides solid-state chemistry domination of old/new nuclear- reactors/spent-fuel-casks/refineries/jet/missile/rocket-engines in austenitic/FCC Ni/Fe-based(so miscalled)``super"alloys(182/82; Hastelloy-X,600,304/304L-SSs,...,690!!!) GENERIC ENDEMIC EXTANT detrimental(synonyms): Wigner's-diseas(WD)[J.Appl.Phys.17,857 (1946)]/Ostwald-ripening/spinodal-decomposition/overageing- embrittlement/thermal-leading-to-mechanical(TLTM)-INstability: Mayo[Google:``If Leaks Could Kill"; at flickr.com search on ``Giant-Magnotoresistance"; find: Siegel[J.Mag.Mag.Mtls.7,312 (1978)]<<<``Fert"-"Gruenberg"(1988/89)2007-physics Nobel/Wolf/ Japan-prizes]necessitating NRC-inspections of 40+25 = 65 Westin- ``KLouse PWRs(12/2006)]-Lai[Met.Trans.AIME,9A,827(1978)]-Sabol- Stickler[Phys.Stat.Sol.(1970)]-Ashpahani[Intl.Conf. H in Metals, Paris(1977]-Russell[Prog.Mtls.Sci.(1983)]-Pollard[last UCS rept. (9/1995)]-Lofaro[BNL/DOE/NRC Repts.]-Pringle[Nuclear-Power:From Physics to Politics(1979)]-Hoffman[animatedsoftware.com], what DOE/NRC MISlabels as ``butt-welds" ``stress-corrosion cracking" endpoint's ROOT-CAUSE ULTIMATE-ORIGIN is WD overageing-embritt- lement caused brittle-fracture cracking from early/ongoing AEC/ DOE-n"u"tional-la"v"atories sabotage!!!

FRAUD/SABOTAGE Killing Nuclear-Reactors Need Modeling!!!: ``Super'' alloys GENERIC ENDEMIC Wigner's-Disease/.../IN-stability: Ethics? SHMETHICS!!!

NASA Astrophysics Data System (ADS)

O'Grady, Joseph; Bument, Arlden; Siegel, Edward

2011-03-01

Carbides solid-state chemistry domination of old/new nuclear-reactors/spent-fuel-casks/refineries/jet/missile/rocket-engines is austenitic/FCC Ni/Fe-based (so miscalled)"super"alloys(182/82;Hastelloy-X,600,304/304L-SSs,...690!!!) GENERIC ENDEMIC EXTANT detrimental(synonyms): Wigner's-disease(WD) [J.Appl.Phys.17,857 (46)]/Ostwald-ripening/spinodal-decomposition/overageing-embrittlement/thermal-leading-to-mechanical(TLTM)-INstability: Mayo[Google: fLeaksCouldKill > ; - Siegel [ J . Mag . Mag . Mtls . 7 , 312 (78) = atflickr . comsearchonGiant - Magnotoresistance [Fert" [PRL(1988)]-"Gruenberg"[PRL(1989)] 2007-Nobel]necessitating NRC inspections on 40+25=65 Westin"KL"ouse PWRs(12/2006)]-Lai [Met.Trans.AIME, 9A,827(78)]-Sabol-Stickler[Phys.Stat.Sol.(70)]-Ashpahani[ Intl.Conf. Hydrogen in Metals, Paris(1977]-Russell [Prog.Mtls.Sci.(1983)]-Pollard [last UCS rept.(9/1995)]-Lofaro [BNL/DOE/NRC Repts.]-Pringle [ Nuclear-Power:From Physics to Politics(1979)]-Hoffman [animatedsoftware.com], what DOE/NRC MISlabels as "butt-welds" "stress-corrosion cracking" endpoint's ROOT-CAUSE ULTIMATE-ORIGIN is WD overageing-embrittlement caused brittle-fracture cracking from early/ongoing AEC/DOE-n"u"tional-la"v"atories sabotage!!!

A Spherical Torus Nuclear Fusion Reactor Space Propulsion Vehicle Concept for Fast Interplanetary Travel

NASA Technical Reports Server (NTRS)

Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

1998-01-01

A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a greater than 5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all ma or systems including payload, central truss, nuclear reactor (including divertor and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, power utilization, and component design.

Fast Breeder Reactors in Sweden: Vision and Reality.

PubMed

Fjaestad, Maja

2015-01-01

The fast breeder is a type of nuclear reactor that aroused much attention in the 1950s and '60s. Its ability to produce more nuclear fuel than it consumes offered promises of cheap and reliable energy. Sweden had advanced plans for a nuclear breeder program, but canceled them in the middle of the 1970s with the rise of nuclear skepticism. The article investigates the nuclear breeder as a technological vision. The nuclear breeder reactor is an example of a technological future that did not meet its industrial expectations. But that does not change the fact that the breeder was an influential technology. Decisions about the contemporary reactors were taken with the idea that in a foreseeable future they would be replaced with the efficient breeder. The article argues that general themes in the history of the breeder reactor can deepen our understanding of the mechanisms behind technological change.

Characterization of fast neutron spectrum in the TRIGA for hardness testing of electronic components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nelson, George W.

1986-07-01

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering Laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiographymore » and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems. (author)« less

Computed tomography automatic exposure control techniques in 18F-FDG oncology PET-CT scanning.

PubMed

Iball, Gareth R; Tout, Deborah

2014-04-01

Computed tomography (CT) automatic exposure control (AEC) systems are now used in all modern PET-CT scanners. A collaborative study was undertaken to compare AEC techniques of the three major PET-CT manufacturers for fluorine-18 fluorodeoxyglucose half-body oncology imaging. An audit of 70 patients was performed for half-body CT scans taken on a GE Discovery 690, Philips Gemini TF and Siemens Biograph mCT (all 64-slice CT). Patient demographic and dose information was recorded and image noise was calculated as the SD of Hounsfield units in the liver. A direct comparison of the AEC systems was made by scanning a Rando phantom on all three systems for a range of AEC settings. The variation in dose and image quality with patient weight was significantly different for all three systems, with the GE system showing the largest variation in dose with weight and Philips the least. Image noise varied with patient weight in Philips and Siemens systems but was constant for all weights in GE. The z-axis mA profiles from the Rando phantom demonstrate that these differences are caused by the nature of the tube current modulation techniques applied. The mA profiles varied considerably according to the AEC settings used. CT AEC techniques from the three manufacturers yield significantly different tube current modulation patterns and hence deliver different doses and levels of image quality across a range of patient weights. Users should be aware of how their system works and of steps that could be taken to optimize imaging protocols.

CARMA3 Is Critical for the Initiation of Allergic Airway Inflammation

PubMed Central

Causton, Benjamin; Ramadas, Ravisankar A.; Cho, Josalyn L.; Jones, Khristianna; Pardo-Saganta, Ana; Rajagopal, Jayaraj; Xavier, Ramnik J.

2015-01-01

Innate immune responses to allergens by airway epithelial cells (AECs) help initiate and propagate the adaptive immune response associated with allergic airway inflammation in asthma. Activation of the transcription factor NF-κB in AECs by allergens or secondary mediators via G protein–coupled receptors (GPCRs) is an important component of this multifaceted inflammatory cascade. Members of the caspase recruitment domain family of proteins display tissue-specific expression and help mediate NF-κB activity in response to numerous stimuli. We have previously shown that caspase recruitment domain–containing membrane-associated guanylate kinase protein (CARMA)3 is specifically expressed in AECs and mediates NF-κB activation in these cells in response to stimulation with the GPCR agonist lysophosphatidic acid. In this study, we demonstrate that reduced levels of CARMA3 in normal human bronchial epithelial cells decreases the production of proasthmatic mediators in response to a panel of asthma-relevant GPCR ligands such as lysophosphatidic acid, adenosine triphosphate, and allergens that activate GPCRs such as Alternaria alternata and house dust mite. We then show that genetically modified mice with CARMA3-deficient AECs have reduced airway eosinophilia and proinflammatory cytokine production in a murine model of allergic airway inflammation. Additionally, we demonstrate that these mice have impaired dendritic cell maturation in the lung and that dendritic cells from mice with CARMA3-deficient AECs have impaired Ag processing. In conclusion, we show that AEC CARMA3 helps mediate allergic airway inflammation, and that CARMA3 is a critical signaling molecule bridging the innate and adaptive immune responses in the lung. PMID:26041536

Gastroprotective Effect of Freeze Dried Stripped Snakehead Fish (Channa striata Bloch.) Aqueous Extract against Aspirin Induced Ulcerogenesis in Pylorus Ligated Rats.

PubMed

Ali Khan, Mohammed Safwan; Mat Jais, Abdul Manan; Hussain, Javeed; Siddiqua, Faiza; Gopala Reddy, A; Shivakumar, P; Madhuri, D

2014-01-01

Channa striata (Bloch.) is a fresh water fish belonging to the family Channidae. The stripped snakehead fish possesses wide range of medicinal properties. In view of traditional use of C. striata for wound healing, the present study was undertaken to investigate the beneficial effects of orally administered freeze dried aqueous extract of Channa striata (AECS) in experimentally induced gastric ulcers in Wistar rats. Aspirin induced ulcerogenesis in pyloric ligation model was used for the assessment of antiulcer activity and Ranitidine (50 mg/kg) was employed as the standard drug. The various gastric parameters like volume of gastric juice, pH, free and total acidities, ulcer index, and levels of antioxidant enzymes like catalase, superoxide dismutase, and lipid peroxidation marker malondialdehyde were determined. AECS at concentrations of 40% and 50% w/v significantly decreased the volume of gastric juice and increased the levels of catalase while considerable decrease in free and total acidities and increase in superoxide dismutase were observed with the treatment of standard drug and AECS (50% w/v). All the test doses of AECS markedly decreased ulcer index and malondialdehyde compared to the standard drug whereas AECS 30% w/v did not alter volume of gastric juice, pH, free and total acidities, catalase, and superoxide dismutase. From these findings, it can be concluded that AECS is devoid of acid neutralizing effects at lower doses and possesses antisecretory and antiulcer activities and this could be related to its antioxidant mechanism.

Steam generator for liquid metal fast breeder reactor

DOEpatents

Gillett, James E.; Garner, Daniel C.; Wineman, Arthur L.; Robey, Robert M.

1985-01-01

Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

NASA Astrophysics Data System (ADS)

Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

2017-01-01

The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

Radiogenic lead as coolant, reflector and moderator in advanced fast reactors

NASA Astrophysics Data System (ADS)

Kulikov, E. G.

2017-01-01

Main purpose of the study is assessing reasonability for recovery, production and application of radiogenic lead as a coolant, neutron moderator and neutron reflector in advanced fast reactors. When performing the study, thermal, physical and neutron-physical properties of natural and radiogenic lead were analyzed. The following results were obtained: 1. Radiogenic lead with high content of isotope 208Pb can be extracted from thorium or mixed thorium-uranium ores because 208Pb is a final product of 232Th natural decay chain. 2. The use of radiogenic lead with high 208Pb content in advanced fast reactors and accelerator-driven systems (ADS) makes it possible to improve significantly their neutron-physical and thermal-hydraulic parameters. 3. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a coolant opens the possibilities for more intense fuel breeding and for application of well-known oxide fuel instead of the promising but not tested enough nitride fuel under the same safety parameters. 4. The use of radiogenic lead with high 208Pb content in ADS as a coolant can upgrade substantially the level of neutron flux in the ADS blanket, which enables effective transmutation of radioactive wastes with low cross-sections of radiative neutron capture.

Strongyloidiasis Epidemiology and Treatment Response in Patients with HIV Infection

PubMed Central

Cortes-Penfield, Nicolas; Moore, Cody; Arduino, Roberto; Serpa, Jose

2017-01-01

Abstract Background We sought to characterize the epidemiology of HIV and S. stercoralis coinfection in an urban HIV cohort, and to investigate the effect of S. stercoralis infection on HIV virologic control and immune recovery. Methods We reviewed the medical records of all HIV-infected patients diagnosed with strongyloidiasis who received care at Thomas Street Health Center (Houston, TX) between 2000 and 2015. For each case we included up to two matched HIV-infected patients without strongyloidiasis (controls). Matching was based on age, sex, ethnicity, baseline CD4 percentage, and HIV viral load at the time of strongyloidiasis diagnosis in the case patient. We recorded patient demographics, comorbidities, CD4 count and percentage, HIV viral load, and absolute eosinophilia count (AEC) at the time of HIV diagnosis, strongyloidiasis diagnosis, and six and twelve months after ivermectin treatment. Results We identified 15 cases of HIV and S.stercoralis coinfection; 13 had at least one available matched control. The mean age of coinfected patients was 45; all were Hispanic, 84.6% were male, and the mean CD4 nadir was 146 cells/ul. At the time of strongyloidiasis diagnosis, the mean CD4 count was 460 cells/ul, HIV RNA viral load 2.07 logs/ml, and AEC was 1,360 cells/μL. At 6 and 12 months after treatment, CD4 counts were 514 and 464 cells/μL, HIV RNA viral loads 1.78 and 2.31 log/mL, and AECs 319 and 362 cells/μL, respectively. Although CD4 counts increased 6 months after treatment, they returned to baseline levels at 12 months; neither change achieved statistical significance. The reduction in AECs after ivermectin treatment was statistically significant (P < 0.001). Matched controls without S.stercoralis had lower AECs at baseline, 6 months, and 12 months; otherwise, there were no differences between cases and controls. Conclusion Strongyloidiasis treatment in HIV-infected patients led to normalization of the AEC at 6 months in most cases, but AECs remained higher than in control patients. Persistently elevated AECs may suggest treatment failure or reinfection. Our study was unable to identify any effect of S. stercoralis infection or treatment on HIV virologic suppression or immunologic recovery; larger studies are warranted to investigate the effect of strongyloidiasis on HIV disease. Disclosures All authors: No reported disclosures.

Improved vortex reactor system

DOEpatents

Diebold, J.P.; Scahill, J.W.

1995-05-09

An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

PREPARATION OF UO$sub 2$ FOR NUCLEAR REACTOR FUEL PELLETS

DOEpatents

Googin, J.M.

1962-06-01

A method is given for preparing high-density UO/sub 2/ compacts. An aqueous uranyl fluoride solution is contacted with an aqueous ammonium hydroxide solution at an ammonium to-uranium ratio of 25: 1 to 30:1 to form a precipitate. The precipitate is separated from the- mother liquor, dried, and contacted with steam at a uniform temperature within the range of 400 to 650 deg C to produce U/ sub 3/O/sub 8/. The U/sub 3/O/sub 8/ is red uced to UO/sub 2/ with hydrogen at a uniform temperature within the range of 550 to 600 deg C. The UO/sub 2/ is then compressed into compacts and sintered. High-density compacts are fabricated to close tolerances without use of a binder and without machining or grinding. (AEC)

PREPARATION OF REFRACTORY OXIDE MICROSPHERE

DOEpatents

Haws, C.C. Jr.

1963-09-24

A method is described of preparing thorium oxide in the form of fused spherical particles about 1 to 2 microns in diameter. A combustible organic solution of thorium nitrate containing additive metal values is dispersed into a reflected, oxygen-fed flame at a temperature above the melting point of the resulting oxide. The metal additive is aluminum at a proportion such as to provide 1 to 10 weight per cent aluminum oxide in the product, silicon at the same proportion, or beryllium at a proportion of 12 to 25 weight per cent beryllium oxide in the product. A minor proportion of uranium values may also be provided in the solution. The metal additive lowers the oxide melting point and allows fusion and sphere formation in conventional equipment. The product particles are suitable for use in thorium oxide slurries for nuclear reactors. (AEC)

GRIPPING TOOL

DOEpatents

Sandrock, R.J.

1961-12-12

A self-actuated gripping tool is described for transferring fuel elements and the like into reactors and other inaccessible locations. The tool will grasp or release the load only when properly positioned for this purpose. In addition, the load cannot be released except when unsupported by the tool, so that jarring or contact will not bring about accidental release of the load. The gripping members or jaws of the device are cam-actuated by an axially slidable shaft which has two lockable positions. A spring urges the shaft into one position and a solenoid is provided to overcome the spring and move it into the other position. The weight of the tool operates a sleeve to lock the shaft in its existing position. Only when the cable supporting the tool is slack is the device capable of being actuated either to grasp or release its load. (AEC)

R and D program for core instrumentation improvements devoted for French sodium fast reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeannot, J. P.; Rodriguez, G.; Jammes, C.

2011-07-01

Under the framework of French R and D studies for Generation IV reactors and more specifically for sodium-cooled fast reactors (SFR); the CEA, EDF and AREVA have launched a joint coordinated research programme. This paper deals with the R and D sets out to achieve better inspection, maintenance, availability and decommissioning. In particular the instrumentation requirements for core monitoring and detection in the case of accidental events. Requirements mainly involve diversifying the means of protection and improving instrumentation performance in terms of responsiveness and sensitivity. Operation feedback from the Phenix and Superphenix prototype reactors and studies, carried out within themore » scope of the EFR projects, has been used to define the needs for instrumentation enhancement. (authors)« less

FY2017 Updates to the SAS4A/SASSYS-1 Safety Analysis Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.

The SAS4A/SASSYS-1 safety analysis software is used to perform deterministic analysis of anticipated events as well as design-basis and beyond-design-basis accidents for advanced fast reactors. It plays a central role in the analysis of U.S. DOE conceptual designs, proposed test and demonstration reactors, and in domestic and international collaborations. This report summarizes the code development activities that have taken place during FY2017. Extensions to the void and cladding reactivity feedback models have been implemented, and Control System capabilities have been improved through a new virtual data acquisition system for plant state variables and an additional Block Signal for a variablemore » lag compensator to represent reactivity feedback for novel shutdown devices. Current code development and maintenance needs are also summarized in three key areas: software quality assurance, modeling improvements, and maintenance of related tools. With ongoing support, SAS4A/SASSYS-1 can continue to fulfill its growing role in fast reactor safety analysis and help solidify DOE’s leadership role in fast reactor safety both domestically and in international collaborations.« less

Response of six neutron survey meters in mixed fields of fast and thermal neutrons.

PubMed

Kim, S I; Kim, B H; Chang, I; Lee, J I; Kim, J L; Pradhan, A S

2013-10-01

Calibration neutron fields have been developed at KAERI (Korea Atomic Energy Research Institute) to study the responses of commonly used neutron survey meters in the presence of fast neutrons of energy around 10 MeV. The neutron fields were produced by using neutrons from the (241)Am-Be sources held in a graphite pile and a DT neutron generator. The spectral details and the ambient dose equivalent rates of the calibration fields were established, and the responses of six neutron survey meters were evaluated. Four single-moderator-based survey meters exhibited an under-responses ranging from ∼9 to 55 %. DINEUTRUN, commonly used in fields around nuclear reactors, exhibited an over-response by a factor of three in the thermal neutron field and an under-response of ∼85 % in the mixed fields. REM-500 (tissue-equivalent proportional counter) exhibited a response close to 1.0 in the fast neutron fields and an under-response of ∼50 % in the thermal neutron field.

Pulmonary Regnase-1 orchestrates the interplay of epithelium and adaptive immune systems to protect against pneumonia.

PubMed

Nakatsuka, Yoshinari; Vandenbon, Alexis; Mino, Takashi; Yoshinaga, Masanori; Uehata, Takuya; Cui, Xiaotong; Sato, Ayuko; Tsujimura, Tohru; Suzuki, Yutaka; Sato, Atsuyasu; Handa, Tomohiro; Chin, Kazuo; Sawa, Teiji; Hirai, Toyohiro; Takeuchi, Osamu

2018-04-25

Inhaled pathogens including Pseudomonas aeruginosa initially encounter airway epithelial cells (AECs), which are poised to evoke cell-intrinsic innate defense, affecting second tier of hematopoietic cell-mediated immune reaction. However, it is largely unknown how pulmonary immune responses mediated by a variety of immune cells are coordinated. Here we show that Regnase-1, an endoribonuclease expressed in AECs and immune cells, plays an essential role in coordinating innate responses and adaptive immunity against P. aeruginosa infection. Intratracheal treatment of mice with heat-killed P. aeruginosa resulted in prolonged disappearance of Regnase-1 consistent with sustained expression of Regnase-1 target inflammatory genes, whereas the transcription factor NF-κB was only transiently activated. AEC-specific deletion of Regnase-1 not only augmented innate defenses against P. aeruginosa but also enhanced secretion of Pseudomonas-specific IgA and Th17 accumulation in the lung, culminating in conferring significant resistance against P. aeruginosa re-infection in vivo. Although Regnase-1 directly controls distinct sets of genes in each of AECs and T cells, degradation of Regnase-1 in both cell types is beneficial for maximizing acquired immune responses. Collectively, these results demonstrate that Regnase-1 orchestrates AEC-mediated and immune cell-mediated host defense against pulmonary bacterial infection.

The Role of Mitochondrial DNA in Mediating Alveolar Epithelial Cell Apoptosis and Pulmonary Fibrosis

PubMed Central

Kim, Seok-Jo; Cheresh, Paul; Jablonski, Renea P.; Williams, David B.; Kamp, David W.

2015-01-01

Convincing evidence has emerged demonstrating that impairment of mitochondrial function is critically important in regulating alveolar epithelial cell (AEC) programmed cell death (apoptosis) that may contribute to aging-related lung diseases, such as idiopathic pulmonary fibrosis (IPF) and asbestosis (pulmonary fibrosis following asbestos exposure). The mammalian mitochondrial DNA (mtDNA) encodes for 13 proteins, including several essential for oxidative phosphorylation. We review the evidence implicating that oxidative stress-induced mtDNA damage promotes AEC apoptosis and pulmonary fibrosis. We focus on the emerging role for AEC mtDNA damage repair by 8-oxoguanine DNA glycosylase (OGG1) and mitochondrial aconitase (ACO-2) in maintaining mtDNA integrity which is important in preventing AEC apoptosis and asbestos-induced pulmonary fibrosis in a murine model. We then review recent studies linking the sirtuin (SIRT) family members, especially SIRT3, to mitochondrial integrity and mtDNA damage repair and aging. We present a conceptual model of how SIRTs modulate reactive oxygen species (ROS)-driven mitochondrial metabolism that may be important for their tumor suppressor function. The emerging insights into the pathobiology underlying AEC mtDNA damage and apoptosis is suggesting novel therapeutic targets that may prove useful for the management of age-related diseases, including pulmonary fibrosis and lung cancer. PMID:26370974

Biodiesel production process intensification using a rotor-stator type generator of hydrodynamic cavitation.

PubMed

Crudo, Daniele; Bosco, Valentina; Cavaglià, Giuliano; Grillo, Giorgio; Mantegna, Stefano; Cravotto, Giancarlo

2016-11-01

Triglyceride transesterification for biodiesel production is a model reaction which is used to compare the conversion efficiency, yield, reaction time, energy consumption, scalability and cost estimation of different reactor technology and energy source. This work describes an efficient, fast and cost-effective procedure for biodiesel preparation using a rotating generator of hydrodynamic cavitation (HC). The base-catalyzed transesterification (methanol/sodium hydroxide) has been carried out using refined and bleached palm oil and waste vegetable cooking oil. The novel HC unit is a continuous rotor-stator type reactor in which reagents are directly fed into the controlled cavitation chamber. The high-speed rotation of the reactor creates micron-sized droplets of the immiscible reacting mixture leading to outstanding mass and heat transfer and enhancing the kinetics of the transesterification reaction which completes much more quickly than traditional methods. All the biodiesel samples obtained respect the ASTM standard and present fatty acid methyl ester contents of >99% m/m in both feedstocks. The electrical energy consumption of the HC reactor is 0.030kWh per L of produced crude biodiesel, making this innovative technology really quite competitive. The reactor can be easily scaled-up, from producing a few hundred to thousands of liters of biodiesel per hour while avoiding the risk of orifices clogging with oil impurities, which may occur in conventional HC reactors. Furthermore it requires minimal installation space due to its compact design, which enhances overall security. Copyright © 2016 Elsevier B.V. All rights reserved.

NEUTRONIC REACTOR

DOEpatents

Wade, E.J.

1958-09-16

This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

Trench fast reactor design using the microcomputer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohach, A.F.; Sankoorikal, J.T.; Schmidt, R.R.

1987-01-01

This project is a study of alternative liquid-metal-cooled fast power reactor system concepts. Specifically, an unconventional primary system is being conceptually designed and evaluated. The project design is based primarily on microcomputer analysis through the use of computational modules. The reactor system concept is a long, narrow pool with a long, narrow reactor called a trench-type pool reactor in it. The reactor consists of five core-blanket modules in a line. Specific power is to be modest, permitting long fuel residence time. Two fuel cycles are currently being considered. The reactor design philosophy is that of the inherently safe concept. Thismore » requires transient analysis dependent on reactivity coefficients: prompt fuel, including Doppler and expansion, fuel expansion, sodium temperature and void, and core expansion. Conceptual reactor design is done on a microcomputer. A part of the trench reactor project is to develop a microcomputer-based system that can be used by the user for scoping studies and design. Current development includes the neutronics and fuel management aspects of the design. Thermal-hydraulic analysis and economics are currently being incorporated into the microcomputer system. The system is menu-driven including preparation of program input data and of output data for displays in graphics form.« less

The combined hybrid system: A symbiotic thermal reactor/fast reactor system for power generation and radioactive waste toxicity reduction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hollaway, W.R.

1991-08-01

If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issuemore » through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MW{sub e} IFR capacity for every three MW{sub e} Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years).« less

Characterisation of an accelerator-based neutron source for BNCT versus beam energy

NASA Astrophysics Data System (ADS)

Agosteo, S.; Curzio, G.; d'Errico, F.; Nath, R.; Tinti, R.

2002-01-01

Neutron capture in 10B produces energetic alpha particles that have a high linear energy transfer in tissue. This results in higher cell killing and a higher relative biological effectiveness compared to photons. Using suitably designed boron compounds which preferentially localize in cancerous cells instead of healthy tissues, boron neutron capture therapy (BNCT) has the potential of providing a higher tumor cure rate within minimal toxicity to normal tissues. This clinical approach requires a thermal neutron source, generally a nuclear reactor, with a fluence rate sufficient to deliver tumorcidal doses within a reasonable treatment time (minutes). Thermal neutrons do not penetrate deeply in tissue, therefore BNCT is limited to lesions which are either superficial or otherwise accessible. In this work, we investigate the feasibility of an accelerator-based thermal neutron source for the BNCT of skin melanomas. The source was designed via MCNP Monte Carlo simulations of the thermalization of a fast neutron beam, generated by 7 MeV deuterons impinging on a thick target of beryllium. The neutron field was characterized at several deuteron energies (3.0-6.5 MeV) in an experimental structure installed at the Van De Graaff accelerator of the Laboratori Nazionali di Legnaro, in Italy. Thermal and epithermal neutron fluences were measured with activation techniques and fast neutron spectra were determined with superheated drop detectors (SDD). These neutron spectrometry and dosimetry studies indicated that the fast neutron dose is unacceptably high in the current design. Modifications to the current design to overcome this problem are presented.

An Ultrasonic Scanning Technique for In-Situ `Bowing' Measurement of Prototype Fast Breeder Reactor Fuel Sub-Assembly

NASA Astrophysics Data System (ADS)

Swaminathan, K.; Asokane, C.; Sylvia, J. I.; Kalyanasundaram, P.; Swaminathan, P.

2012-02-01

An ultrasonic under-sodium scanner has been developed for deployment in Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India. Its purpose is to scan the above-core plenum for detection, if any, of displacement of sub-assemblies. During its burn-up in the reactor, the head of a Fuel Sub-Assembly (FSA) may undergo a lateral shift from its original position (called `bowing') due to the fast neutron induced damage on its structural material. A simple scanning technique has been developed for measuring the extent of bowing in-situ. This paper describes a PC-controlled mock-up of the scanner used to implement the scanning technique and the results obtained of scanning a mock-up FSA head under water. The details of the liquid-sodium proof transducer developed for use in the PFBR scanner and its performance are also discussed.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bustraan, M.; Coehoorn, J.; Veenema, J.J.

This report is a collection of separate contributions on some aspects of the work done on STEK up to February 1970. A description is given of STEK together with the philosophy of its design, i.e. integral measurements of fission product cross sections by a sample oscillator technique in fast reactor spectra. The influences fission products may have on fast breeder reactors are briefly demonstrated by an example. A description of the facility and of the sample oscillator and sample exchange mechanism is given. Some preliminary results of measurements of reactor parameters and the neutron spectrum in the first fast zonemore » in STEK are given. For the use of lead as material for the buffer an argumentation is given. The proposed program for the measurements of the integral fission product cross sections is outlined. The procurement of some, highly active, samples of actual fission products is briefly sketched. (auth)« less

Estimating the Temperature Experienced by Biomass Particles during Fast Pyrolysis Using Microscopic Analysis of Biochars

DOE PAGES

Thompson, Logan C.; Ciesielski, Peter N.; Jarvis, Mark W.; ...

2017-07-12

Here, biomass particles can experience variable thermal conditions during fast pyrolysis due to differences in their size and morphology, and from local temperature variations within a reactor. These differences lead to increased heterogeneity of the chemical products obtained in the pyrolysis vapors and bio-oil. Here we present a simple, high-throughput method to investigate the thermal history experienced by large ensembles of particles during fast pyrolysis by imaging and quantitative image analysis. We present a correlation between the surface luminance (darkness) of the biochar particle and the highest temperature that it experienced during pyrolysis. Next, we apply this correlation to large,more » heterogeneous ensembles of char particles produced in a laminar entrained flow reactor (LEFR). The results are used to interpret the actual temperature distributions delivered by the reactor over a range of operating conditions.« less