An Innovative Hybrid Loop-Pool SFR Design and Safety Analysis Methods: Today and Tomorrow

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

Hongbin Zhang; Haihua Zhao; Vincent Mousseau

2008-04-01

Investment in commercial sodium cooled fast reactor (SFR) power plants will become possible only if SFRs achieve economic competitiveness as compared to light water reactors and other Generation IV reactors. Toward that end, we have launched efforts to improve the economics and safety of SFRs from the thermal design and safety analyses perspectives at Idaho National Laboratory. From the thermal design perspective, an innovative hybrid loop-pool SFR design has been proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to further improve economics and safety. From the safety analysesmore » perspective, we have initiated an effort to develop a high fidelity reactor system safety code.« less

75 FR 56597 - University of Wisconsin; University of Wisconsin Nuclear Reactor Environmental Assessment and...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-16

...,000 gallons being typical. The licensee maintains a pool leak surveillance program. The pool water leak surveillance program continues to monitor the pool water evaporation rate, the pool water make-up volume, and pool water radioactivity. The pool leak surveillance program indicated that approximately 2...

Emergency core cooling system

DOEpatents

Schenewerk, William E.; Glasgow, Lyle E.

1983-01-01

A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.

A new safety channel based on ¹⁷N detection in research reactors.

PubMed

Seyfi, Somayye; Gharib, Morteza

2015-10-01

Tehran research reactor (TRR) is a representative of pool type research reactors using light water, as coolant and moderator. This reactor is chosen as a prototype to demonstrate and prove the feasibility of (17)N detection as a new redundant channel for reactor power measurement. In TRR, similar to other pool type reactors, neutron detectors are immersed in the pool around the core as the main power measuring devices. In the present article, a different approach, using out of water neutron detector, is employed to measure reactor power. This new method is based on (17)O (n,p) (17)N reaction taking place inside the core and subsequent measurement of delayed neutrons emitted due to (17)N disintegration. Count and measurement of neutrons around outlet water pipe provides a reliable redundant safety channel to measure reactor power. Results compared with other established channels indicate a good agreement and shows a linear interdependency with true thermal power. Safety of reactor operation is improved with installation & use of this new power measuring channel. The new approach may equally serve well as a redundant channel in all other types of reactors having coolant comprised of oxygen in its molecular constituents. Contrary to existing channels, this one is totally out of water and thus is an advantage over current instrumentations. It is proposed to employ the same idea on other reactors (nuclear power plants too) to improve safety criteria. Copyright © 2015 Elsevier Ltd. All rights reserved.

The Muon System of the Daya Bay Reactor Antineutrino Experiment

DOE PAGES

An, F. P.; Hackenburg, R. W.; Brown, R. E.; ...

2014-10-05

The Daya Bay experiment consists of functionally identical antineutrino detectors immersed in pools of ultrapure water in three well-separated underground experimental halls near two nuclear reactor complexes. These pools serve both as shields against natural, low-energy radiation, and as water Cherenkov detectors that efficiently detect cosmic muons using arrays of photomultiplier tubes. Each pool is covered by a plane of resistive plate chambers as an additional means of detecting muons. Design, construction, operation, and performance of these muon detectors are described. (auth)

Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

NASA Astrophysics Data System (ADS)

Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

2010-12-01

Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

PWR upper/lower internals shield

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

Homyk, W.A.

1995-03-01

During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use ofmore » lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.« less

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA660, INTERIOR. REACTOR INSIDE TANK. METAL ...

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

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA-660, INTERIOR. REACTOR INSIDE TANK. METAL WORK PLATFORM ABOVE. THE REACTOR WAS IN A SMALL WATER-FILLED POOL. INL NEGATIVE NO. 66-6373. Unknown Photographer, ca. 1966 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

LPT. EBOR (TAN646) interior, installing reactor in STF pool ("vault"). ...

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

LPT. EBOR (TAN-646) interior, installing reactor in STF pool ("vault"). Pressure vessel shows core barrel and outlet nozzle (next to man below) to inner duct weld, which is prepared and in position for stress relieving. Camera facing southeast. Photographer: Comiskey. Date: January 20, 1965. INEEL negative no. 65-239 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

NUCLEAR REACTOR CONTROL SYSTEM

DOEpatents

Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

1959-11-01

A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

Conceptual design of BNCT facility based on the TRR medical room

NASA Astrophysics Data System (ADS)

Golshanian, M.; Rajabi, A. A.; Kasesaz, Y.

2017-10-01

This paper presents a conceptual design of the Boron Neutron Capture Therapy (BNCT) facility based on the medical room of Tehran Research Reactor (TRR). The medical room is located behind the east wall of the reactor pool. The designed beam line is an in-pool Beam Shaping Assembly (BSA) which is considered between the reactor core and the medical room wall. The final designed BSA can provide 2.96× 109 n/cm2ṡs epithermal neutron flux at the irradiation position with acceptable beam contamination to use as a clinical BNCT.

Development work for a borax internal core-catcher for a gas-cooled fast reactor

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

Donne, M.D.; Dorner, S.; Schumacher, G.

1978-07-01

Preliminary thermal calculations show that a corecatcher, which is able to cope with the complete meltdown of the core and blankets of a 1000-MW(electric) gas-cooled fast reactor, appears to be feasible. This core-catcher is based on borax (Na/sub 2/B/sub 4/O/sub 7/) dissolving the oxide fuel and the fission products occurring in oxide form. The borax is contained in steel boxes forming a 2.2-m-thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel (PCRV), just underneath the reactor core. After a complete meltdown accident, the fission products, in oxide form, are dispersed in the pool formedmore » by the liquid borax. The metallic fission products are contained in the steel lying below the borax pool and in contact with the water-cooled PCRV liner. The volumetric power density of the molten core is conveniently reduced as it is dissolved in the borax, and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system.« less

Method development and validation for simultaneous determination of IEA-R1 reactor’s pool water uranium and silicon content by ICP OES

NASA Astrophysics Data System (ADS)

Ulrich, J. C.; Guilhen, S. N.; Cotrim, M. E. B.; Pires, M. A. F.

2018-03-01

IPEN’s research reactor, IEA-R1, an open pool type research reactor moderated and cooled by light water. High quality water is a key factor in preventing the corrosion of the spent fuel stored in the pool. Leaching of radionuclides from the corroded fuel cladding may be prevented by an efficient water treatment and purification system. However, as a safety management policy, IPEN has adopted a water chemistry control which periodically monitors the levels of uranium (U) and silicon (Si) in the pool’s reactor, since IEA-R1 employs U3Si2-Al dispersion fuel. An analytical method was developed and validated for the determination of uranium and silicon by ICP OES. This work describes the validation process, in a context of quality assurance, including the parameters selectivity, linearity, quantification limit, precision and recovery.

System and process for the production of syngas and fuel gasses

DOEpatents

Bingham, Dennis N.; Kllingler, Kerry M.; Turner, Terry D.; Wilding, Bruce M.; Benefiel, Bradley C.

2014-04-01

The production of gasses and, more particularly, to systems and methods for the production of syngas and fuel gasses including the production of hydrogen are set forth. In one embodiment system and method includes a reactor having a molten pool of a material comprising sodium carbonate. A supply of conditioned water is in communication with the reactor. A supply of carbon containing material is also in communication with the reactor. In one particular embodiment, the carbon containing material may include vacuum residuum (VR). The water and VR may be kept at desired temperatures and pressures compatible with the process that is to take place in the reactor. When introduced into the reactor, the water, the VR and the molten pool may be homogenously mixed in an environment in which chemical reactions take place including the production of hydrogen and other gasses.

System and process for the production of syngas and fuel gasses

DOEpatents

Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M; Benefiel, Bradley C

2015-04-21

The production of gasses and, more particularly, to systems and methods for the production of syngas and fuel gasses including the production of hydrogen are set forth. In one embodiment system and method includes a reactor having a molten pool of a material comprising sodium carbonate. A supply of conditioned water is in communication with the reactor. A supply of carbon containing material is also in communication with the reactor. In one particular embodiment, the carbon containing material may include vacuum residuum (VR). The water and VR may be kept at desired temperatures and pressures compatible with the process that is to take place in the reactor. When introduced into the reactor, the water, the VR and the molten pool may be homogenously mixed in an environment in which chemical reactions take place including the production of hydrogen and other gasses.

SPERT Destructive Test - I on Aluminum, Highly Enriched Plate Type Core

ScienceCinema

None

2018-01-16

SPERT - Special Power Excursion Reactor Tests Destructive Test number 1 On Aluminum, Highly Enriched Plate Type Core. A test studying the behavior of the reactor under destructive conditions on a light water moderated pool-type reactor with a plate-type core.

Dismantling the nuclear research reactor Thetis

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

Michiels, P.

The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3 m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storagemore » rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were used, which must be removed in controlled conditions. The ventilation system is considered free from nuclear contamination but it contains asbestos. This paper covers the organization of the dismantling work, the technical execution aspect and conclusions already known (dismantling is ongoing as this is written). (authors)« less

Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

NASA Astrophysics Data System (ADS)

Susyadi; Endiah, P. H.; Sukmanto, D.; Andi, S. E.; Syaiful, B.; Hendro, T.; Geni, R. S.

2018-02-01

RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

BWR zero pressure containment

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

Dillmann, C.W.; Townsend, H.E.; Nesbitt, L.B.

1992-02-25

This patent describes the operation of a nuclear reactor system, the system including a containment defining a drywall space wherein a nuclear reactor is disposed, there being a suppression pool in the containment with the suppression pool having a wetwell space above a level of the pool to which an non-condensable gases entering the suppression pool can vent. It comprises: continuously exhausting the wetwell space to remove gas mixture therefrom while admitting inflow of air from an atmospheric source thereof to the wetwell during normal operation by blocking off the inflow during a loss-of-coolant-accident whenever a pressure in the wetwellmore » space is above a predetermined value, and subjecting the gas subsequent to its removal from the wetwell to a treatment operation to separate any particulate material entrained therein from the gas mixture.« less

LPT. Shield test facility test building interior (TAN646). Camera facing ...

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

LPT. Shield test facility test building interior (TAN-646). Camera facing south. Distant pool contained EBOR reactor; near pool was intended for fuel rod storage. Other post-1970 activity equipment remains in pool. INEEL negative no. HD-40-9-4 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Water inventory management in condenser pool of boiling water reactor

DOEpatents

Gluntz, Douglas M.

1996-01-01

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

Water inventory management in condenser pool of boiling water reactor

DOEpatents

Gluntz, D.M.

1996-03-12

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

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

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

Interfacial heat transfer in multiphase molten pools with gas injection

NASA Astrophysics Data System (ADS)

Bilbao Y Leon, Rosa Marina

1998-12-01

In the very unlikely event of a severe reactor accident involving core meltdown and pressure vessel failure, it is vital to identify the circumstances that would allow the molten core material to cool down and resolidify, bringing core debris to a safe and stable state. In this type of accident, the molten material which escapes from the reactor pressure vessel will accumulate as a molten pool in the reactor cavity below. To achieve coolability of the corium in this configuration it has been proposed to flood the cavity with water from above forming a layered structure where upward heat loss from the molten pool to the water will cause the core material to quench and solidify. The effectiveness of this procedure depends largely on the rate of upward heat loss as well as on the formation and stability of an upper crust. In this situation the molten pool becomes a three phase mixture: the solid and liquid slurry formed by the molten pool cooled to a temperature below the temperature of liquidus, agitated by the gases formed in the concrete ablation process. The present work quantifies the partition of the heat losses upward and downward considering the influence of the solid fraction in the pool and the viscosity effects, and the rate of heat loss through a solid layer. To complete this task a intermediate scale experimental test section has been designed and built at the University of Wisconsin - Madison, in which simulant materials are used to model the process of heat and mass transfer which involves the molten pool, the solid layer atop and the coolant layer above. The design includes volumetric heating, gas injection from the bottom and solids within the pool. New experimental results showing the heat transfer behavior for pools with different viscosities and various solid fractions are presented. The current results indicate a power split which favors heat transfer upward to the coolant simulant above by a 2:1 or 3:1 ratio. In addition, the power split is unaffected by the viscosity of the pool, the solid fractions in the pool and the superficial velocity.

The detector system of the Daya Bay reactor neutrino experiment

DOE PAGES

An, F. P.

2015-12-15

The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin 22θ 13 and the effective mass splitting Δm 2 ee. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrummore » due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors’ baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This study describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.« less

The Underwater Spectrometric System Based on CZT Detector for Survey of the Bottom of MR Reactor Pool - 13461

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

Potapov, Victor; Safronov, Alexey; Ivanov, Oleg

2013-07-01

The underwater spectrometer system for detection of irradiated nuclear fuel on the pool bottom of the reactor was elaborated. During the development process metrological studies of CdZnTe (CZT) detectors were conducted. These detectors are designed for spectrometric measurements in high radiation fields. A mathematical model based on the Monte Carlo method was created to evaluate the capability of such a system. A few experimental models were realized and the characteristics of the spectrometric system are represented. (authors)

Fuel subassembly leak test chamber for a nuclear reactor

DOEpatents

Divona, Charles J.

1978-04-04

A container with a valve at one end is inserted into a nuclear reactor coolant pool. Once in the pool, the valve is opened by a mechanical linkage. An individual fuel subassembly is lifted into the container by a gripper; the valve is then closed providing an isolated chamber for the subassembly. A vacuum is drawn on the chamber to encourage gaseous fission product leakage through any defects in the cladding of the fuel rods comprising the subassembly; this leakage may be detected by instrumentation, and the need for replacement of the assembly ascertained.

Design of conduction cooling system for a high current HTS DC reactor

NASA Astrophysics Data System (ADS)

Dao, Van Quan; Kim, Taekue; Le Tat, Thang; Sung, Haejin; Choi, Jongho; Kim, Kwangmin; Hwang, Chul-Sang; Park, Minwon; Yu, In-Keun

2017-07-01

A DC reactor using a high temperature superconducting (HTS) magnet reduces the reactor’s size, weight, flux leakage, and electrical losses. An HTS magnet needs cryogenic cooling to achieve and maintain its superconducting state. There are two methods for doing this: one is pool boiling and the other is conduction cooling. The conduction cooling method is more effective than the pool boiling method in terms of smaller size and lighter weight. This paper discusses a design of conduction cooling system for a high current, high temperature superconducting DC reactor. Dimensions of the conduction cooling system parts including HTS magnets, bobbin structures, current leads, support bars, and thermal exchangers were calculated and drawn using a 3D CAD program. A finite element method model was built for determining the optimal design parameters and analyzing the thermo-mechanical characteristics. The operating current and inductance of the reactor magnet were 1,500 A, 400 mH, respectively. The thermal load of the HTS DC reactor was analyzed for determining the cooling capacity of the cryo-cooler. The study results can be effectively utilized for the design and fabrication of a commercial HTS DC reactor.

Best-estimate coupled RELAP/CONTAIN analysis of inadvertent BWR ADS valve opening transient

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

Feltus, M.A.; Muftuoglu, A.K.

1993-01-01

Noncondensible gases may become dissolved in boiling water reactor (BWR) water-level instrumentation during normal operations. Any dissolved noncondensible gases inside these water columns may come out of solution during rapid depressurization events and displace water from the reference leg piping, resulting in a false high level. Significant errors in water-level indication are not expected to occur until the reactor pressure vessel (RPV) pressure has dropped below [approximately]450 psig. These water level errors may cause a delay or failure in emergency core cooling system (ECCS) actuation. The RPV water level is monitored using the pressure of a water column having amore » varying height (reactor water level) that is compared to the pressure of a water column maintained at a constant height (reference level). The reference legs have small-diameter pipes with varying lengths that provide a constant head of water and are located outside the drywell. The amount of noncondensible gases dissolved in each reference leg is very dependent on the amount of leakage from the reference leg and its geometry and interaction of the reactor coolant system with the containment, i.e., torus or suppression pool, and reactor building. If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response. In the postulated inadvertent opening of all seven automatic depressurization system (ADS) valves, the ECCS signal on high drywell pressure would be circumvented because the ADS valves discharge directly into the suppression pool. A best-estimate analysis of such an inadvertent opening of all ADS valves would have to consider the thermal-hydraulic coupling between the pool, drywell, reactor building, and RPV.« less

The World at Your Feet: Immersive Interactive Displays Might Have a Bright Future in Education

ERIC Educational Resources Information Center

Simkins, Michael

2006-01-01

A reactor is an example of an immersive interactive play in which animated images are projected onto the floor. A reactor allows people to walk on images and interact with them using their feet. With reactors, people can stomp on kernels of popcorn, shoot a pool using their big toes, or wade through a shallow surf on pristine beaches. This…

Annual report, FY 1979 Spent fuel and fuel pool component integrity.

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

Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.

International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-..mu..m) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion.more » A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report.« less

Interior of the Plum Brook Reactor Facility

NASA Image and Video Library

1961-02-21

A view inside the 55-foot high containment vessel of the National Aeronautics and Space Administration (NASA) Plum Brook Reactor Facility in Sandusky, Ohio. The 60-megawatt test reactor went critical for the first time in 1961 and began its full-power research operations in 1963. From 1961 to 1973, this reactor performed some of the nation’s most advanced nuclear research. The reactor was designed to determine the behavior of metals and other materials after long durations of irradiation. The materials would be used to construct a nuclear-powered rocket. The reactor core, where the chain reaction occurred, sat at the bottom of the tubular pressure vessel, seen here at the center of the shielding pool. The core contained fuel rods with uranium isotopes. A cooling system was needed to reduce the heat levels during the reaction. A neutron-impervious reflector was also employed to send many of the neutrons back to the core. The Plum Brook Reactor Facility was constructed from high-density concrete and steel to prevent the excess neutrons from escaping the facility, but the water in the pool shielded most of the radiation. The water, found in three of the four quadrants served as a reflector, moderator, and coolant. In this photograph, the three 20-ton protective shrapnel shields and hatch have been removed from the top of the pressure tank revealing the reactor tank. An overhead crane could be manipulated to reach any section of this room. It was used to remove the shrapnel shields and transfer equipment.

Apparatus for draining lower drywell pool water into suppresion pool in boiling water reactor

DOEpatents

Gluntz, Douglas M.

1996-01-01

An apparatus which mitigates temperature stratification in the suppression pool water caused by hot water drained into the suppression pool from the lower drywell pool. The outlet of a spillover hole formed in the inner bounding wall of the suppression pool is connected to and in flow communication with one end of piping. The inlet end of the piping is above the water level in the suppression pool. The piping is routed down the vertical downcomer duct and through a hole formed in the thin wall separating the downcomer duct from the suppression pool water. The piping discharge end preferably has an elevation at or near the bottom of the suppression pool and has a location in the horizontal plane which is removed from the point where the piping first emerges on the suppression pool side of the inner bounding wall of the suppression pool. This enables water at the surface of the lower drywell pool to flow into and be discharged at the bottom of the suppression pool.

77 FR 7610 - Notice of Availability of Environmental Assessment and Finding of No Significant Impact for...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-13

... and reinforced concrete floors acting as diaphragms in distributing loads to vertically resisting... reinforced concrete foundation. The reactor is fueled with standard low-enriched TRIGA (Training, Research... cooled by a light water primary system consisting of the reactor pool and a heat removal system to remove...

Dry transfer system for spent fuel: Project report, A system designed to achieve the dry transfer of bare spent fuel between two casks. Final report

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

Dawson, D.M.; Guerra, G.; Neider, T.

1995-12-01

This report describes the system developed by EPRI/DOE for the dry transfer of spent fuel assemblies outside the reactor spent fuel pool. The system is designed to allow spent fuel assemblies to be removed from a spent fuel pool in a small cask, transported to the transfer facility, and transferred to a larger cask, either for off-site transportation or on-site storage. With design modifications, this design is capable of transferring single spent fuel assemblies from dry storage casks to transportation casks or visa versa. One incentive for the development of this design is that utilities with limited lifting capacity ormore » other physical or regulatory constraints are limited in their ability to utilize the current, more efficient transportation and storage cask designs. In addition, DOE, in planning to develop and implement the multi-purpose canister (MPC) system for the Civilian Radioactive Waste Management System, included the concept of an on-site dry transfer system to support the implementation of the MPC system at reactors with limitations that preclude the handling of the MPC system transfer casks. This Dry Transfer System can also be used at reactors wi decommissioned spent fuel pools and fuel in dry storage in non-MPC systems to transfer fuel into transportation casks. It can also be used at off-reactor site interim storage facilities for the same purpose.« less

Natural Convection Heat Transfer in a Rectangular Liquid Metal Pool With Bottom Heating and Top Cooling

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

Lee, Il S.; Yu, Yong H.; Son, Hyoung M.

2006-07-01

An experimental study is performed to investigate the natural convection heat transfer characteristics with subcooled coolant to create engineering database for basic applications in a lead alloy cooled reactor. Tests are performed in the ALTOS (Applied Liquid-metal Thermal Operation Study) apparatus as part of MITHOS (Metal Integrated Thermo Hydrodynamic Operation System). A relationship is determined between the Nusselt number Nu and the Rayleigh number Ra in the liquid metal rectangular pool. Results are compared with correlations and experimental data in the literature. Given the similar Ra condition, the present test results for Nu of the liquid metal pool with topmore » subcooling are found to be similar to those predicted by the existing correlations or experiments. The current test results are utilized to develop natural convection heat transfer correlations applicable to low Prandtl number Pr fluids that are heated from below and cooled by the external coolant above. Results from this study are slated to be used in designing BORIS (Battery Optimized Reactor Integral System), a small lead cooled modular fast reactor for deployment at remote sites cycled with MOBIS (Modular Optimized Brayton Integral System) for electricity generation, tied with NAVIS (Naval Application Vessel Integral System) for ship propulsion, joined with THAIS (Thermochemical Hydrogen Acquisition Integral System) for hydrogen production, and coupled with DORIS (Desalination Optimized Reactor Integral System) for seawater desalination. Tests are performed with Wood's metal (Pb-Bi-Sn-Cd) filling a rectangular pool whose lower surface is heated and upper surface cooled by forced convection of water. The test section is 20 cm long, 11.3 cm high and 15 cm wide. The simulant has a melting temperature of 78 deg. C. The constant temperature and heat flux condition was realized for the bottom heating once the steady state had been met. The test parameters include the heated bottom surface temperature of the liquid metal pool, the input power to the bottom surface of the section, and the coolant temperature. (authors)« less

Analysis of dose rates received around the storage pool for irradiated control rods in a BWR nuclear power plant.

PubMed

Ródenas, J; Abarca, A; Gallardo, S

2011-08-01

BWR control rods are activated by neutron reactions in the reactor. The dose produced by this activity can affect workers in the area surrounding the storage pool, where activated rods are stored. Monte Carlo (MC) models for neutron activation and dose assessment around the storage pool have been developed and validated. In this work, the MC models are applied to verify the expected reduction of dose when the irradiated control rod is hanged in an inverted position into the pool. 2010 Elsevier Ltd. All rights reserved.

Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

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

Licht, J.; Dionne, B.; Sikik, E.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showingmore » agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm 2 and temporary heat flux limit of 600 W/cm 2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.« less

Using SAFRAN Software to Assess Radiological Hazards from Dismantling of Tammuz-2 Reactor Core at Al-tuwaitha Nuclear Site

NASA Astrophysics Data System (ADS)

Abed Gatea, Mezher; Ahmed, Anwar A.; jundee kadhum, Saad; Ali, Hasan Mohammed; Hussein Muheisn, Abbas

2018-05-01

The Safety Assessment Framework (SAFRAN) software has implemented here for radiological safety analysis; to verify that the dose acceptance criteria and safety goals are met with a high degree of confidence for dismantling of Tammuz-2 reactor core at Al-tuwaitha nuclear site. The activities characterizing, dismantling and packaging were practiced to manage the generated radioactive waste. Dose to the worker was considered an endpoint-scenario while dose to the public has neglected due to that Tammuz-2 facility is located in a restricted zone and 30m berm surrounded Al-tuwaitha site. Safety assessment for dismantling worker endpoint-scenario based on maximum external dose at component position level in the reactor pool and internal dose via airborne activity while, for characterizing and packaging worker endpoints scenarios have been done via external dose only because no evidence for airborne radioactivity hazards outside the reactor pool. The in-situ measurements approved that reactor core components are radiologically activated by Co-60 radioisotope. SAFRAN results showed that the maximum received dose for workers are (1.85, 0.64 and 1.3mSv/y) for activities dismantling, characterizing and packaging of reactor core components respectively. Hence, the radiological hazards remain below the low level hazard and within the acceptable annual dose for workers in radiation field

Analysis of standard reference materials by absolute INAA

NASA Astrophysics Data System (ADS)

Heft, R. E.; Koszykowski, R. F.

1981-07-01

Three standard reference materials: flyash, soil, and ASI 4340 steel, are analyzed by a method of absolute instrumental neutron activation analysis. Two different light water pool-type reactors were used to produce equivalent analytical results even though the epithermal to thermal flux ratio in one reactor was higher than that in the other by a factor of two.

Default operational intervention levels (OILs) for severe nuclear power plant or spent fuel pool emergencies.

PubMed

McKenna, T; Kutkov, V; Vilar Welter, P; Dodd, B; Buglova, E

2013-05-01

Experience and studies show that for an emergency at a nuclear power plant involving severe core damage or damage to the fuel in spent fuel pools, the following actions may need to be taken in order to prevent severe deterministic health effects and reduce stochastic health effects: (1) precautionary protective actions and other response actions for those near the facility (i.e., within the zones identified by the International Atomic Energy Agency) taken immediately upon detection of facility conditions indicating possible severe damage to the fuel in the core or in the spent fuel pool; and (2) protective actions and other response actions taken based on environmental monitoring and sampling results following a release. This paper addresses the second item by providing default operational intervention levels [OILs, which are similar to the U.S. derived response levels (DRLs)] for promptly assessing radioactive material deposition, as well as skin, food, milk and drinking water contamination, following a major release of fission products from the core or spent fuel pool of a light water reactor (LWR) or a high power channel reactor (RBMK), based on the International Atomic Energy Agency's guidance.

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

Sarta, Jose A.; Castiblanco, Luis A

With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energias Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer ofmore » the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.« less

PBF (PER620) interior of Reactor Room. Camera facing south from ...

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

PBF (PER-620) interior of Reactor Room. Camera facing south from stairway platform in southwest corner (similar to platform in view at left). Reactor was beneath water in circular tank. Fuel was stored in the canal north of it. Platform and apparatus at right is reactor bridge with control rod mechanisms and actuators. The entire apparatus swung over the reactor and pool during operations. Personnel in view are involved with decontamination and preparation of facility for demolition. Note rails near ceiling for crane; motor for rollup door at upper center of view. Date: March 2004. INEEL negative no. HD-41-3-2 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

JEN-1 Reactor Control System; SISTEMA DE CONTROL DEL REACTOR JEN-1

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

Cantillo, M.F.; Nuno, C.M.; Andreu, J.L.M.

1963-01-01

ABS>The JEN-1 3Mw power swimming pool reactor electrical control circuits are described. Start-up, power generation in the core, and shutdown are controlled by the reactor control system. This control system guarantees in each moment the safety conditions during reactor operation. Each circuit was represented by a scheme, complemented with a description of its function, components, and operation theory. Components described include: scram circuit; fission counter control circuit; servo control circuit; control circuit of safety sheets; control circuits of primary, secondary, and clean-up pump motors and tower fan motor; primary valve motor circuit; center cubicle alarm circuit; and process alarm circuit.more » (auth)« less

Scoping studies of vapor behavior during a severe accident in a metal-fueling reactor

NASA Astrophysics Data System (ADS)

Spencer, B. W.; Marchaterre, J. F.

1985-04-01

The consequences of fuel melting and pin failures for a reactivity-insertion type accident in a sodium-cooled, pool-type reactor fueled with a metal alloy fuel were examined. The principal gas and vapor species released are shown to be Xe, Cs, and bond sodium contained within the fuel porosity. Condensation of sodium vapor as it expands into the upper sodium pool in a jet mixing regime may occur as rapidly as the vapor emerges from the disrupted core. If the predictions of rapid direct-contact condensation can be verified experimentally for the sodium system, the ability of vapor expansion to perform appreciable work on the system and the ability of an expanding vapor bubble to transport fuel and fission produce species to the cover gas region where they may be released to the containment are largely eliminated. The radionuclide species except for fission gas are largely retained within the core and sodium pool.

Loss of DHR sequences at Browns Ferry Unit One - accident-sequence analysis

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

Cook, D.H.; Grene, S.R.; Harrington, R.M.

1983-05-01

This study describes the predicted response of Unit One at the Browns Ferry Nuclear Plant to a postulated loss of decay heat removal (DHR) capability following scram from full power with the power conversion system unavailable. In accident sequences without DHR capability, the residual heat removal (RHR) system functions of pressure suppression pool cooling and reactor vessel shutdown cooling are unavailable. Consequently, all decay heat energy is stored in the pressure suppression pool with a concomitant increase in pool temperature and primary containment pressure. With the assumption that DHR capability is not regained during the lengthy course of this accidentmore » sequence, the containment ultimately fails by overpressurization. Although unlikely, this catastrophic failure might lead to loss of the ability to inject cooling water into the reactor vessel, causing subsequent core uncovery and meltdown. The timing of these events and the effective mitigating actions that might be taken by the operator are discussed in this report.« less

Detectability prediction for a thermoacoustic sensor in the breazeale nuclear reactor pool

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

Smith, James; Hrisko, Joshua; Garrett, Steven

2016-03-01

Laboratory experiments have suggested that thermoacoustic engines can be in- corporated within nuclear fuel rods. Such engines would radiate sounds that could be used to measure and acoustically-telemeter information about the op- eration of the nuclear reactor (e.g., coolant temperature or uxes of neutrons or other energetic particles) or the physical condition of the nuclear fuel itself (e.g., changes in temperature, evolved gases) that are encoded as the frequency and/or amplitude of the radiated sound [IEEE Measurement and Instrumen- tation 16(3), 18-25 (2013)]. For such acoustic information to be detectable, it is important to characterize the vibroacoustical environments within reactors.more » Measurements will be presented of the background noise spectra (with and with- out coolant pumps) and reverberation times within the 70,000 gallon pool that cools and shields the fuel in the 1 MW research reactor on Penn State's campus using two hydrophones, a piezoelectric projector, and an accelerometer. Sev- eral signal-processing techniques will be demonstrated to enhance the measured results. Background vibrational measurement were also taken at the 250 MW Advanced Test Reactor, located at the Idaho National Laboratory, using ac- celerometers mounted outside the reactor's pressure vessel and on plumbing will also be presented. The detectability predictions made in the thesis were validated in September 2015 using a nuclear ssion-heated thermoacoustic sensor that was placed in the core of the Breazeale Nuclear Reactor on Penn State's campus. Some features of the thermoacoustic device used in that experiment will also be revealed. [Work supported by the U.S. Department of Energy.]« less

Pressure suppression containment system

DOEpatents

Gluntz, Douglas M.; Townsend, Harold E.

1994-03-15

A pressure suppression containment system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The wetwell pool includes a plenum for receiving the non-condensable gas carried with steam from the drywell following a loss-of coolant-accident (LOCA). The wetwell plenum is vented to a plenum above the GDCS pool following the LOCA for suppressing pressure rise within the containment vessel. A method of operation includes channeling steam released into the drywell following the LOCA into the wetwell pool for cooling along with the non-condensable gas carried therewith. The GDCS pool is then drained by gravity, and the wetwell plenum is vented into the GDCS plenum for channeling the non-condensable gas thereto.

Pressure suppression containment system

DOEpatents

Gluntz, D.M.; Townsend, H.E.

1994-03-15

A pressure suppression containment system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The wetwell pool includes a plenum for receiving the non-condensable gas carried with steam from the drywell following a loss-of-coolant-accident (LOCA). The wetwell plenum is vented to a plenum above the GDCS pool following the LOCA for suppressing pressure rise within the containment vessel. A method of operation includes channeling steam released into the drywell following the LOCA into the wetwell pool for cooling along with the non-condensable gas carried therewith. The GDCS pool is then drained by gravity, and the wetwell plenum is vented into the GDCS plenum for channeling the non-condensable gas thereto. 6 figures.

Application of the Monte Carlo method to estimate doses due to neutron activation of different materials in a nuclear reactor

NASA Astrophysics Data System (ADS)

Ródenas, José

2017-11-01

All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc.

Intelligent uranium fission converter for neutron production on the periphery of the nuclear reactor core (MARIA reactor in Swierk - Poland)

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

Gryzinski, M.A.; Wielgosz, M.

The multipurpose, high flux research reactor MARIA in Otwock - Swierk is an open-pool type, water and beryllium moderated and graphite reflected. There are two not occupied experimental H1 and H2 horizontal channels with complex of empty rooms beside them. Making use of these two channels is not in conflict with other research or commercial employing channels. They can work simultaneously, moreover commercial channels covers the cost of reactor working. Such conditions give beneficial possibility of creating epithermal neutron stand for researches in various field at the horizontal channel H2 of MARIA reactor (co-organization of research at H1 channel ismore » additionally planned). At the front of experimental channels the neutron flux is strongly thermalized - neutrons with energies above 0.625 eV constitute only ∼2% of the total flux. This thermalized neutron flux will be used to achieve high flux of epithermal neutrons at the level of 2x10{sup 9} n cm{sup -2}s{sup -1} by uranium neutron converter (fast neutron production - conversion of reactor core thermal neutrons to fast neutrons - and then filtering, moderating and finally cutting of unwanted gamma radiation). The intelligent converter will be placed in the reactor pool, near the front of the H2 channel. It will replace one graphite block at the periphery of MARIA graphite reflector. The converter will consist of 20 fuel elements - low enriched uranium plates. A fuel plate will be a part which will measure 110 mm wide by 380 mm long and will consist of a thin layer of uranium sealed between two aluminium plates. These plates, once assembled, form the fuel element used in converter. The plates will be positioned vertically. There are several important requirements which should be taken into account at the converter design stage: -maximum efficiency of the converter for neutrons conversion, -cooling of the converter need to be integrated with the cooling circuit of the reactor pool and if needed equipped with self-cooling system (enhanced comparing to the cooling properties inherent with regular rector pool water flows), -proper cooling conditions can be ensured by an appropriate water flow, so the resistance to flow has to be optimised, -the requirement of the minimum resistance to water flow leads to the openwork design of the fuel element separator, which, on the other hand, has to be strong enough to ensure the needed strength for mechanical load due to the fuel weight and forces associated with the water flow, -the possibility of changing beam and flux qualities by rotating the converter or repositioning the converter plates by moving or replacing with another materials. In order to minimize the neutron activation of the fuel in the converter, the possibility was predicted to remove the converter and to replace it with an aluminium dummy for the time when the beam at the channel H2 is not used. This means that both, the converter and the dummy, have to be easily removable from the converter socket. There has to be also the place in the water pool, near the research stand or in technological pool, where the converter can be safely stored (this place have to be proper for operation with plates i.e. changing amount of plates). Thermal and neutron load of the fuel plates in the converter will be inhomogeneous. In order to equalize these loads, the converter should be designed in such way that it would be possible to change the order of fuel plates. Moreover replacing the amount of the plates gives the opportunity to obtain different fluxes of neutrons (quantitatively and qualitatively i.e. energetically). The project of the converter is based on Monte Carlo calculation concerning neutron production and on Computational Fluid Dynamics (CFD) i.e. modelling of converter for thermodynamical aspects. (authors)« less

Utilization of the Philippine Research Reactor as a training facility for nuclear power plant operators

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

Palabrica, R.J.

1981-01-01

The Philippines has a 1-MW swimming-pool reactor facility operated by the Philippine Atomic Energy Commission (PAEC). The reactor is light-water moderated and cooled, graphite reflected, and fueled with 90% enriched uranium. Since it became critical in 1963 it has been utilized for research, radioisotope production, and training. It was used initially in the training of PAEC personnel and other research institutions and universities. During the last few years, however, it has played a key role in training personnel for the Philippine Nuclear Power Project (PNPP).

ETR CRITICAL FACILITY, TRA654. CONTEXTUAL VIEW. CAMERA ON ROOF OF ...

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

ETR CRITICAL FACILITY, TRA-654. CONTEXTUAL VIEW. CAMERA ON ROOF OF MTR BUILDING AND FACING SOUTH. ETR AND ITS COOLANT BUILDING AT UPPER PART OF VIEW. ETR COOLING TOWER NEAR TOP EDGE OF VIEW. EXCAVATION AT CENTER IS FOR ETR CF. CENTER OF WHICH WILL CONTAIN POOL FOR REACTOR. NOTE CHOPPER TUBE PROCEEDING FROM MTR IN LOWER LEFT OF VIEW, DIAGONAL TOWARD LEFT. INL NEGATIVE NO. 56-4227. Jack L. Anderson, Photographer, 12/18/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Proposed Design and Operation of a Heat Pipe Reactor using the Sandia National Laboratories Annular Core Test Facility and Existing UZrH Fuel Pins

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

Wright, Steven A.; Lipinski, Ronald J.; Pandya, Tara

2005-02-06

Heat Pipe Reactors (HPR) for space power conversion systems offer a number of advantages not easily provided by other systems. They require no pumping, their design easily deals with freezing and thawing of the liquid metal, and they can provide substantial levels of redundancy. Nevertheless, no reactor has ever been operated and cooled with heat pipes, and the startup and other operational characteristics of these systems remain largely unknown. Signification deviations from normal reactor heat removal mechanisms exist, because the heat pipes have fundamental heat removal limits due to sonic flow issues at low temperatures. This paper proposes an earlymore » prototypic test of a Heat Pipe Reactor (using existing 20% enriched nuclear fuel pins) to determine the operational characteristics of the HPR. The proposed design is similar in design to the HOMER and SAFE-300 HPR designs (Elliot, Lipinski, and Poston, 2003; Houts, et. al, 2003). However, this reactor uses existing UZrH fuel pins that are coupled to potassium heat pipes modules. The prototype reactor would be located in the Sandia Annular Core Research Reactor Facility where the fuel pins currently reside. The proposed reactor would use the heat pipes to transport the heat from the UZrH fuel pins to a water pool above the core, and the heat transport to the water pool would be controlled by adjusting the pressure and gas type within a small annulus around each heat pipe. The reactor would operate as a self-critical assembly at power levels up to 200 kWth. Because the nuclear heated HPR test uses existing fuel and because it would be performed in an existing facility with the appropriate safety authorization basis, the test could be performed rapidly and inexpensively. This approach makes it possible to validate the operation of a HPR and also measure the feedback mechanisms for a typical HPR design. A test of this nature would be the world's first operating Heat Pipe Reactor. This reactor is therefore called 'HPR-1'.« less

Metal fires and their implications for advanced reactors.

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

Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean

This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in thesemore » areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety analysis capabilities of the advanced-reactor community for directly relevant scenarios. Beyond the focus on the thermally-interacting and smoldering sodium pool fires, experimental and analysis capabilities for sodium spray fires have also been developed in this project.« less

Hydrogen or formate: Alternative key players in methanogenic degradation.

PubMed

Schink, Bernhard; Montag, Dominik; Keller, Anja; Müller, Nicolai

2017-06-01

Hydrogen and formate are important electron carriers in methanogenic degradation in anoxic environments such as sediments, sewage sludge digestors and biogas reactors. Especially in the terminal steps of methanogenesis, they determine the energy budgets of secondary (syntrophically) fermenting bacteria and their methanogenic partners. The literature provides considerable data on hydrogen pool sizes in such habitats, but little data exist for formate concentrations due to technical difficulties in formate determination at low concentration. Recent evidence from biochemical and molecular biological studies indicates that several secondary fermenters can use both hydrogen and formate for electron release, and may do so even simultaneously. Numerous strictly anaerobic bacteria contain enzymes which equilibrate hydrogen and formate pools to energetically equal values, and recent measurements in sewage digestors and biogas reactors indicate that - beyond occasional fluctuations - the pool sizes of hydrogen and formate are indeed energetically nearly equivalent. Nonetheless, a thermophilic archaeon from a submarine hydrothermal vent, Thermococcus onnurineus, can obtain ATP from the conversion of formate to hydrogen plus bicarbonate at 80°C, indicating that at least in this extreme environment the pools of formate and hydrogen are likely to be sufficiently different to support such an unusual type of energy conservation. © 2017 Society for Applied Microbiology and John Wiley & Sons Ltd.

Nuclear reactor construction with bottom supported reactor vessel

DOEpatents

Sharbaugh, John E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

Pressure suppression system

DOEpatents

Gluntz, D.M.

1994-10-04

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein. 3 figs.

Pressure suppression system

DOEpatents

Gluntz, Douglas M.

1994-01-01

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein.

Astéroides et satellites de Saturne: quelques résultats récents et contribution éventuelle des données Hipparcos.

NASA Astrophysics Data System (ADS)

Viateau, B.; Rapaport, M.

48 astéroides et 2 satellites de Saturne étaient au programme de la mission Hipparcos, et diverses propositions ont été faites pour l'utilisation de ces données. Les auteurs présentent quelques résultats récents concernant ces objets, et susceptibles de 1) donner un supplément d'intére^t aux données astrométriques fournies par Hipparcos, 2) permettre de préciser les objectifs contenus dans diverses propositions.

Operating manual for the Bulk Shielding Reactor

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

Not Available

1983-04-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR.

Operating manual for the Bulk Shielding Reactor

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

Not Available

1987-03-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR.

Control console replacement at the WPI Reactor

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

Not Available

1992-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Radiant vessel auxiliary cooling system

DOEpatents

Germer, John H.

1987-01-01

In a modular liquid-metal pool breeder reactor, a radiant vessel auxiliary cooling system is disclosed for removing the residual heat resulting from the shutdown of a reactor by a completely passive heat transfer system. A shell surrounds the reactor and containment vessel, separated from the containment vessel by an air passage. Natural circulation of air is provided by air vents at the lower and upper ends of the shell. Longitudinal, radial and inwardly extending fins extend from the shell into the air passage. The fins are heated by radiation from the containment vessel and convect the heat to the circulating air. Residual heat from the primary reactor vessel is transmitted from the reactor vessel through an inert gas plenum to a guard or containment vessel designed to contain any leaking coolant. The containment vessel is conventional and is surrounded by the shell.

Passive containment cooling system

DOEpatents

Billig, P.F.; Cooke, F.E.; Fitch, J.R.

1994-01-25

A passive containment cooling system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel and is vented to the drywell. An isolation pool is disposed above the GDCS pool and includes an isolation condenser therein. The condenser has an inlet line disposed in flow communication with the drywell for receiving the non-condensable gas along with any steam released therein following a loss-of-coolant accident (LOCA). The condenser also has an outlet line disposed in flow communication with the drywell for returning to the drywell both liquid condensate produced upon cooling of the steam and the non-condensable gas for reducing pressure within the containment vessel following the LOCA. 1 figure.

Passive containment cooling system

DOEpatents

Billig, Paul F.; Cooke, Franklin E.; Fitch, James R.

1994-01-01

A passive containment cooling system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel and is vented to the drywell. An isolation pool is disposed above the GDCS pool and includes an isolation condenser therein. The condenser has an inlet line disposed in flow communication with the drywell for receiving the non-condensable gas along with any steam released therein following a loss-of-coolant accident (LOCA). The condenser also has an outlet line disposed in flow communication with the drywell for returning to the drywell both liquid condensate produced upon cooling of the steam and the non-condensable gas for reducing pressure within the containment vessel following the LOCA.

Israel’s Attack on Osiraq: A Model for Future Preventive Strikes?

DTIC Science & Technology

2005-07-01

Chauvet . Shamir told Chauvet , “Israel holds France exclusively responsible for the results liable to arise from operation of the reactor and misuse...of the nuclear fuel.” Chauvet argued, “Acquisition of nuclear arms would be lunacy on the part of Iraq. After all, Israel’s Jewish and Arab...McKinnon, the tapes from aircraft number seven and eight reveal the reactor dome completely caved in and a destroyed cooling pool.57 However

Dismantling of Loop-Type Channel Equipment of MR Reactor in NRC 'Kurchatov Institute' - 13040

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

Volkov, Victor; Danilovich, Alexey; Zverkov, Yuri

2013-07-01

In 2009 the project of decommissioning of MR and RTF reactors was developed and approved by the Expert Authority of the Russian Federation (Gosexpertiza). The main objective of the decommissioning works identified in this project: - complete dismantling of reactor equipment and systems; - decontamination of reactor premises and site in accordance with the established sanitary and hygienic standards. At the preparatory stage (2008-2010) of the project the following works were executed: loop-type channels' dismantling in the storage pool; experimental fuel assemblies' removal from spent fuel repositories in the central hall; spent fuel assembly removal from the liquid-metal-cooled loop-type channelmore » of the reactor core and its placement into the SNF repository; and reconstruction of engineering support systems to the extent necessary for reactor decommissioning. The project assumes three main phases of dismantling and decontamination: - dismantling of equipment/pipelines of cooling circuits and loop-type channels, and auxiliary reactor equipment (2011-2012); - dismantling of equipment in underground reactor premises and of both MR and RTF in-vessel devices (2013-2014); - decontamination of reactor premises; rehabilitation of the reactor site; final radiation survey of reactor premises, loop-type channels and site; and issuance of the regulatory authorities' de-registration statement (2015). In 2011 the decommissioning license for the two reactors was received and direct MR decommissioning activities started. MR primary pipelines and loop-type facilities situated in the underground reactor hall were dismantled. Works were also launched to dismantle the loop-type channels' equipment in underground reactor premises; reactor buildings were reconstructed to allow removal of dismantled equipment; and the MR/RTF decommissioning sequence was identified. In autumn 2011 - spring 2012 results of dismantling activities performed are: - equipment from underground rooms (No. 66, 66A, 66B, 72, 64, 63) - as well as from water and gas loop corridors - was dismantled, with the total radwaste weight of 53 tons and the total removed activity of 5,0 x 10{sup 10} Bq; - loop-type channel equipment from underground reactor hall premises was dismantled; - 93 loop-type channels were characterized, chopped and removed, with radwaste of 2.6 x 10{sup 13} Bq ({sup 60}Co) and 1.5 x 10{sup 13} Bq ({sup 137}Cs) total activity removed from the reactor pool, fragmented and packaged. Some of this waste was placed into the high-level waste (HLW) repository of the Center. Dismantling works were executed with application of remotely operated mechanisms, which promoted decrease of radiation impact on the personnel. The average individual dose for the personnel was 1.9 mSv/year in 2011, and the collective dose is estimated as 0.0605 man x Sv/year. (authors)« less

Declassification of radioactive water from a pool type reactor after nuclear facility dismantling

NASA Astrophysics Data System (ADS)

Arnal, J. M.; Sancho, M.; García-Fayos, B.; Verdú, G.; Serrano, C.; Ruiz-Martínez, J. T.

2017-09-01

This work is aimed to the treatment of the radioactive water from a dismantled nuclear facility with an experimental pool type reactor. The main objective of the treatment is to declassify the maximum volume of water and thus decrease the volume of radioactive liquid waste to be managed. In a preliminary stage, simulation of treatment by the combination of reverse osmosis (RO) and evaporation have been performed. Predicted results showed that the combination of membrane and evaporation technologies would result in a volume reduction factor higher than 600. The estimated time to complete the treatment was around 650 h (25-30 days). For different economical and organizational reasons which are explained in this paper, the final treatment of the real waste had to be reduced and only evaporation was applied. The volume reduction factor achieved in the real treatment was around 170, and the time spent for treatment was 194 days.

Spent fuel pool storage calculations using the ISOCRIT burnup credit tool

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

Kucukboyaci, Vefa; Marshall, William BJ J

2012-01-01

In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse's state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion,more » thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc.) with a quick turnaround.« less

Evaluation of Fe(II) oxidation at an acid mine drainage site using laboratory-scale reactors

NASA Astrophysics Data System (ADS)

Brown, Juliana; Burgos, William

2010-05-01

Acid mine drainage (AMD) is a severe environmental threat to the Appalachian region of the Eastern United States. The Susquehanna and Potomac River basins of Pennsylvania drain to the Chesapeake Bay, which is heavily polluted by acidity and metals from AMD. This study attempted to unravel the complex relationships between AMD geochemistry, microbial communities, hydrodynamic conditions, and the mineral precipitates for low-pH Fe mounds formed downstream of deep mine discharges, such as Lower Red Eyes in Somerset County, PA, USA. This site is contaminated with high concentrations of Fe (550 mg/L), Mn (115 mg/L), and other trace metals. At the site 95% of dissolved Fe(II) and 56% of total dissolved Fe is removed without treatment, across the mound, but there is no change in the concentration of trace metals. Fe(III) oxides were collected across the Red Eyes Fe mound and precipitates were analyzed by X-ray diffraction, electron microscopy and elemental analysis. Schwertmannite was the dominant mineral phase with traces of goethite. The precipitates also contained minor amounts of Al2O3, MgO,and P2O5. Laboratory flow-through reactors were constructed to quantify Fe(II) oxidation and Fe removal over time at terrace and pool depositional facies. Conditions such as residence time, number of reactors in sequence and water column height were varied to determine optimal conditions for Fe removal. Reactors with sediments collected from an upstream terrace oxidized more than 50% of dissolved Fe(II) at a ten hour residence time, while upstream pool sediments only oxidized 40% of dissolved Fe(II). Downstream terrace and pool sediments were only capable of oxidizing 25% and 20% of Fe(II), respectively. Fe(II) oxidation rates measured in the reactors were determined to be between 3.99 x 10-8and 1.94 x 10-7mol L-1s-1. The sediments were not as efficient for total dissolved Fe removal and only 25% was removed under optimal conditions. The removal efficiency for all sediments decreased as residence time decreased and as water column depth increased. Control reactors with Co-60 irradiated sediments showed an increase in Fe concentration as a result of dissolution of the sediments; thus, it was concluded that Fe(II) oxidation in the reactors was a result of biological processes and not abiotic oxidation. It was also concluded that Fe(II) oxidation and removal rates were dependent upon geochemical gradients (pH, Fe(II) concentration) rather than depositional facies. Fluorescent in situ hybridization was also performed on field and reactor samples to determine which microbial communities were responsible for the highest Fe(II) oxidation rates.

Modélisation par éléments finis 3D du champ magnétostatique dans les enroulements des réactances cuirassées de grande puissance. Comparaison avec le calcul en 2D

NASA Astrophysics Data System (ADS)

Ngnegueu, Triomphant; Terme, Claude; Mailhot, Michel

1993-03-01

In this paper, the finite element method is applied for the computation of the magnetostatic field in the windings of a shell-form reactor. The modeling is carried out in 3D, using FLUX3D, a software developed at the Laboratoire d'Electrotechnique de Grenoble. The results are compared to those obtained in 2D. These calculation results are also compared to some test results. Dans cet article, nous décrivons une application de la méthode des éléments finis pour la modélisation du champ magnétostatique dans les enroulements d'une réactance cuirassée de grande puissance. La modélisation est conduite en 3D, en utilisant le logiciel FLUX3D. Les résultats du calcul sont comparés avec ceux obtenus en 2D. Quelques comparaisons sont aussi effectuées avec des résultats de mesure.

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

Snepvangers, J.J.M.

Equipment and results are described connected with irradiation studies of UO/sub 2/ fuels, fuel element testing in pressurized water loops, graphite irradiation, and steel irradiations with and without temperature control. The apparatus described is associated with a 20-Mw pool-type research reactor. (T.F.H.)

Recent MELCOR and VICTORIA Fission Product Research at the NRC

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

Bixler, N.E.; Cole, R.K.; Gauntt, R.O.

1999-01-21

The MELCOR and VICTORIA severe accident analysis codes, which were developed at Sandia National Laboratories for the U. S. Nuclear Regulatory Commission, are designed to estimate fission product releases during nuclear reactor accidents in light water reactors. MELCOR is an integrated plant-assessment code that models the key phenomena in adequate detail for risk-assessment purposes. VICTORIA is a more specialized fission- product code that provides detailed modeling of chemical reactions and aerosol processes under the high-temperature conditions encountered in the reactor coolant system during a severe reactor accident. This paper focuses on recent enhancements and assessments of the two codes inmore » the area of fission product chemistry modeling. Recently, a model for iodine chemistry in aqueous pools in the containment building was incorporated into the MELCOR code. The model calculates dissolution of iodine into the pool and releases of organic and inorganic iodine vapors from the pool into the containment atmosphere. The main purpose of this model is to evaluate the effect of long-term revolatilization of dissolved iodine. Inputs to the model include dose rate in the pool, the amount of chloride-containing polymer, such as Hypalon, and the amount of buffering agents in the containment. Model predictions are compared against the Radioiodine Test Facility (RTF) experiments conduced by Atomic Energy of Canada Limited (AECL), specifically International Standard Problem 41. Improvements to VICTORIA's chemical reactions models were implemented as a result of recommendations from a peer review of VICTORIA that was completed last year. Specifically, an option is now included to model aerosols and deposited fission products as three condensed phases in addition to the original option of a single condensed phase. The three-condensed-phase model results in somewhat higher predicted fission product volatilities than does the single-condensed-phase model. Modeling of U02 thermochemistry was also improved, and results in better prediction of vaporization of uranium from fuel, which can react with released fission products to affect their volatility. This model also improves the prediction of fission product release rates from fuel. Finally, recent comparisons of MELCOR and VICTORIA with International Standard Problem 40 (STORM) data are presented. These comparisons focus on predicted therrnophoretic deposition, which is the dominant deposition mechanism. Sensitivity studies were performed with the codes to examine experimental and modeling uncertainties.« less

The WPI reactor-readying for the next generation

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

Bobek, L.M.

1993-01-01

Built in 1959, the 10-kW open-pool nuclear training reactor at Worcester Polytechnic Institute (WPI) was one of the first such facilities in the nation located on a university campus. Since then, the reactor and its related facilities have been used to train two generations of nuclear engineers and scientists for the nuclear industry. With the use of nuclear technology playing an increasing role in many segments of the economy, WPI with its nuclear reactor facility is committed to continuing its mission of training future nuclear engineers and scientists. The WPI reactor includes a 6-in. beam port, graphite thermal column, andmore » in-core sample facility. The reactor, housed in an open 8000-gal tank of water, is designed so that the core is readily accessible. Both the control console and the peripheral counting equipment used for student projects and laboratory exercises are located in the reactor room. This arrangement provides convenience and flexibility in using the reactor for foil activations in neutron flux measurements, diffusion measurements, radioactive decay measurements, and the neutron activation of samples for analysis. In 1988, the reactor was successfully converted to low-enriched uranium fuel.« less

Control console replacement at the WPI Reactor. [Final report

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

Not Available

1992-12-31

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

The U.S. Geological Survey's TRIGA® reactor

USGS Publications Warehouse

DeBey, Timothy M.; Roy, Brycen R.; Brady, Sally R.

2012-01-01

The U.S. Geological Survey (USGS) operates a low-enriched uranium-fueled, pool-type reactor located at the Federal Center in Denver, Colorado. The mission of the Geological Survey TRIGA® Reactor (GSTR) is to support USGS science by providing information on geologic, plant, and animal specimens to advance methods and techniques unique to nuclear reactors. The reactor facility is supported by programs across the USGS and is organizationally under the Associate Director for Energy and Minerals, and Environmental Health. The GSTR is the only facility in the United States capable of performing automated delayed neutron analyses for detecting fissile and fissionable isotopes. Samples from around the world are submitted to the USGS for analysis using the reactor facility. Qualitative and quantitative elemental analyses, spatial elemental analyses, and geochronology are performed. Few research reactor facilities in the United States are equipped to handle the large number of samples processed at the GSTR. Historically, more than 450,000 sample irradiations have been performed at the USGS facility. Providing impartial scientific information to resource managers, planners, and other interested parties throughout the world is an integral part of the research effort of the USGS.

Characterization of Sodium Thermal Hydraulics with Optical Fiber Temperature Sensors

NASA Astrophysics Data System (ADS)

Weathered, Matthew Thomas

The thermal hydraulic properties of liquid sodium make it an attractive coolant for use in Generation IV reactors. The liquid metal's high thermal conductivity and low Prandtl number increases efficiency in heat transfer at fuel rods and heat exchangers, but can also cause features such as high magnitude temperature oscillations and gradients in the coolant. Currently, there exists a knowledge gap in the mechanisms which may create these features and their effect on mechanical structures in a sodium fast reactor. Two of these mechanisms include thermal striping and thermal stratification. Thermal striping is the oscillating temperature field created by the turbulent mixing of non-isothermal flows. Usually this occurs at the reactor core outlet or in piping junctions and can cause thermal fatigue in mechanical structures. Meanwhile, thermal stratification results from large volumes of non-isothermal sodium in a pool type reactor, usually caused by a loss of coolant flow accident. This stratification creates buoyancy driven flow transients and high temperature gradients which can also lead to thermal fatigue in reactor structures. In order to study these phenomena in sodium, a novel method for the deployment of optical fiber temperature sensors was developed. This method promotes rapid thermal response time and high spatial temperature resolution in the fluid. The thermal striping and stratification behavior in sodium may be experimentally analyzed with these sensors with greater fidelity than ever before. Thermal striping behavior at a junction of non-isothermal sodium was fully characterized with optical fibers. An experimental vessel was hydrodynamically scaled to model thermal stratification in a prototypical sodium reactor pool. Novel auxiliary applications of the optical fiber temperature sensors were developed throughout the course of this work. One such application includes local convection coefficient determination in a vessel with the corollary application of level sensing. Other applications were cross correlation velocimetry to determine bulk sodium flow rate and the characterization of coherent vortical structures in sodium with temperature frequency data. The data harvested, instrumentation developed and techniques refined in this work will help in the design of more robust reactors as well as validate computational models for licensing sodium fast reactors.

Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

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

Not Available

1995-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomicmore » facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor`s Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced.« less

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

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

Licht, J. R.; Bergeron, A.; Dionne, B.

2015-12-01

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux ofmore » 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.« less

Radiation dose distributions due to sudden ejection of cobalt device.

PubMed

Abdelhady, Amr

2016-09-01

The evaluation of the radiation dose during accident in a nuclear reactor is of great concern from the viewpoint of safety. One of important accident must be analyzed and may be occurred in open pool type reactor is the rejection of cobalt device. The study is evaluating the dose rate levels resulting from upset withdrawal of co device especially the radiation dose received by the operator in the control room. Study of indirect radiation exposure to the environment due to skyshine effect is also taken into consideration in order to evaluate the radiation dose levels around the reactor during the ejection trip. Microshield, SHLDUTIL, and MCSky codes were used in this study to calculate the radiation dose profiles during cobalt device ejection trip inside and outside the reactor building. Copyright © 2016 Elsevier Ltd. All rights reserved.

Pool-site fuel inspection and examination techniques applied by the Kraftwerk Union Aktiengesellschaft Fuel Service. [PWR; BWR

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

Knaab, H.; Knecht, K.

The need for pool-site inspection and examination of fuel assemblies was recognized by Kraftwerk Union Aktiengesellschaft with the commissioning of the first nuclear power stations. A wet sipping method has demonstrated high reliability in detection of leaking fuel assemblies. The visual inspection system is a versatile tool. It can be supplemented by attaching devices for oxide thickness measurement or surface replication. Repair of leaking pressurized water reactor fuel assemblies has improved fuel utilization. Applied methods and typical results are described.

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

Kreitman, Paul J.; Sirianni, Steve R.; Pillard, Mark M.

Entergy recently performed an Extended Power Up-rate (EPU) on their Grand Gulf Nuclear Station, near Port Gibson, Mississippi. To support the EPU, a new Steam Dryer Assembly was installed during the last refueling outage. Due to limited access into the containment, the large Replacement Steam Dryer (RSD) had to be brought into the containment in pieces and then final assembly was completed on the refueling floor before installation into the reactor. Likewise, the highly contaminated Original Steam Dryer (OSD) had to be segmented into manageable sections, loaded into specially designed shielded containers, and rigged out of containment where they willmore » be safely stored until final disposal is accomplished at an acceptable waste repository. Westinghouse Nuclear Services was contracted by Entergy to segment, package and remove the OSD from containment. This work was performed on critical path during the most recent refueling outage. The segmentation was performed underwater to minimize radiation exposure to the workers. Special hydraulic saws were developed for the cutting operations based on Westinghouse designs previously used in Sweden to segment ABB Reactor Internals. The mechanical cutting method was selected because of its proven reliability and the minimal cutting debris that is generated by the process. Maintaining stability of the large OSD sections during cutting was accomplished using a custom built support stand that was installed into the Moisture Separator Pool after the Moisture Separator was installed back in the reactor vessel. The OSD was then moved from the Steam Dryer Pool to the Moisture Separator Pool for segmentation. This scenario resolved the logistical challenge of having two steam dryers and a moisture separator in containment simultaneously. A water filtration/vacuum unit was supplied to maintain water clarity during the cutting and handling operations and to collect the cutting chips. (authors)« less

Release of plutonium isotopes into the environment from the Fukushima Daiichi Nuclear Power Plant accident: what is known and what needs to be known.

PubMed

Zheng, Jian; Tagami, Keiko; Uchida, Shigeo

2013-09-03

The Fukushima Daiichi Nuclear Power Plant (FDNPP) accident has caused serious contamination in the environment. The release of Pu isotopes renewed considerable public concern because they present a large risk for internal radiation exposure. In this Critical Review, we summarize and analyze published studies related to the release of Pu from the FDNPP accident based on environmental sample analyses and the ORIGEN model simulations. Our analysis emphasizes the environmental distribution of released Pu isotopes, information on Pu isotopic composition for source identification of Pu releases in the FDNPP-damaged reactors or spent fuel pools, and estimation of the amounts of Pu isotopes released from the FDNPP accident. Our analysis indicates that a trace amount of Pu isotopes (∼2 × 10(-5)% of core inventory) was released into the environment from the damaged reactors but not from the spent fuel pools located in the reactor buildings. Regarding the possible Pu contamination in the marine environment, limited studies suggest that no extra Pu input from the FDNPP accident could be detected in the western North Pacific 30 km off the Fukushima coast. Finally, we identified knowledge gaps remained on the release of Pu into the environment and recommended issues for future studies.

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

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

DOE PAGES

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek; ...

2017-03-27

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

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

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Radiation Transport Calculation of the UGXR Collimators for the Jules Horowitz Reactor (JHR)

NASA Astrophysics Data System (ADS)

Chento, Yelko; Hueso, César; Zamora, Imanol; Fabbri, Marco; Fuente, Cristina De La; Larringan, Asier

2017-09-01

Jules Horowitz Reactor (JHR), a major infrastructure of European interest in the fission domain, will be built and operated in the framework of an international cooperation, including the development and qualification of materials and nuclear fuel used in nuclear industry. For this purpose UGXR Collimators, two multi slit gamma and X-ray collimation mechatronic systems, will be installed at the JHR pool and at the Irradiated Components Storage pool. Expected amounts of radiation produced by the spent fuel and X-ray accelerator implies diverse aspects need to be verified to ensure adequate radiological zoning and personnel radiation protection. A computational methodology was devised to validate the Collimators design by means of coupling different engineering codes. In summary, several assessments were performed by means of MCNP5v1.60 to fulfil all the radiological requirements in Nominal scenario (TEDE < 25µSv/h) and in Maintenance scenario (TEDE < 2mSv/h) among others, detailing the methodology, hypotheses and assumptions employed.

Characterisation of MR reactor pond in nNRC 'Kurchatov institute' before dismantling work

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

Stepanov, Alexey; Simirsky, Yury; Semin, Ilya

2013-07-01

In this work complex α-, β-, γ-spectrometric research of water, bottom slimes and deposits on walls of the reactor pond and the storage pond of the MR reactor was made. Identify, that the main dose forming radionuclide, during dismantling work on the reactor MR, is Cs-137. It is shown, that specific activity of radionuclides in bottom slimes considerably exceed specific activity of radionuclides in water from ponds, and near to high level radioactive waste. It is detected that decreasing the water level in reactor ponds on 1 m, increase the exposure dose rate at a distance 1 m from themore » pond in 2 times. The observed increase in exposure dose rate can be explained by contribution on dose rate the cesium-137 deposed on walls of the storage pond. Effectiveness of cleaning of walls of the pool of storage from deposits by a water jet of high pressure is investigated. (authors)« less

Cold weather effects on Dresden Unit 1

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

Anagnostopoulos, H.

1995-03-01

Dresden Unit 1 is in the final stages of a decommissioning effort directed at preparing the unit to enter a SAFSTOR status. Following an extended sub-zero cold wave, about 55,000 gallons of water were discovered in the lowest elevation of the spherical reactor enclosure. Cold weather had caused the freezing and breaking of several service water lines that had not been completely isolated. Two days later, at a regularly scheduled decommissioning meeting, the event was communicated to the decommissioning team, who quickly recognized the potential for freezing of a 42 inches diameter Fuel Transfer Tube that connects the sphere tomore » the Spent Fuel Pool. The team directed that the pool gates between the adjacent Spent Fuel Pool and the Fuel Transfer Pool be installed, and a portable source of heat was installed on the Fuel Transfer Tube. It was later determined that, with the fuel pool gates removed, and with a worst case freeze break at the 502 elevation on the Fuel Transfer Tube (in the Sphere), the fuel in the Spent Fuel Pool could be uncovered to a level 3 below the top of active fuel.« less

Electrolytic cell with reference electrode

DOEpatents

Kessie, Robert W.

1989-01-01

A reference electrode device is provided for a high temperature electrolytic cell used to electrolytically recover uranium from spent reactor fuel dissolved in an anode pool, the device having a glass tube to enclose the electrode and electrolyte and serve as a conductive membrane with the cell electrolyte, and an outer metal tube about the glass tube to serve as a shield and basket for any glass sections broken by handling of the tube to prevent their contact with the anode pool, the metal tube having perforations to provide access between the bulk of the cell electrolyte and glass membrane.

Reference electrode for electrolytic cell

DOEpatents

Kessie, R.W.

1988-07-28

A reference electrode device is provided for a high temperature electrolytic cell used to electrolytically recover uranium from spent reactor fuel dissolved in an anode pool, the device having a glass tube to enclose the electrode and electrolyte and serve as a conductive membrane with the cell electrolyte, and an outer metal tube about the glass tube to serve as a shield and basket for any glass sections broken by handling of the tube to prevent their contact with the anode pool, the metal tube having perforations to provide access between the bulk of the cell electrolyte and glass membrane. 4 figs.

Preliminary Design of Critical Function Monitoring System of PGSFR

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

NONE

2015-07-01

A PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) is under development at Korea Atomic Energy Research Institute. A critical function monitoring system of the PGSFR is preliminarily studied. The functions of CFMS are to display critical plant variables related to the safety of the plant during normal and accident conditions and guide the operators corrective actions to keep the plant in a safe condition and mitigate the consequences of accidents. The minimal critical functions of the PGSFR are composed of reactivity control, reactor core cooling, reactor coolant system integrity, primary heat transfer system(PHTS) heat removal, sodium water reaction mitigation, radiation controlmore » and containment conditions. The variables and alarm legs of each critical function of the PGSFR are as follows; - Reactivity control: The variables of reactivity control function are power range neutron flux instrumentation, intermediate range neutron flux instrumentation, source range neutron flux instrumentation, and control rod bottom contacts. The alarm leg to display the reactivity controls consists of status of control drop malfunction, high post trip power and thermal reactivity addition. - Reactor core cooling: The variables are PHTS sodium level, hot pool temperature of PHTS, subassembly exit temperature, cold pool temperature of the PHTS, PHTS pump current, and PHTS pump breaker status. The alarm leg consists of high core delta temperature, low sodium level of the PHTS, high subassembly exit temperature, and low PHTS pump load. - Reactor coolant system integrity: The variables are PHTS sodium level, cover gas pressure, and safeguard vessel sodium level. The alarm leg is composed of low sodium level of PHTS, high cover gas pressure and high sodium level of the safety guard vessel. - PHTS heat removal: The variables are PHTS sodium level, hot pool temperature of PHTS, core exit temperature, cold pool temperature of the PHTS, flow rate of passive residual heat removal system, flow rate of active residual heat removal system, and temperatures of air heat exchanger temperature of residual heat removal systems. The alarm legs are composed of two legs of a 'passive residual heat removal system not cooling' and 'active residual heat removal system not cooling'. - Sodium water reaction mitigation: The variables are intermediate heat transfer system(IHTS) pressure, pressure and temperature and level of sodium dump tank, the status of rupture disk, hydrogen concentration in IHTS and direct variable of sodium-water-reaction measure. The alarm leg consists of high IHTS pressure, the status of sodium water reaction mitigation system and the indication of direct measure. - Radiation control: The variables are radiation of PHTS, radiation of IHTS, and radiation of containment purge. The alarm leg is composed of high radiation of PHTS and IHTS, and containment purge system. - Containment condition: The variables are containment pressure, containment isolation status, and sodium fire. The alarm leg consists of high containment pressure, status of containment isolation and status of sodium fire. (authors)« less

Decommissioning of the High Flux Beam Reactor at Brookhaven National Laboratory.

PubMed

Hu, Jih-Perng; Reciniello, Richard N; Holden, Norman E

2012-08-01

The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory was a heavy-water cooled and moderated reactor that achieved criticality on 31 October 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost 3 y for safety and environmental reviews. In November 1999, the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR's cleanup performed during 1999-2009, to allow the BNL facilities to be re-accessed by the public, will be described in the paper.

Dual-phase reactor plant with partitioned isolation condenser

DOEpatents

Hui, Marvin M.

1992-01-01

A nuclear energy plant housing a boiling-water reactor utilizes an isolation condenser in which a single chamber is partitioned into a distributor plenum and a collector plenum. Steam accumulates in the distributor plenum and is conveyed to the collector plenum through an annular manifold that includes tubes extending through a condenser pool. The tubes provide for a transfer of heat from the steam, forming a condensate. The chamber has a disk-shaped base, a cylindrical sidewall, and a semispherical top. This geometry results in a compact design that exhibits significant performance and cost advantages over prior designs.

CFD Analysis of Upper Plenum Flow for a Sodium-Cooled Small Modular Reactor

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

Kraus, A.; Hu, R.

2015-01-01

Upper plenum flow behavior is important for many operational and safety issues in sodium fast reactors. The Prototype Gen-IV Sodium Fast Reactor (PGSFR), a pool-type, 150 MWe output power design, was used as a reference case for a detailed characterization of upper plenum flow for normal operating conditions. Computational Fluid Dynamics (CFD) simulation was utilized with detailed geometric modeling of major structures. Core outlet conditions based on prior system-level calculations were mapped to approximate the outlet temperatures and flow rates for each core assembly. Core outlet flow was found to largely bypass the Upper Internal Structures (UIS). Flow curves overmore » the shield and circulates within the pool before exiting the plenum. Cross-flows and temperatures were evaluated near the core outlet, leading to a proposed height for the core outlet thermocouples to ensure accurate assembly-specific temperature readings. A passive scalar was used to evaluate fluid residence time from core outlet to IHX inlet, which can be used to assess the applicability of various methods for monitoring fuel failure. Additionally, the gas entrainment likelihood was assessed based on the CFD simulation results. Based on the evaluation of velocity gradients and turbulent kinetic energies and the available gas entrainment criteria in the literature, it was concluded that significant gas entrainment is unlikely for the current PGSFR design.« less

Enrichment of extremophilic exoelectrogens in microbial electrolysis cells using Red Sea brine pools as inocula.

PubMed

Shehab, Noura A; Ortiz-Medina, Juan F; Katuri, Krishna P; Hari, Ananda Rao; Amy, Gary; Logan, Bruce E; Saikaly, Pascal E

2017-09-01

Applying microbial electrochemical technologies for the treatment of highly saline or thermophilic solutions is challenging due to the lack of proper inocula to enrich for efficient exoelectrogens. Brine pools from three different locations (Valdivia, Atlantis II and Kebrit) in the Red Sea were investigated as potential inocula sources for enriching exoelectrogens in microbial electrolysis cells (MECs) under thermophilic (70°C) and hypersaline (25% salinity) conditions. Of these, only the Valdivia brine pool produced high and consistent current 6.8±2.1A/m 2 -anode in MECs operated at a set anode potential of +0.2V vs. Ag/AgCl (+0.405V vs. standard hydrogen electrode). These results show that exoelectrogens are present in these extreme environments and can be used to startup MEC under thermophilic and hypersaline conditions. Bacteroides was enriched on the anode of the Valdivia MEC, but it was not detected in the open circuit voltage reactor seeded with the Valdivia brine pool. Copyright © 2017 Elsevier Ltd. All rights reserved.

Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

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

Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh

2008-07-15

The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configurationmore » with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)« less

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

75 FR 62151 - Notice of Availability of Environmental Assessment and Finding of No Significant Impact for the...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-07

... years of experience in decommissioning health physics practices. All reactor and pool components will be... from lead paint and asbestos. WPI has committed to compliance with applicable occupational health and safety requirements, primarily the federal Occupational Safety and Health Act (OSHA) of 1973. Accordingly...

PBF (PER620) interior. Detail view of actuator platform and control ...

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

PBF (PER-620) interior. Detail view of actuator platform and control rod mechanism. Camera facing easterly from floor level. Reactor pool at lower left of view. Date: March 2004. INEEL negative no. HD-41-3-3 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Clay-based matrices incorporating radioactive silts: A case study of sediments from spent fuel pool

NASA Astrophysics Data System (ADS)

Antonenko, Mikhail; Myshkin, Vyacheslav; Grigoriev, Alexander; Chubreev, Dmitry

2018-03-01

Radioactive silt sediments from uranium reactors may be effectively and safely included by ceramic compounds. The purpose of the paper is to determine the influence of composition and preparation conditions on physicochemical and mechanical properties of clay-based matrices containing radioactive silt. Clay matrices were prepared from four minerals, took from Siberian regions, as kaolin, loan, bentonite and red clay, and they included radioactive silt sediments collected from Spent Fuel Pool of a Uranium-graphite Reactor. The rate of 137Cs leaching from the matrices of different compositions was studied. The results of the studies allowed determining the optimal compositions and the preparation conditions of the matrices. It has been shown that red clay from "Zykovskaya" career (Krasnoyarsk region, Russia) is preferable for use as a matrix for incorporating the silt sediments compared to kaolin, loam and bentonite due to the maximum values tensile strength and minimal change in ultimate strength for compression after irradiation, freezing and water exposure. Nevertheless, 137Cs leaching rate of all studied composites did not exceed 10-3 g/cm2.day.

Reactor pressure vessel nozzle

DOEpatents

Challberg, Roy C.; Upton, Hubert A.

1994-01-01

A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough.

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

Pluquet, Alain

Cette théetudie les techniques d'identication de l'electron dans l'experience D0 au laboratoire Fermi pres de Chicago Le premier chapitre rappelle quelques unes des motivations physiques de l'experience physique des jets physique electrofaible physique du quark top Le detecteur D0 est decrit en details dans le second chapitre Le troisieme cha pitre etudie les algorithmes didentication de lelectron trigger reconstruction ltres et leurs performances Le quatrieme chapitre est consacre au detecteur a radiation de transition TRD construit par le Departement dAstrophysique Physique des Particules Physique Nucleaire et dInstrumentation Associee de Saclay il presente son principe sa calibration et ses performances Ennmore » le dernier chapitre decrit la methode mise au point pour lanalyse des donnees avec le TRD et illustre son emploi sur quelques exemples jets simulant des electrons recherche du quark top« less

Westinghouse Small Modular Reactor nuclear steam supply system design

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

Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development andmore » integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)« less

Strategy proposed by Electricite de France in the development of automatic tools

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

Castaing, C.; Cazin, B.

1995-03-01

The strategy proposed by EDF in the development of a means to limit personal and collective dosimetry is recent. It follows in the steps of a policy that consisted of developing remote operation means for those activities of inspection and maintenance on the reactor, pools bottom, steam generators (SGs), also reactor building valves; target activities because of their high dosimetric cost. One of the main duties of the UTO (Technical Support Department), within the EDF, is the maintenance of Pressurized Water Reactors in French Nuclear Power Plant Operations (consisting of 54 units) and the development and monitoring of specialized tools.more » To achieve this, the UTO has started a national think-tank on the implementation of the ALARA process in its field of activity and created an ALARA Committee responsible for running and monitoring it, as well as a policy for developing tools. This point will be illustrated in the second on reactor vessel heads.« less

Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

PubMed

Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

2017-12-01

The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

A molten salt process for producing neodymium and neodymium-iron alloys

NASA Astrophysics Data System (ADS)

Sharma, Ram A.; Seefurth, Randall N.

1989-12-01

The production of low-cost neodymium metal in a stirred tank reactor by the reduction of Nd2O3 with sodium in the presence of CaCl2-KCl-NaCl melts by the overall reaction Nd2O3+3CaCl2+6Na→2Nd+3CaO+6NaCl at ˜750 °C is described. The metal produced is recovered from the salt medium by dissolving it in a Nd-Zn or Nd-Fe alloy pool. In the case of Nd-Zn alloy pools, product yields (percentages of theoretical neodymium produced) in excess of 94 pct are obtained when using salt ratios, i.e., the amounts of salt per gram of neodymium produced, ≥3.5 and excess reductant ≥10 pct. The alloy produced is of high quality, and following vacuum distillation of the zinc, can be used in producing General Motors’ MAGNEQUENCH alloy for permanent magnets. In the case of Nd-Fe pools, the yield is also ˜95 pct with a salt ratio as low as 3.5. The yield is found to depend on the salt composition and salt ratio, and to decrease at salt ratios below 3.25. Stirrer position has little effect on yield, while increasing the temperature and placing fins in the reactor increase the yield. The Nd-Fe alloy produced is of as good quality as that produced using Ca reductant and is suitable for direct use in preparing the MAGNEQUENCH alloy.

Quelques considerations sur la traduction medicale et pharmaceutique (Some Considerations in Medical and Pharmaceutical Translation)

ERIC Educational Resources Information Center

Sliosberg, A.

1971-01-01

Paper presented during the meeting of the Section Presse et Documentation" of the 29th International Congress of Pharmaceutical Science of the International Pharmaceutical Federation, London, September 10, 1969. (VM)

Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

NASA Astrophysics Data System (ADS)

Spaccapaniccia, C.; Planquart, P.; Buchlin, J. M. AB(; ), AC(; )

2018-01-01

The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

Use of LEU in the aqueous homogeneous medical isotope production reactor

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

Ball, R.M.

1997-08-01

The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its largemore » negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution.« less

Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor using RELAP5 and TEMPEST: Part 1, Models and simulation results

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

Morris, D.G.; Wendel, M.W.; Chen, N.C.J.

A study was conducted to examine decay heat removal requirements in the High Flux Isotope Reactor (HFIR) following shutdown from 85 MW. The objective of the study was to determine when forced flow through the core could be terminated without causing the fuel to melt. This question is particularly relevant when a station blackout caused by an external event is considered. Analysis of natural circulation in the core, vessel upper plenum, and reactor pool indicates that 12 h of forced flow will permit a safe shutdown with some margin. However, uncertainties in the analysis preclude conclusive proof that 12 hmore » is sufficient. As a result of the study, two seismically qualified diesel generators were installed in HFIR. 9 refs., 4 figs.« less

SAFEGUARDS REPORT FOR THE NORTHROP PULSE RADIATION FACILITY

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

Feinauer, E.; Thomas, R.D.

1961-03-22

Ae description is given of the Northrop pulse Radiation Facility, (NPRF), which consists of a TRlGA Mark-F reactor and associated supporting equipment. The NPRF was designed to operate in the following modes: Mode 1-100 kw steady-state operation; Mode II--Pulsed operation up to a maximum transient giving a maximum measured fuel element temperature of 470 deg C, which corresponds to an energy release of about 18 Mw-sec (approximately 1.9% sigma K/ K). The movable reactor will be operated in three general areas in the pool: adjacent to the exposure room; adjacent to the beam ponts; or at intermediate positions. Based onmore » the analyses presented and operating experience with the prototype TRIGA Mark F and other TRlGA reactors, it is concluded that operation of the NPRF does not present any undue hazard to the health and safety of the operating personnel or the public. (auth)« less

Quelques problemes poses a la grammaire casuelle (Some Problems Regarding Case Grammar)

ERIC Educational Resources Information Center

Fillmore, Charles J.

1975-01-01

Discusses problems related to case grammar theory, including: the organizations of a case grammar; determination of semantic roles; definition and hierarchy of cases; cause-effect relations; and formalization and notation. (Text is in French.) (AM)

Electrolysis cell for reprocessing plutonium reactor fuel

DOEpatents

Miller, William E.; Steindler, Martin J.; Burris, Leslie

1986-01-01

An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals, the cell including a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket and the anode basket being extendable into the lower pool to dissolve at least some metallic contaminants, the anode basket containing the spent fuel acting as a second anode when in the electrolyte.

Electrolysis cell for reprocessing plutonium reactor fuel

DOEpatents

Miller, W.E.; Steindler, M.J.; Burris, L.

1985-01-04

An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals is claimed. The cell includes a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket. The anode basket is extendable into the lower pool to dissolve at least some metallic contaminants; the anode basket contains the spent fuel acting as a second anode when in the electrolyte.

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

Schwantes, Jon M.

Kelly Fitzgerald Kelly Fitzgerald assisted with laboratory testing for an ongoing R&D project known as Electrochemically Modulated Separation (EMS) for on-line rapid preseparations of actinides prior to mass spectrometry analysis. Ryne Burgess Ryne Burgess used SCALE 5.1 ORIGEN-ARP to predict isotope libraries for the Units 1, 2 and 3 reactors and Unit 4 spent fuel pool for comparing against measurements of environmental sampled collected at the site in order to identify the source terms of the accident. Comparison of the cesium 134/137 and cesium 136/137 ratios observed in environmental samples and ORIGEN-ARP predictions indicated that the Unit 4 Spent Fuelmore » Pool did not significantly contribute to radionuclide release during the Fukushima Daiichi accident.« less

Investigation on Main Radiation Source at Operation Floor of Fukushima Daiichi Nuclear Power Station Unit 4

NASA Astrophysics Data System (ADS)

Hirayama, Hideo; Kondo, Kenjiro; Suzuki, Seishiro; Hamamoto, Shimpei; Iwanaga, Kohei

2017-09-01

Pulse height distributions were measured using a LaBr3 detector set in a 1 cm lead collimator to investigate main radiation source at the operation floor of Fukushima Daiichi Nuclear Power Station Unit 4. It was confirmed that main radiation source above the reactor well was Co-60 from the activated steam dryer in the DS pool (Dryer-Separator pool) and that at the standby area was Cs-134 and Cs-137 from contaminated buildings and debris at the lower floor. Full energy peak count rate of Co-60 was reduced about 1/3 by 12mm lead sheet placed on the floor of the fuel handling machine.

Reactor pressure vessel nozzle

DOEpatents

Challberg, R.C.; Upton, H.A.

1994-10-04

A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough. 2 figs.

Quenching behavior of molten pool with different strategies – A review

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

Shrikant,, E-mail: 2014rmt9018@mnit.ac.in; Pandel, U.; Duchaniya, R. K.

After the major severe accident in nuclear reactor, there has been lot of concerns regarding long term core melt stabilization following a severe accident in nuclear reactors. Numerous strategies have been though for quenching and stabilization of core melt like top flooding, bottom flooding, indirect cooling, etc. However, the effectiveness of these schemes is yet to be determined properly, for which, lot of experiments are needed. Several experiments have been performed for coolability of melt pool under bottom flooding as well as for indirect cooling. Besides these tests are very scattered because they involve different simulants material initial temperatures andmore » masses of melt, which makes it very complex to judge the effectiveness of a particular technique and advantage over the other. In this review paper, a study has been carried on different cooling techniques of simulant materials with same mass. Three techniques have been compared here and the results are discussed. Under top flooding technique it took several hours to cool the melt under without decay heat condition. In bottom flooding technique was found to be the best technique among in indirect cooling technique, top flooded technique, and bottom flooded technique.« less

Calculation of natural convection test at Phenix using the NETFLOW++ code

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

Mochizuki, H.; Kikuchi, N.; Li, S.

2012-07-01

The present paper describes modeling and analyses of a natural convection of the pool-type fast breeder reactor Phenix. The natural convection test was carried out as one of the End of Life Tests of the Phenix. Objective of the present study is to assess the applicability of the NETFLOW++ code which has been verified thus far using various water facilities and validated using the plant data of the loop-type FBR 'Monju' and the loop-type experimental fast reactor 'Joyo'. The Phenix primary heat transport system is modeled based on the benchmark documents available from IAEA. The calculational model consists of onlymore » the primary heat transport system with boundary conditions on the secondary-side of IHX. The coolant temperature at the primary pump inlet, the primary coolant temperature at the IHX inlet and outlet, the secondary coolant temperatures and other parameters are calculated by the code where the heat transfer between the hot and cold pools is explicitly taken into account. A model including the secondary and tertiary systems was prepared, and the calculated results also agree well with the measured data in general. (authors)« less

Flow characteristics of Korea multi-purpose research reactor

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

Heonil Kim; Hee Taek Chae; Byung Jin Jun

1995-09-01

The construction of Korea Multi-purpose Research Reactor (KMRR), a 30 MW{sub th} open-tank-in-pool type, is completed. Various thermal-hydraulic experiments have been conducted to verify the design characteristics of the KMRR. This paper describes the commissioning experiments to determine the flow distribution of KMRR core and the flow characteristics inside the chimney which stands on top of the core. The core flow is distributed to within {+-}6% of the average values, which is sufficiently flat in the sense that the design velocity in the fueled region is satisfied. The role of core bypass flow to confine the activated core coolant inmore » the chimney structure is confirmed.« less

Heat transfer of molten metal layers in severe accidents

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

Wong, Seung Kai; Walton, A.; Yang, Zhilin

1997-12-01

In some scenarios of severe accidents of light water reactors, a layer of molten metal from internal structural components of the pressure vessel is predicted to occur on top of a ceramic core debris in the lower head. The layer transfers the heat generated in the ceramic pool to the side wall of the vessel, causing the latter to melt. This problem has been investigated by Theofanous et al. for the advanced light water reactor AP600 in the context of the accident management strategy of ex-vessel cooling, and the conclusion was drawn that the melting does not seriously compromise themore » integrity of the pressure vessel.« less

Fast reactor power plant design having heat pipe heat exchanger

DOEpatents

Huebotter, P.R.; McLennan, G.A.

1984-08-30

The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

Fast reactor power plant design having heat pipe heat exchanger

DOEpatents

Huebotter, Paul R.; McLennan, George A.

1985-01-01

The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

Neutron detection of the Triga Mark III reactor, using nuclear track methodology

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

Espinosa, G., E-mail: espinosa@fisica.unam.mx; Golzarri, J. I.; Raya-Arredondo, R.

Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 10{sup 12} n cm{sup −2} s{sup −1}, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer{sup ®} PADC as neutron detection material, covered by 3 mm Plexiglas{sup ®} as converter. After exposure, plasticmore » detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux density measured at low reactor power.« less

Tools for placing the radiological health hazard in perspective following a severe emergency at a light water reactor (LWR) or its spent fuel pool.

PubMed

McKenna, Thomas; Welter, Phillip Vilar; Callen, Jessica; Martincic, Rafael; Dodd, Brian; Kutkov, Vladimir

2015-01-01

Experience from past nuclear and radiological emergencies shows that placing the radiological health hazard in perspective and having a definition of "safe" are required in order to prevent members of the public, those responsible for protecting the public (i.e., decision makers), and others from taking inappropriate and damaging actions that are not justified based on the radiological health hazard. The principle concerns of the public during a severe nuclear power plant or spent fuel pool emergency are "Am I safe?" and "What should I do to be safe?" However, these questions have not been answered to the satisfaction of the public, despite various protective actions being implemented to ensure their safety. Instead, calculated doses or various measured quantities (e.g., ambient dose rate or radionuclide concentrations) are used to describe the situation to the public without placing them into perspective in terms of the possible radiological health hazard, or if they have, it has been done incorrectly. This has contributed to members of the public taking actions that do more harm than good in the belief that they are protecting themselves. Based on established international guidance, this paper provides a definition of "safe" for the radiological health hazard for use in nuclear or radiological emergencies and a system for putting the radiological health hazard in perspective for quantities most commonly measured after a release resulting from a severe emergency at a light water reactor or its spent fuel pool.

Physical and chemical differences between natural and artificial pools in blanket peatlands

NASA Astrophysics Data System (ADS)

Turner, Ed; Baird, Andy; Billett, Mike; Chapman, Pippa; Dinsmore, Kerry; Holden, Joseph

2014-05-01

Natural pools are common features of many northern peatlands. Numerous artificial pools are being created behind dams installed during drain-blocking, a common peatland restoration technique, significantly increasing the area of open water. Natural pools are known to be major sources of GHGs (e.g. Hamilton et al. 1994), but the reasons they are such 'hotspots' is poorly understood. We hypothesize that pools act as 'biochemical reactors' of particulate and dissolved organic carbon (POC and DOC) transported from surrounding peat that is processed into a range of products including CH4 and CO2. Therefore, understanding the processes operating in both natural and artificial pool systems is fundamental to elucidating this hypothesis. Water levels and temperature have been continuously monitored at six natural and six artificial pools within the 'Flow Country' blanket peatland in northern Scotland since May 2013. Bi-weekly sampling of waters from pools, peat matrix through-flow (via piezometers) and surface flow has been conducted for analysis of DOC, POC, DIC, CH4diss and CO2diss, together with GHG flux measurements from pool surfaces and adjacent peat. We show that, to date, pool water levels rapidly respond to rainfall, although artificial pools appear to respond with greater magnitude. For example, over the course of same rainfall event (20-23 June 2013), natural and artificial pool levels increased between 5.3 and 9.8 cm, and 12.5 and 22.6 cm respectively. Temperature measured at c. 5 cm from the base of each pool shows distinct diurnal fluctuations, which are of greater magnitude in all but one of the natural pools compared to the artificial pools: over the same period (20-23 July 2013), the maximum diurnal variation at the artificial pool site was 5.1 °C compared to 9.2 °C within the natural pools. Vegetation cover is generally higher in artificial pools and may have a moderating effect on variations in pool temperature. Results of pool-water DOC analysis from regular sampling at the study site and a wider regional survey indicate DOC concentrations are consistently higher in artificial pools. The implications of these preliminary results in relation to the carbon cycle and GHGs of blanket peatlands are briefly discussed. Hamilton, J. D., Kelly, C. A., Rudd, J. W. M., Hesslein, R. H. and Roulet, N. T. (1994) Flux to the atmosphere of CH4 and CO2 from wetland ponds on the Hudson Bay lowlands (HBLs). Journal of Geophysical Research 99, 1495-1510.

Lévitation magnétique par association d'aimants permanents et de supraconducteurs à haute température critique

NASA Astrophysics Data System (ADS)

Hiebel, P.; Tixador, P.; Chaud, X.

1995-06-01

Since their discovery in the years 1986/87, the high critical temperature superconductors have reached nowadays performances interesting enough to conceive passive magnetic bearings and suspensions which would combined permanent magnets and naturally stable superconducting pellets. After underlining the principal factors that affect the superconductormagnet interaction, different experimental results are given about vertical and axial forces with some stiffness values. The magnetization curve of a superconductor help to understand the hysteretic behavior of the force as a function of the distance between superconductor and magnet. So called simple and hybrid structures of superconducting magnetic suspension are presented. Finally simple numerical simulations allow to draw some interesting conclusions about both geometry and best fitting structure of permanent magnets. Depuis leur découverte dans les années 1986/87, les supraconducteurs à haute température critique ont désormais atteint des performances intéressantes et rendent envisageables des paliers et suspensions magnétiques passives associant aimants permanents et pastilles supraconductrices naturellement stables. Après avoir indiqué les termes importants influençant l'interaction supraconducteur - aimant, différents relevés expérimentaux sont donnés pour les forces verticales et transversales avec quelques valeurs de raideurs. La courbe d'aimantation d'un supraconducteur permet de comprendre le comportement hystérétique de la force en fonction de la distance supraconducteur-aimant. Les structures dites simple et hybride des suspensions magnétiques supraconductrices sont présentées. Enfin quelques simulations numériques simples permettent de dégager quelques conclusions intéressantes quant aux géométries respectives et aux structures d'aimants permanents les mieux adaptées.

A User Guide to PARET/ANL

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

Olson, A. P.; Dionne, B.; Marin-Lafleche, A.

2015-01-01

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA, and has been used by the Reactor Conversion Program tomore » determine the expected transient behavior of a large number of reactors. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.« less

The Potential of the LFR and the ELSY Project

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

Cinotti, L; Smith, C F; Sienicki, J J

2007-03-12

This paper presents the current status of the development of the Lead-cooled Fast Reactor (LFR) in support of Generation IV (GEN IV) Nuclear Energy Systems. The approach being taken by the GIF plan is to address the research priorities of each member state in developing an integrated and coordinated research program to achieve common objectives, while avoiding duplication of effort. The integrated plan being prepared by the LFR Provisional System Steering Committee of the GIF, known as the LFR System research Plan (SRP) recognizes two principal technology tracks for pursuit of LFR technology: (1) a small, transportable system of 10-100more » MWe size that features a very long refueling interval, (2) a larger-sized system rated at about 600 MWe, intended for central station power generation and waste transmutation. This paper, in particular, describes the ongoing activities to develop the Small Secure Transportable Autonomous Reactor (SSTAR) and the European Lead-cooled SYstem (ELSY), the two research initiatives closely aligned with the overall tracks of the SRP and outlines the Proliferation-resistant Environment-friendly Accident-tolerant Continual & Economical Reactors (PEACER) conceived with particular focus on burning/transmuting of long-living TRU waste and fission fragments of concern, such as Tc and I. The current reference design for the SSTAR is a 20 MWe natural circulation pool-type reactor concept with a small shippable reactor vessel. Specific features of the lead coolant, the nitride fuel containing transuranics, the fast spectrum core, and the small size combine to promote a unique approach to achieve proliferation resistance, while also enabling fissile self-sufficiency, autonomous load following, simplicity of operation, reliability, transportability, as well as a high degree of passive safety. Conversion of the core thermal power into electricity at a high plant efficiency of 44% is accomplished utilizing a supercritical carbon dioxide Brayton cycle power converter. The ELSY reference design is a 600 MWe pool-type reactor cooled by pure lead. This concept has been under development since September 2006, and is sponsored by the Sixth Framework Programme of EURATOM. The ELSY project is being performed by a consortium consisting of twenty organizations including seventeen from Europe, two from Korea and one from the USA. ELSY aims to demonstrate the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features while fully complying with the Generation IV goal of minor actinide (MA) burning capability. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Simplicity is expected to reduce both the capital cost and the construction time; these are also supported by the compactness of the reactor building (reduced footprint and height). The reduced footprint would be possible due to the elimination of the Intermediate Cooling System, the reduced elevation the result of the design approach of reduced-height components.« less

The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

DOE PAGES

Bonano, Evaristo J.; Kalinina, Elena A.; Swift, Peter N.

2018-02-26

Current practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-centurymore » when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.« less

Modeling evaporation from spent nuclear fuel storage pools: A diffusion approach

NASA Astrophysics Data System (ADS)

Hugo, Bruce Robert

Accurate prediction of evaporative losses from light water reactor nuclear power plant (NPP) spent fuel storage pools (SFPs) is important for activities ranging from sizing of water makeup systems during NPP design to predicting the time available to supply emergency makeup water following severe accidents. Existing correlations for predicting evaporation from water surfaces are only optimized for conditions typical of swimming pools. This new approach modeling evaporation as a diffusion process has yielded an evaporation rate model that provided a better fit of published high temperature evaporation data and measurements from two SFPs than other published evaporation correlations. Insights from treating evaporation as a diffusion process include correcting for the effects of air flow and solutes on evaporation rate. An accurate modeling of the effects of air flow on evaporation rate is required to explain the observed temperature data from the Fukushima Daiichi Unit 4 SFP during the 2011 loss of cooling event; the diffusion model of evaporation provides a significantly better fit to this data than existing evaporation models.

The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

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

Bonano, Evaristo J.; Kalinina, Elena A.; Swift, Peter N.

Current practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-centurymore » when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.« less

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

Lessons Learned in Protection of the Public for the Accident at the Fukushima Daiichi Nuclear Power Plant.

PubMed

Callen, Jessica; Homma, Toshimitsu

2017-06-01

What insights can the accident at the Fukushima Daiichi nuclear power plant provide in the reality of decision making on actions to protect the public during a severe reactor and spent fuel pool emergency? In order to answer this question, and with the goal of limiting the consequences of any future emergencies at a nuclear power plant due to severe conditions, this paper presents the main actions taken in response to the emergency in the form of a timeline. The focus of this paper is those insights concerning the progression of an accident due to severe conditions at a light water reactor nuclear power plant that must be understood in order to protect the public.

A temps nouveaux, solutions nouvelles: quelques propositions (New Times, New Solutions: Some Proposals).

ERIC Educational Resources Information Center

Capelle, Guy

1983-01-01

Serious problems in education in Latin America arising from political, economic, and social change periodically put in question the status, objectives, and manner of French second-language instruction. A number of solutions to general and specific pedagogical problems are proposed. (MSE)

The R/D of high power proton accelerator technology in China

NASA Astrophysics Data System (ADS)

Xialing, Guan

2002-12-01

In China, a multipurpose verification system as a first phase of our ADS program consists of a low energy accelerator (150 MeV/3 mA proton LINAC) and a swimming pool light water subcritical reactor. In this paper the activities of HPPA technology related to ADS in China, which includes the intense proton ECR source, the RFQ accelerator and some other technology of HPPA, are described.

LMFBR system-wide transient analysis: the state of the art and US validation needs

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

Khatib-Rahbar, M.; Guppy, J.G.; Cerbone, R.J.

1982-01-01

This paper summarizes the computational capabilities in the area of liquid metal fast breeder reactor (LMFBR) system-wide transient analysis in the United States, identifies various numerical and physical approximations, the degree of empiricism, range of applicability, model verification and experimental needs for a wide class of protected transients, in particular, natural circulation shutdown heat removal for both loop- and pool-type plants.

Linear free energy correlations for fission product release from the Fukushima-Daiichi nuclear accident.

PubMed

Abrecht, David G; Schwantes, Jon M

2015-03-03

This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the initial source of the radionuclides to the environment to be from active reactors rather than the spent fuel pool. Linear correlations of the form In χ = −α ((ΔGrxn°(TC))/(RTC)) + β were obtained between the deposited concentrations, and the reduction potentials of the fission product oxide species using multiple reduction schemes to calculate ΔG°rxn (TC). These models allowed an estimate of the upper bound for the reactor temperatures of TC between 2015 and 2060 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, and 151Sm through atmospheric venting during the first month following the accident were obtained, indicating that large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.

Neutron spectrometry and dosimetry study at two research nuclear reactors using Bonner sphere spectrometer (BSS), rotational spectrometer (ROSPEC) and cylindrical nested neutron spectrometer (NNS).

PubMed

Atanackovic, J; Matysiak, W; Hakmana Witharana, S S; Aslam, I; Dubeau, J; Waker, A J

2013-01-01

Neutron spectrometry and subsequent dosimetry measurements were undertaken at the McMaster Nuclear Reactor (MNR) and AECL Chalk River National Research Universal (NRU) Reactor. The instruments used were a Bonner sphere spectrometer (BSS), a cylindrical nested neutron spectrometer (NNS) and a commercially available rotational proton recoil spectrometer. The purposes of these measurements were to: (1) compare the results obtained by three different neutron measuring instruments and (2) quantify neutron fields of interest. The results showed vastly different neutron spectral shapes for the two different reactors. This is not surprising, considering the type of the reactors and the locations where the measurements were performed. MNR is a heavily shielded light water moderated reactor, while NRU is a heavy water moderated reactor. The measurements at MNR were taken at the base of the reactor pool, where a large amount of water and concrete shielding is present, while measurements at NRU were taken at the top of the reactor (TOR) plate, where there is only heavy water and steel between the reactor core and the measuring instrument. As a result, a large component of the thermal neutron fluence was measured at MNR, while a negligible amount of thermal neutrons was measured at NRU. The neutron ambient dose rates at NRU TOR were measured to be between 0.03 and 0.06 mSv h⁻¹, while at MNR, these values were between 0.07 and 2.8 mSv h⁻¹ inside the beam port and <0.2 mSv h⁻¹ between two operating beam ports. The conservative uncertainty of these values is 15 %. The conservative uncertainty of the measured integral neutron fluence is 5 %. It was also found that BSS over-responded slightly due to a non-calibrated response matrix.

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.

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

Marques, J.G.; Ramos, A.R.; Fernandes, A.C.

The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, sincemore » the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor power operational fluctuations; - Study of gamma radiation effects only. The fission neutron spectrum can be limitative for some of the tests, as most neutrons are in the 1-2 MeV energy range. Significant progress has been made lately in compact neutron generators using D-D and D-T fusion reactions, achieving higher neutron fluxes and longer lifetime than previously available. The advantages of using compact neutron generators for testing of electronic components and circuits will be also discussed. (authors)« less

Opportunities for Materials Science and Biological Research at the OPAL Research Reactor

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

Kennedy, S. J.

Neutron scattering techniques have evolved over more than 1/2 century into a powerful set of tools for determination of atomic and molecular structures. Modern facilities offer the possibility to determine complex structures over length scales from {approx}0.1 nm to {approx}500 nm. They can also provide information on atomic and molecular dynamics, on magnetic interactions and on the location and behaviour of hydrogen in a variety of materials. The OPAL Research Reactor is a 20 megawatt pool type reactor using low enriched uranium fuel, and cooled by water. OPAL is a multipurpose neutron factory with modern facilities for neutron beam research,more » radioisotope production and irradiation services. The neutron beam facility has been designed to compete with the best beam facilities in the world. After six years in construction, the reactor and neutron beam facilities are now being commissioned, and we will commence scientific experiments later this year. The presentation will include an outline of the strengths of neutron scattering and a description of the OPAL research reactor, with particular emphasis on it's scientific infrastructure. It will also provide an overview of the opportunities for research in materials science and biology that will be possible at OPAL, and mechanisms for accessing the facilities. The discussion will emphasize how researchers from around the world can utilize these exciting new facilities.« less

An investigation into the feasibility of thorium fuels utilization in seed-blanket configurations for TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

Thorium is one of the elements that needs to be explored for nuclear fuel research and development. One of the popular core configurations of thorium fuel is seed-blanket configuration or also known as Radkowsky Thorium Fuel concept. The seed will act as a supplier of neutrons, which will be placed inside of the core. The blanket, on the other hand, is the consumer of neutrons that is located at outermost of the core. In this work, a neutronic analysis of seed-blanket configuration for the TRIGA PUSPATI Reactor (RTP) is carried out using Monte Carlo method. The reactor, which has been operated since 1982 use uranium zirconium hydride (U-ZrH1.6) as the fuel and have multiple uranium weight which are 8.5, 12 and 20 wt.%. The pool type reactor is one and only research reactor that located in Malaysia. The design of core included the Uranium Zirconium Hydride located at the centre of the core that will act as the seed to supply neutron. The thorium oxide that will act as blanket situated outside of seed region will receive neutron to transmute 232Th to 233U. The neutron multiplication factor or criticality of each configuration is estimated. Results show that the highest initial criticality achieved is 1.30153.

Traduction et langues de specialite: Approches theoriques et considerations pedagogiques (Translation and Specialty Languages: Theoretical Approaches and Pedagogic Considerations).

ERIC Educational Resources Information Center

Guevel, Zelie, Ed.; Valentine, Egan, Ed.

Essays on the teaching of translation and on specialized translation, all in French, include: "Perspectives d'optimisation de la formation du traducteur: quelques reflexions" ("Perspectives on Optimization of Training of Translation Teachers: Some Reflections") (Egan Valentine); "L'enseignement de la revision…

Psychologie des discours et didactique des textes (Psychology of Discourse and the Teaching of Texts).

ERIC Educational Resources Information Center

Bronckart, Jean-Paul, Ed.

1995-01-01

This collection of articles on the nature of discourse and writing instruction include: "Une demarche de psychologie de discours; quelques aspects introductifs" ("An Application of Discourse Psychology; Introductory Thoughts") (Jean-Paul Bronckart); "Les procedes de prise en charge enonciative dans trois genres de texts expositifs" ("The Processes…

[Not Available].

PubMed

Ettalbi, S; Ibnouzahir, M; Droussi, H; Wahbi, S; Bahaichar, N; Boukind, E H

2009-06-30

La brûlure est un accident qui reste toujours très fréquent au Maroc, ce qui fait d'elle un problème de la santé publique. Les brûlures, quand elles sont graves ou profondes, entraînent de façon quasi inéluctable des séquelles fonctionnelles et esthétiques. A travers deux observations de deux enfants présentant des séquelles de brûlures graves, ayant retenti péjorativement sur leurs scolarités, on a essayé de mettre en évidence quelques facteurs incriminés dans cette tragédie (feu, petites bouteilles de gaz et le manque d'infrastructure, du personnel médical et paramédical, du matériel ainsi que de la prévention) comme étant une grande cause dans la survenue de ces séquelles. Le but de notre travail est d'énumérer ces différents facteurs intriqués, ainsi que de proposer quelques solutions, tout en insistant sur la prévention.

Recent upgrades and new scientific infrastructure of MARIA research reactor, Otwock-Swierk, Poland

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

NONE

The MARIA reactor is open-pool type, water and beryllium moderated. It has two independent primary cooling systems: fuel and pool cooling system. Each fuel assembly is cooled down separately in pressurized channels with individual performances characterization. The fuel assemblies consist of five layers of bent plates or six concentric tubes. Currently it is one of the most powerful research reactors in Europe with operation availability at least up to 2030. Its nominal thermal power is 30 MW. It is characterized by high neutron flux density: up to 3x10{sup 14} n cm{sup -2} s{sup -1} in case of thermal neutrons, andmore » up to 2x10{sup 13} n cm{sup -2} s{sup -1} in case of fast neutrons. The reactor is operated for ca. 4000 h per year. The reactor facility is equipped with fully equipped three hot cells with shielding up to 10{sup 15} Bq. Adjacent to the reactor facility, the radio-pharmaceutics plant (POLATOM) and Material Research Laboratory are located. They are equipped with a number of hot cells with instrumentation. The transport system of radioactive materials from reactor facility to Material Research Laboratory is available. During 2014 the MARIA reactor has been operated with three different types of fuel the same time: previous 36% enriched fuel, and two types of new LEU fuels. In the meantime, molybdenum irradiation programme has been developed. Maria is a multifunctional research tool, with a notable application in production of radioisotopes, radio-pharmaceutics manufacturing (ca. 600 TBq/y), {sup 99}Mo for medical scintigraphy (ca. 6000 TBq/y), neutron transmutation doping of silicon single crystals, wide scientific research based on neutron beams utilization. From the beginning MARIA reactor was intended for loop and fuel testing research activities. Currently it is used mostly as material testing and irradiation facility and for that reason it has wide experimental capabilities. There are eight horizontal irradiation channels from among whom six of them are equipped with instrumentation for condensed matter physics research: - H3 - spectrometer and diffractometer with double monochromator; - H4 - small angle scattering spectrometer; - H5 - polarized neutrons spectrometer; - H6, H7 - two 3-axial crystal neutron spectrometers; - H8 - neutron radiography stand. For two horizontal channels are ongoing exploitation programs: - H2 - station with epithermal neutron beam produced in uranium converter is being developed. Intelligent converter will be installed on the periphery of reactor core. The intensity of the beam will be at the level 2x10{sup 9} n cm{sup -2}s{sup -1} what makes the beam unique in the Europe. - H1 - special pneumatic horizontal mail is being developed for irradiation material samples in the vicinity of the core i.e. in the distal part of the H1 channel. The number of neutron irradiation facilities in MARIA reactor is increasing every year. Numerous of thermal neutron irradiation channels including fast hydraulic rabbit system and large size channels for fast neutron irradiation are used routinely. Recently new in-pile facility with ITER-like neutron energy spectrum for 14 MeV neutron irradiation has been constructed. Taking into account its performance and ability of almost incessant operation the facility appears as one of the most powerful 14 MeV neutron sources. The facility shall be used for material research connected with thermonuclear devices (ITER) and 4. generation nuclear reactors. The system of independent fuels channels used in MARIA reactor appear to be very flexible and very convenient to be used as irradiation channels for uranium targets for {sup 99}Mo production. Currently, MARIA reactor supplies ca. 18% world production of {sup 99}Mo. The MARIA reactor research activities are still extended. The current scientific projects are connected e.g. with silicon neutron transmutation doping, in-pile gamma heating measurements, French calculation codes implementation (TRIPOLI4, APOLLO2). The horizontal neutron beams utilization is also developed. The MARIA reactor, due to its primary application connected with loop and fuel testing, is very convenient for testing the nuclear instrumentation, control and measurement systems.« less

Numerical Simulation of Hydrodynamics of a Heavy Liquid Drop Covered by Vapor Film in a Water Pool

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

Ma, W.M.; Yang, Z.L.; Giri, A.

2002-07-01

A numerical study on the hydrodynamics of a droplet covered by vapor film in water pool is carried out. Two level set functions are used as to implicitly capture the interfaces among three immiscible fluids (melt-drop, vapor and coolant). This approach leaves only one set of conservation equations for the three phases. A high-order Navier-Stokes solver, called Cubic-Interpolated Pseudo-Particle (CIP) algorithm, is employed in combination with level set approach, which allows large density ratios (up to 1000), surface tension and jump in viscosity. By this calculation, the hydrodynamic behavior of a melt droplet falling into a volatile coolant is simulated,more » which is of great significance to reveal the mechanism of steam explosion during a hypothetical severe reactor accident. (authors)« less

Pore diffusion limits removal of monochloramine in treatment of swimming pool water using granular activated carbon.

PubMed

Skibinski, Bertram; Götze, Christoph; Worch, Eckhard; Uhl, Wolfgang

2018-04-01

Overall apparent reaction rates for the removal of monochloramine (MCA) in granular activated carbon (GAC) beds were determined using a fixed-bed reactor system and under conditions typical for swimming pool water treatment. Reaction rates dropped and quasi-stationary conditions were reached quickly. Diffusional mass transport in the pores was shown to be limiting the overall reaction rate. This was reflected consistently in the Thiele modulus, in the effect of temperature, pore size distribution and of grain size on the reaction rates. Pores <2.5 times the diameter of the monochloramine molecule were shown to be barely accessible for the monochloramine conversion reaction. GACs with a significant proportion of large mesopores were found to have the highest overall reactivity for monochloramine removal. Copyright © 2017 Elsevier Ltd. All rights reserved.

LPT. Aerial of low power test facility (TAN640 and 641) ...

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

LPT. Aerial of low power test facility (TAN-640 and -641) and shield test facility (TAN-645 and -646). Camera facing south. Low power reactor cells at left, then one-story control building; diagonal fence; shield test control building, then (high-bay) pool room. In foreground are electrical pad, water tanks and guard house. Photographer: Lowin. Date: February 24, 1965. INEEL negative no. 65-987 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Israel’s Attack on Osiraq: A Model for Future Preventive Strikes?

DTIC Science & Technology

2004-09-01

destroying Israel. July 28, 1980 Israeli Foreign Minister Yitzhak Shamir met with French Ambassador to Israel, Jean-Pierre Chauvet . Shamir told Chauvet ... Chauvet argued, “Acquisition of nuclear arms would be lunacy on the part of Iraq. After all, Israel’s Jewish and Arab populations are intermingled, and... caved in and a destroyed cooling pool.57 However, Perlmutter claims a specially equipped F-15 flew by the reactor after the bombing on a special

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

Jaech, J.L.

The use of a pooling technique in leak testing Plutonium Recycle Test Reactor fuel elements to reduce the number of tests is discussed. Since the proportion of defectives in this case is small, application of the method would suggest that the group size be large. It was suggested that additional savings might be introduced by subgrouping the originally grouped items in the event of a positive result, rather than testing them individually. An investigation was made to determine optimum subgrouping sizes. (M.C.G.)

Operation TOMODACHI: A Model for American Disaster Response Efforts and the Collective use of Military Forces Abroad

DTIC Science & Technology

2012-01-01

Plant in Fukushima Daiichi (approximately 170 miles North of Tokyo). The plant consisted of six nuclear reactors and a series of spent-fuel pools...should be praised for the decision to allow family members to voluntarily evacuate areas within 200 miles of the Fukushima - Daiichi Nuclear Plant... Disaster ” (power point presentation, Airlift Tanker Association, Nashville, TN, November 4, 2011) 3 Hisaya Sugiyama, “AIA Summary of Fukushima

MELCOR model for an experimental 17x17 spent fuel PWR assembly.

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

Cardoni, Jeffrey

2010-11-01

A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

Current status of the development of high density LEU fuel for Russian research reactors

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

Vatulin, A.; Dobrikova, I.; Suprun, V.

2008-07-15

One of the main directions of the Russian RERTR program is to develop U-Mo fuel and fuel elements/FA with this fuel. The development is carried out both for existing reactors, and for new advanced designs of reactors. Many organizations in Russia, i.e. 'TVEL', RDIPE, RIAR, IRM, NPCC participate in the work. Two fuels are under development: dispersion and monolithic U-Mo fuel, as well two types of FA to use the dispersion U-Mo fuel: with tubular type fuel elements and with pin type fuel elements. The first stage of works was successfully completed. This stage included out-pile, in-pile and post irradiationmore » examinations of U-Mo dispersion fuel in experimental tubular and pin fuel elements under parameters similar to operation conditions of Russian design pool-type research reactors. The results received both in Russia and abroad enabled to go on to the next stage of development which includes irradiation tests both of full-scale IRT pin-type and tube-type fuel assemblies with U-Mo dispersion fuel and of mini-fuel elements with modified U-Mo dispersion fuel and monolithic fuel. The paper gives a generalized review of the results of U-Mo fuel development accomplished by now. (author)« less

Development of a Research Reactor Protocol for Neutron Multiplication Measurements

DOE PAGES

Arthur, Jennifer Ann; Bahran, Rian Mustafa; Hutchinson, Jesson D.; ...

2018-03-20

A new series of subcritical measurements has been conducted at the zero-power Walthousen Reactor Critical Facility (RCF) at Rensselaer Polytechnic Institute (RPI) using a 3He neutron multiplicity detector. The Critical and Subcritical 0-Power Experiment at Rensselaer (CaSPER) campaign establishes a protocol for advanced subcritical neutron multiplication measurements involving research reactors for validation of neutron multiplication inference techniques, Monte Carlo codes, and associated nuclear data. There has been increased attention and expanded efforts related to subcritical measurements and analyses, and this work provides yet another data set at known reactivity states that can be used in the validation of state-of-the-art Montemore » Carlo computer simulation tools. The diverse (mass, spatial, spectral) subcritical measurement configurations have been analyzed to produce parameters of interest such as singles rates, doubles rates, and leakage multiplication. MCNP ®6.2 was used to simulate the experiment and the resulting simulated data has been compared to the measured results. Comparison of the simulated and measured observables (singles rates, doubles rates, and leakage multiplication) show good agreement. This work builds upon the previous years of collaborative subcritical experiments and outlines a protocol for future subcritical neutron multiplication inference and subcriticality monitoring measurements on pool-type reactor systems.« less

Validation of DRAGON4/DONJON4 simulation methodology for a typical MNSR by calculating reactivity feedback coefficient and neutron flux

NASA Astrophysics Data System (ADS)

Al Zain, Jamal; El Hajjaji, O.; El Bardouni, T.; Boukhal, H.; Jaï, Otman

2018-06-01

The MNSR is a pool type research reactor, which is difficult to model because of the importance of neutron leakage. The aim of this study is to evaluate a 2-D transport model for the reactor compatible with the latest release of the DRAGON code and 3-D diffusion of the DONJON code. DRAGON code is then used to generate the group macroscopic cross sections needed for full core diffusion calculations. The diffusion DONJON code, is then used to compute the effective multiplication factor (keff), the feedback reactivity coefficients and neutron flux which account for variation in fuel and moderator temperatures as well as the void coefficient have been calculated using the DRAGON and DONJON codes for the MNSR research reactor. The cross sections of all the reactor components at different temperatures were generated using the DRAGON code. These group constants were used then in the DONJON code to calculate the multiplication factor and the neutron spectrum at different water and fuel temperatures using 69 energy groups. Only one parameter was changed where all other parameters were kept constant. Finally, Good agreements between the calculated and measured have been obtained for every of the feedback reactivity coefficients and neutron flux.

Linear Free Energy Correlations for Fission Product Release from the Fukushima-Daiichi Nuclear Accident

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

Abrecht, David G.; Schwantes, Jon M.

This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes, et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the source of the radionuclides to be from active reactors rather than the spent fuel pool. Linear correlations of the form ln χ = -α (ΔG rxn°(T C))/(RT C)+β were obtained between the deposited concentration and the reduction potential of the fission product oxide species using multiple reduction schemes to calculate ΔG° rxn(T C). These models allowedmore » an estimate of the upper bound for the reactor temperatures of T C between 2130 K and 2220 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, 151Sm through atmospheric venting and releases during the first month following the accident were performed, and indicate large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.« less

Development of a Research Reactor Protocol for Neutron Multiplication Measurements

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

Arthur, Jennifer Ann; Bahran, Rian Mustafa; Hutchinson, Jesson D.

A new series of subcritical measurements has been conducted at the zero-power Walthousen Reactor Critical Facility (RCF) at Rensselaer Polytechnic Institute (RPI) using a 3He neutron multiplicity detector. The Critical and Subcritical 0-Power Experiment at Rensselaer (CaSPER) campaign establishes a protocol for advanced subcritical neutron multiplication measurements involving research reactors for validation of neutron multiplication inference techniques, Monte Carlo codes, and associated nuclear data. There has been increased attention and expanded efforts related to subcritical measurements and analyses, and this work provides yet another data set at known reactivity states that can be used in the validation of state-of-the-art Montemore » Carlo computer simulation tools. The diverse (mass, spatial, spectral) subcritical measurement configurations have been analyzed to produce parameters of interest such as singles rates, doubles rates, and leakage multiplication. MCNP ®6.2 was used to simulate the experiment and the resulting simulated data has been compared to the measured results. Comparison of the simulated and measured observables (singles rates, doubles rates, and leakage multiplication) show good agreement. This work builds upon the previous years of collaborative subcritical experiments and outlines a protocol for future subcritical neutron multiplication inference and subcriticality monitoring measurements on pool-type reactor systems.« less

The effect of core configuration on temperature coefficient of reactivity in IRR-1

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

Bettan, M.; Silverman, I.; Shapira, M.

1997-08-01

Experiments designed to measure the effect of coolant moderator temperature on core reactivity in an HEU swimming pool type reactor were performed. The moderator temperature coefficient of reactivity ({alpha}{sub {omega}}) was obtained and found to be different in two core loadings. The measured {alpha}{sub {omega}} of one core loading was {minus}13 pcm/{degrees}C at the temperature range of 23-30{degrees}C. This value of {alpha}{sub {omega}} is comparable to the data published by the IAEA. The {alpha}{sub {omega}} measured in the second core loading was found to be {minus}8 pcm/{degrees}C at the same temperature range. Another phenomenon considered in this study is coremore » behavior during reactivity insertion transient. The results were compared to a core simulation using the Dynamic Simulator for Nuclear Power Plants. It was found that in the second core loading factors other than the moderator temperature influence the core reactivity more than expected. These effects proved to be extremely dependent on core configuration and may in certain core loadings render the reactor`s reactivity coefficient undesirable.« less

Motor fuels and chemicals from coal via the Sasol Synthol route

NASA Astrophysics Data System (ADS)

Hoogendoorn, J. C.

1981-03-01

The production of synthetic motor fuels and chemicals from coal by the Sasol procedures is discussed. This process is based on the Fischer-Tropsch reaction by passing hydrogen and carbon monoxide in a specific ratio over iron catalysts at elevated temperatures and pressures. Two parallel reactor systems are discussed. The smaller system employs fixed-bed reactors, using a precipitated iron catalyst and produces predominantly heavy hydrocarbons of an aliphatic nature with carbon chains up to 100. These straight-chain hydrocarbons yield excellent waxes and high quality diesel oil. The larger system uses a powdered iron catalyst in a circulating fluid-bed reactor, a concept developed from American catalytic cracker technology. This system has the advantage of high production capacity and scale-up potential, and produces light olefins which can be used either as petrochemical feedstock or refined and added to the motor fuel pool, and ethylene which is augmented by ethane cracking. Analysis of product selectivities and values shows that co-production of chemicals and motor fuels from coal is profitable and efficient.

Decommissioning ALARA programs Cintichem decommissioning experience

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

Adler, J.J.; LaGuardia, T.S.

1995-03-01

The Cintichem facility, originally the Union Carbide Nuclear Company (UCNC) Research Center, consisted primarily of a 5MW pool type reactor linked via a four-foot-wide by twelve-foot-deep water-filled canal to a bank of five adjacent hot cells. Shortly after going into operations in the early 1960s, the facility`s operations expanded to provide various reactor-based products and services to a multitude of research, production, medical, and education groups. From 1968 through 1972, the facility developed a process of separating isotopes from mixed fission products generated by irradiating enriched Uranium target capsules. By the late 1970s, 20 to 30 capsules were being processedmore » weekly, with about 200,000 curies being produced per week. Several isotopes such as Mo{sup 99}, I{sup 131}, and Xe{sup 133} were being extracted for medical use.« less

L'acquisition d'une language seconde: Quelques developpements theoriques recents (Second Language Acquisition: Some Recent Theoretical Developments).

ERIC Educational Resources Information Center

Py, Bernard, Ed.

1994-01-01

This collection of articles on second language learning includes: "Action, langage et discours. Les fondements d'une psychologie du langage" ("Action, Language, and Discourse. Foundations of a Psychology of Language") (Jean-Paul Bronckart); "Contextes socio-culturels et appropriation des languages secondes: l'apprentissage en milieu social et la…

Canadian Association for the Study of Adult Education. Proceedings of the Annual Conference (4th, Montreal, Quebec, Canada, May 28-30, 1985).

ERIC Educational Resources Information Center

Canadian Association for the Study of Adult Education, Guelph (Ontario).

These proceedings contain 28 papers (20 in English and 8 in French), including the following: "Beyond Ideology: The Case of the Corporate Classroom" (Zinman); "De quelques dimensions paradoxales de l'education interculturelle" (Ollivier); "Ideology, Indoctrination and the Language of Physics" (Winchester);…

IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

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

Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra

IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors.more » The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)« less

Development of a Reduced-Order Three-Dimensional Flow Model for Thermal Mixing and Stratification Simulation during Reactor Transients

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

Hu, Rui

2017-09-03

Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developedmore » at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the multi-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.« less

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

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

Licht, J. R.; Bergeron, A.; Dionne, B.

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cmmore » 2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).« less

An efficient modeling method for thermal stratification simulation in a BWR suppression pool

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

Haihua Zhao; Ling Zou; Hongbin Zhang

2012-09-01

The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safetymore » analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.« less

SIKA—the multiplexing cold-neutron triple-axis spectrometer at ANSTO

NASA Astrophysics Data System (ADS)

Wu, C.-M.; Deng, G.; Gardner, J. S.; Vorderwisch, P.; Li, W.-H.; Yano, S.; Peng, J.-C.; Imamovic, E.

2016-10-01

SIKA is a new cold-neutron triple-axis spectrometer receiving neutrons from the cold source CG4 of the 20MW Open Pool Australian Light-water reactor. As a state-of-the-art triple-axis spectrometer, SIKA is equipped with a large double-focusing pyrolytic graphite monochromator, a multiblade pyrolytic graphite analyser and a multi-detector system. In this paper, we present the design, functions, and capabilities of SIKA, and discuss commissioning experimental results from powder and single-crystal samples to demonstrate its performance.

Noble gas, iodine, and cesium transport in a postulated loss of decay heat removal accident at Browns Ferry

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

Wichner, R.P.; Hodge, S.A.; Weber, C.F.

1984-08-01

This report presents an analysis of the movement of noble gas, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto themore » drywell floor. The analysis of fission product transport presented in this report is based on the no-operator-action sequence and provides an estimate of fission product inventories, as a function of time, within 14 control volumes outside the core, with the atmosphere considered as the final control volume in the transport sequence. As in the case of accident sequences previously studied, we find small barrier for noble gas ejection to air, these gases being effectively purged from the drywell and reactor building by steam and concrete degradation gases. However, significant decay of krypton isotopes occurs during the long delay times involved in this sequence. In contrast, large degrees of holdup for iodine and cesium are projected due to the chemical reactivity of these elements. Only about 2 x 10/sup -4/% of the initial iodine and cesium activity are predicted to be released to the atmosphere. Principal barriers for release are deposition on reactor vessel and containment walls. A significant amount of iodine is captured in the water pool formed in the reactor building basement after actuation of the fire protection system.« less

A propos de quelques experiences de francais fonctionnel en milieu hispanaphone (A Commentary on Experiments in Functional French in a Spanish-Speaking Milieu)

ERIC Educational Resources Information Center

Baltzer, Francois

1978-01-01

A discussion of the adaptation of audiovisual methods to respond to various specific needs in Mexico City. Some of the topics discussed are: meeting needs of people involved in special fields, particularly science, technology and economics; and the use of television for functional French instruction. (AMH)

Aspect Epidemiologique des Sequelles de Brulures a Marrakech, Maroc, a Travers Deux Observations

PubMed Central

Ettalbi, S.; Ibnouzahir, M.; Droussi, H.; Wahbi, S.; Bahaichar, N.; Boukind, E.H.

2009-01-01

Summary La brûlure est un accident qui reste toujours très fréquent au Maroc, ce qui fait d'elle un problème de la santé publique. Les brûlures, quand elles sont graves ou profondes, entraînent de façon quasi inéluctable des séquelles fonctionnelles et esthétiques. A travers deux observations de deux enfants présentant des séquelles de brûlures graves, ayant retenti péjorativement sur leurs scolarités, on a essayé de mettre en évidence quelques facteurs incriminés dans cette tragédie (feu, petites bouteilles de gaz et le manque d'infrastructure, du personnel médical et paramédical, du matériel ainsi que de la prévention) comme étant une grande cause dans la survenue de ces séquelles. Le but de notre travail est d'énumérer ces différents facteurs intriqués, ainsi que de proposer quelques solutions, tout en insistant sur la prévention. PMID:21991156

Fukushima Accident: Sequence of Events and Lessons Learned

NASA Astrophysics Data System (ADS)

Morse, Edward C.

2011-10-01

The Fukushima Dai-Ichi nuclear power station suffered a devastating Richter 9.0 earthquake followed by a 14.0 m tsunami on 11 March 2011. The subsequent loss of power for emergency core cooling systems resulted in damage to the fuel in the cores of three reactors. The relief of pressure from the containment in these three reactors led to sufficient hydrogen gas release to cause explosions in the buildings housing the reactors. There was probably subsequent damage to a spent fuel pool of a fourth reactor caused by debris from one of these explosions. Resultant releases of fission product isotopes in air were significant and have been estimated to be in the 3 . 7 --> 6 . 3 ×1017 Bq range (~10 MCi) for 131I and 137Cs combined, or approximately one tenth that of the Chernobyl accident. A synopsis of the sequence of events leading up to this large release of radioactivity will be presented, along with likely scenarios for stabilization and site cleanup in the future. Some aspects of the isotope monitoring programs, both locally and at large, will also be discussed. An assessment of radiological health risk for the plant workers as well as the general public will also be presented. Finally, the impact of this accident on design and deployment of nuclear generating stations in the future will be discussed.

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

Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

NASA Astrophysics Data System (ADS)

Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

2016-01-01

The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

Containment Sodium Chemistry Models in MELCOR.

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

Louie, David; Humphries, Larry L.; Denman, Matthew R

To meet regulatory needs for sodium fast reactors’ future development, including licensing requirements, Sandia National Laboratories is modernizing MELCOR, a severe accident analysis computer code developed for the U.S. Nuclear Regulatory Commission (NRC). Specifically, Sandia is modernizing MELCOR to include the capability to model sodium reactors. However, Sandia’s modernization effort primarily focuses on the containment response aspects of the sodium reactor accidents. Sandia began modernizing MELCOR in 2013 to allow a sodium coolant, rather than water, for conventional light water reactors. In the past three years, Sandia has been implementing the sodium chemistry containment models in CONTAIN-LMR, a legacy NRCmore » code, into MELCOR. These chemistry models include spray fire, pool fire and atmosphere chemistry models. Only the first two chemistry models have been implemented though it is intended to implement all these models into MELCOR. A new package called “NAC” has been created to manage the sodium chemistry model more efficiently. In 2017 Sandia began validating the implemented models in MELCOR by simulating available experiments. The CONTAIN-LMR sodium models include sodium atmosphere chemistry and sodium-concrete interaction models. This paper presents sodium property models, the implemented models, implementation issues, and a path towards validation against existing experimental data.« less

Benchmarking criticality analysis of TRIGA fuel storage racks.

PubMed

Robinson, Matthew Loren; DeBey, Timothy M; Higginbotham, Jack F

2017-01-01

A criticality analysis was benchmarked to sub-criticality measurements of the hexagonal fuel storage racks at the United States Geological Survey TRIGA MARK I reactor in Denver. These racks, which hold up to 19 fuel elements each, are arranged at 0.61m (2 feet) spacings around the outer edge of the reactor. A 3-dimensional model was created of the racks using MCNP5, and the model was verified experimentally by comparison to measured subcritical multiplication data collected in an approach to critical loading of two of the racks. The validated model was then used to show that in the extreme condition where the entire circumference of the pool was lined with racks loaded with used fuel the storage array is subcritical with a k value of about 0.71; well below the regulatory limit of 0.8. A model was also constructed of the rectangular 2×10 fuel storage array used in many other TRIGA reactors to validate the technique against the original TRIGA licensing sub-critical analysis performed in 1966. The fuel used in this study was standard 20% enriched (LEU) aluminum or stainless steel clad TRIGA fuel. Copyright © 2016. Published by Elsevier Ltd.

The pre-conceptual design of the nuclear island of ASTRID

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

Saez, M.; Menou, S.; Uzu, B.

The CEA is involved in a substantial effort on the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) pre-conceptual design in cooperation with EDF, as experienced Sodium-cooled Fast Reactor (SFR) operator, AREVA, as experienced SFR Nuclear Island engineering company and components designer, ALSTOM POWER as energy conversion system designer and COMEX NUCLEAIRE as mechanical systems designer. The CEA is looking for other partnerships, in France and abroad. The ASTRID preliminary design is based on a sodium-cooled pool reactor of 1500 MWth generating about 600 MWe, which is required to guarantee the representativeness of the reactor core and the main componentsmore » with regard to future commercial reactors. ASTRID lifetime target is 60 years. Two Energy Conversion Systems are studied in parallel until the end of 2012: Rankine steam cycle or Brayton gas based energy conversion cycle. ASTRID design is guided by the following major objectives: improved safety, simplification of structures, improved In Service Inspection and Repair (ISIR), improved manufacturing conditions for cost reduction and increased quality, reduction of risks related to sodium fires and water/sodium reaction, and improved robustness against external hazards. The core is supported by a diagrid, which lay on a strong back to transfer the weight to the main vessel. AREVA is involved in a substantial effort in order to improve the core support structure in particular regarding the ISIR and the connection to primary pump. In the preliminary design, the primary system is formed by the main vessel and the upper closure comprising the reactor roof, two rotating plugs - used for fuel handling - and the components plugs located in the roof penetrations. The Above Core Structure deflects the sodium flow in the hot pool and provides support to core instrumentation and guidance of the control rod drive mechanisms. The number of the major components in the main vessel, primary pumps, Intermediate Heat Exchangers, and Decay Heat Exchangers are now under consideration. Under normal conditions, power release is achieved using the steam/water plant (in case of Rankine steam cycle) or the gas plant (in case of Brayton gas cycle). The diverse design and operating modes of Decay Heat Removal systems provide protection against common cause failures. A Decay Heat Removal system through the reactor vault is in particular studied with the objective to complement Direct Reactor Cooling systems. At this stage of the studies, the secondary system comprises four independent sodium loops (two and three sodium loops configurations are also investigated). Each loop includes one mechanical pump (or a large capacity Annular Linear Induction Electromagnetic Pump), and three modular Steam Generator Units characterized by once through straight tube units with a ferritic tube bundle; nevertheless, helical coil steam generator with tubes made of Alloy 800, and inverted type steam generator with a ferritic tube bundle are also investigated. The limited power of each modular Steam Generator Unit allows the whole secondary loop to withstand a large water/sodium reaction consecutive to the postulated simultaneous rupture of all the heat exchange tubes of one module. The arrangement of the components is based on the 'Regain' concept, in which the secondary pump is situated at a low level in the circuit; conventional arrangement, as SUPERPHENIX type, is a back-up option. Alternative arrangements based on gas cycles are also studied together with Na-gas heat exchanger design. This paper presents a status of the ASTRID pre-conceptual design. The most promising options are highlighted as well as less risky and back-up options. (authors)« less

Quelques Facteurs Sociaux Agissant sur la Formation Permanente et l'Education Informelle en Algerie (Social Factors Acting upon Lifelong Learning and Informal Education in Algeria).

ERIC Educational Resources Information Center

Haddab, Mustapha

1994-01-01

Analyzes conditions that have led to an increase in private and collective educational initiatives in Algeria, highlighting political and socioeconomic changes since 1988. Indicates that after a long period of a public education monopoly, social factors have led to the development of alternative educational opportunities that are more responsive…

Nitrogen release from rock and soil under simulated field conditions

USGS Publications Warehouse

Holloway, J.M.; Dahlgren, R.A.; Casey, W.H.

2001-01-01

A laboratory study was performed to simulate field weathering and nitrogen release from bedrock in a setting where geologic nitrogen has been suspected to be a large local source of nitrate. Two rock types containing nitrogen, slate (1370 mg N kg-1) and greenstone (480 mg N kg-1), were used along with saprolite and BC horizon sand from soils derived from these rock types. The fresh rock and weathered material were used in batch reactors that were leached every 30 days over 6 months to simulate a single wet season. Nitrogen was released from rock and soil materials at rates between 10-20 and 10-19 mo1 N cm-2 s-1. Results from the laboratory dissolution experiments were compared to in situ soil solutions and available mineral nitrogen pools from the BC horizon of both soils. Concentrations of mineral nitrogen (NO3- + NH4+) in soil solutions reached the highest levels at the beginning of the rainy season and progressively decreased with increased leaching. This seasonal pattern was repeated for the available mineral nitrogen pool that was extracted using a KCl solution. Estimates based on these laboratory release rates bracket stream water NO3-N fluxes and changes in the available mineral nitrogen pool over the active leaching period. These results confirm that geologic nitrogen, when present, may be a large and reactive pool that may contribute as a non-point source of nitrate contamination to surface and ground waters. ?? 2001 Elsevier Science B.V. All rights reserved.

Multiple discharge cylindrical pump collector

DOEpatents

Dunn, Charlton; Bremner, Robert J.; Meng, Sen Y.

1989-01-01

A space-saving discharge collector 40 for the rotary pump 28 of a pool-type nuclear reactor 10. An annular collector 50 is located radially outboard for an impeller 44. The annular collector 50 as a closed outer periphery 52 for collecting the fluid from the impeller 44 and producing a uniform circumferential flow of the fluid. Turning means comprising a plurality of individual passageways 54 are located in an axial position relative to the annular collector 50 for receiving the fluid from the annular collector 50 and turning it into a substantially axial direction.

Radiation shielding evaluation of the BNCT treatment room at THOR: a TORT-coupled MCNP Monte Carlo simulation study.

PubMed

Chen, A Y; Liu, Y-W H; Sheu, R J

2008-01-01

This study investigates the radiation shielding design of the treatment room for boron neutron capture therapy at Tsing Hua Open-pool Reactor using "TORT-coupled MCNP" method. With this method, the computational efficiency is improved significantly by two to three orders of magnitude compared to the analog Monte Carlo MCNP calculation. This makes the calculation feasible using a single CPU in less than 1 day. Further optimization of the photon weight windows leads to additional 50-75% improvement in the overall computational efficiency.

Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

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

Billone, M. C.; Burtseva, T. A.

2016-08-30

The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

Support de la famille dans l'education: Quelques aspects de la realite grecque (Support of the Family for Education: Aspects of the Situation in Greece).

ERIC Educational Resources Information Center

Varnava-Skoura, Gella

1992-01-01

Describes extended family structure in Greece and offers a profile of the family backgrounds of university students. Finds that the cultural capital and sociolinguistic codes of families are not determining factors for university entry in Greece. University students come from clerical and mixed families, who are willing to make necessary financial…

Simulation 󈨔 Symposium.

DTIC Science & Technology

1980-11-21

defensive , and both the question and the answer seemed to generate supporting reactions from the audience. Discrete Event Simulation The session on...R. Toscano / A. Maceri / F. Maceri (Italy) Analyse numerique de quelques problemes de contact en theorie des membranes 3:40 - 4:00 p.m. COFFEE BREAK...Switzerland Stockage de chaleur faible profondeur : Simulation par elements finis 3:40 - 4:00 p.m. A. Rizk Abu El-Wafa / M. Tawfik / M.S. Mansour (Egypt) Digital

Quelques applications de la decomposition de l'entropie en psychologie et pedagogie (Some Applications of the Resolution of Entropy in Psychology and Pedagogy)

ERIC Educational Resources Information Center

Gillet, Louis

1971-01-01

Psychological and educational measurement is carried out according to the type of model used and data collected. The H entropy which shows the dispersion of the data can be divided into intragroup and intergroup entropy. Choice of colors, sociometrical choice, and the communications are three situations where this resolution can be applied. (MF)

Le grand séisme de Huaxian (1556) : quelques documents chinois

NASA Astrophysics Data System (ADS)

Poirier, Jean-Paul

2017-03-01

The strong earthquake that struck Shaanxi, Shanxi and several other Chinese provinces in 1556 is generally considered as the deadliest of all earthquakes. It is said that the Chinese annals reported 830,000 casualties. We give here a translation into French of the relevant passage of the annals, as well as of a testimony of a survivor Qin Keda, and of a text engraved on a stela.

Quelques aspects du folklore de la region Roannaise autour de 1950 (Some Aspects of Folklore of the Roanne Region about 1950).

ERIC Educational Resources Information Center

Long, Jacqueline

1971-01-01

This article examines several aspects of folklore characteristic of the region of Roanne, France, during the 1950's. The town of Roanne, located between Clermont Ferrand and Lyon on the Loire River, is described in terms of its festive activities during serveral key holidays. The erosion of various customs and traditions, an inevitable result of…

Radiological Impact of Tritium from Gaseous Effluent Releases at Cook Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Young, Joshua Allan

The purpose of this study was to investigate the washout of tritiated water by snow and rain from gaseous effluent releases at Donald C. Cook Nuclear Power Plant. Primary concepts studied were determination of washout coefficients for rainfall and snowfall; correlations between rainfall and snow fall tritium concentrations with tritium concentrations in the spent fuel pool, reactor cooling systems, and tritium release rates; and calculations of received doses from the process of recapture. The dose calculations are under the assumption of a maximally exposed individual to get the most conservative estimate of the effect that washout of tritiated water has on individuals around the plant site. This study is in addition to previous work that has been conducted at Cook Nuclear Power Plant for several years. The calculated washout coefficients were typically within the range of 1x10-7s -1 to 1x10-5s-1. A strong correlation between tritium concentration within the spent fuel pool and the tritium release rates was determined.

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

Boyack, B.E.

The PIUS reactor utilizes simplified, inherent, passive, or other innovative means to accomplish safety functions. Accordingly, the PIUS reactor is subject to the requirements of 10CFR52.47(b)(2)(i)(A). This regulation requires that the applicant adequately demonstrate the performance of each safety feature, interdependent effects among the safety features, and a sufficient data base on the safety features of the design to assess the analytical tools used for safety analysis. Los Alamos has assessed the quality and completeness of the existing and planned data bases used by Asea Brown Boveri to validate its safety analysis codes and other relevant data bases. Only amore » limited data base of separate effect and integral tests exist at present. This data base is not adequate to fulfill the requirements of 10CFR52.47(b)(2)(i)(A). Asea Brown Boveri has stated that it plans to conduct more separate effect and integral test programs. If appropriately designed and conducted, these test programs have the potential to satisfy most of the data base requirements of 10CFR52.47(b)(2)(i)(A) and remedy most of the deficiencies of the currently existing combined data base. However, the most important physical processes in PIUS are related to reactor shutdown because the PIUS reactor does not contain rodded shutdown and control systems. For safety-related reactor shutdown, PIUS relies on negative reactivity insertions from the moderator temperature coefficient and from boron entering the core from the reactor pool. Asea Brown Boveri has neither developed a direct experimental data base for these important processes nor provided a rationale for indirect testing of these key PIUS processes. This is assessed as a significant shortcoming. In preparing the conclusions of this report, test documentation and results have been reviewed for only one integral test program, the small-scale integral tests conducted in the ATLE facility.« less

AP1000{sup R} severe accident features and post-Fukushima considerations

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

Scobel, J. H.; Schulz, T. L.; Williams, M. G.

2012-07-01

The AP1000{sup R} passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at Fukushima without compromising core and containment integrity. The AP1000 plant shuts down the reactor, cools the core, containment and spent fuel pool for more than 3 days using passive systems that do not require AC or DC power or operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. To provide defense-in-depth for design extension conditions, themore » AP1000 plant has engineered features that mitigate the effects of core damage. Engineered features retain damaged core debris within the reactor vessel as a key feature. Other aspects of the design protect containment integrity during severe accidents, including unique features of the AP1000 design relative to passive containment cooling with water and air, and hydrogen management. (authors)« less

Optimization of semi-continuous anaerobic digestion of sugarcane straw co-digested with filter cake: Effects of macronutrients supplementation on conversion kinetics.

PubMed

Janke, Leandro; Weinrich, Sören; Leite, Athaydes F; Schüch, Andrea; Nikolausz, Marcell; Nelles, Michael; Stinner, Walter

2017-12-01

Anaerobic digestion of sugarcane straw co-digested with sugarcane filter cake was investigated with a special focus on macronutrients supplementation for an optimized conversion process. Experimental data from batch tests and a semi-continuous experiment operated in different supplementation phases were used for modeling the conversion kinetics based on continuous stirred-tank reactors. The semi-continuous experiment showed an overall decrease in the performance along the inoculum washout from the reactors. By supplementing nitrogen alone or in combination to phosphorus and sulfur the specific methane production significantly increased (P<0.05) by 17% and 44%, respectively. Although the two-pool one-step model has fitted well to the batch experimental data (R 2 >0.99), the use of the depicted kinetics did not provide a good estimation for process simulation of the semi-continuous process (in any supplementation phase), possibly due to the different feeding modes and inoculum source, activity and adaptation. Copyright © 2017 Elsevier Ltd. All rights reserved.

A broad-group cross-section library based on ENDF/B-VII.0 for fast neutron dosimetry Applications

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

Alpan, F.A.

2011-07-01

A new ENDF/B-VII.0-based coupled 44-neutron, 20-gamma-ray-group cross-section library was developed to investigate the latest evaluated nuclear data file (ENDF) ,in comparison to ENDF/B-VI.3 used in BUGLE-96, as well as to generate an objective-specific library. The objectives selected for this work consisted of dosimetry calculations for in-vessel and ex-vessel reactor locations, iron atom displacement calculations for reactor internals and pressure vessel, and {sup 58}Ni(n,{gamma}) calculation that is important for gas generation in the baffle plate. The new library was generated based on the contribution and point-wise cross-section-driven (CPXSD) methodology and was applied to one of the most widely used benchmarks, themore » Oak Ridge National Laboratory Pool Critical Assembly benchmark problem. In addition to the new library, BUGLE-96 and an ENDF/B-VII.0-based coupled 47-neutron, 20-gamma-ray-group cross-section library was generated and used with both SNLRML and IRDF dosimetry cross sections to compute reaction rates. All reaction rates computed by the multigroup libraries are within {+-} 20 % of measurement data and meet the U. S. Nuclear Regulatory Commission acceptance criterion for reactor vessel neutron exposure evaluations specified in Regulatory Guide 1.190. (authors)« less

Results of a Demonstration Assessment of Passive System Reliability Utilizing the Reliability Method for Passive Systems (RMPS)

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

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia

2015-04-26

Advanced small modular reactor designs include many advantageous design features such as passively driven safety systems that are arguably more reliable and cost effective relative to conventional active systems. Despite their attractiveness, a reliability assessment of passive systems can be difficult using conventional reliability methods due to the nature of passive systems. Simple deviations in boundary conditions can induce functional failures in a passive system, and intermediate or unexpected operating modes can also occur. As part of an ongoing project, Argonne National Laboratory is investigating various methodologies to address passive system reliability. The Reliability Method for Passive Systems (RMPS), amore » systematic approach for examining reliability, is one technique chosen for this analysis. This methodology is combined with the Risk-Informed Safety Margin Characterization (RISMC) approach to assess the reliability of a passive system and the impact of its associated uncertainties. For this demonstration problem, an integrated plant model of an advanced small modular pool-type sodium fast reactor with a passive reactor cavity cooling system is subjected to a station blackout using RELAP5-3D. This paper discusses important aspects of the reliability assessment, including deployment of the methodology, the uncertainty identification and quantification process, and identification of key risk metrics.« less

A Passive System Reliability Analysis for a Station Blackout

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

Brunett, Acacia; Bucknor, Matthew; Grabaskas, David

2015-05-03

The latest iterations of advanced reactor designs have included increased reliance on passive safety systems to maintain plant integrity during unplanned sequences. While these systems are advantageous in reducing the reliance on human intervention and availability of power, the phenomenological foundations on which these systems are built require a novel approach to a reliability assessment. Passive systems possess the unique ability to fail functionally without failing physically, a result of their explicit dependency on existing boundary conditions that drive their operating mode and capacity. Argonne National Laboratory is performing ongoing analyses that demonstrate various methodologies for the characterization of passivemore » system reliability within a probabilistic framework. Two reliability analysis techniques are utilized in this work. The first approach, the Reliability Method for Passive Systems, provides a mechanistic technique employing deterministic models and conventional static event trees. The second approach, a simulation-based technique, utilizes discrete dynamic event trees to treat time- dependent phenomena during scenario evolution. For this demonstration analysis, both reliability assessment techniques are used to analyze an extended station blackout in a pool-type sodium fast reactor (SFR) coupled with a reactor cavity cooling system (RCCS). This work demonstrates the entire process of a passive system reliability analysis, including identification of important parameters and failure metrics, treatment of uncertainties and analysis of results.« less

Experimental study of terrestrial plant litter interaction with aqueous solutions

NASA Astrophysics Data System (ADS)

Fraysse, F.; Pokrovsky, O. S.; Meunier, J.-D.

2010-01-01

Quantification of silicon and calcium recycling by plants is hampered by the lack of physico-chemical data on reactivity of plant litter in soil environments. We applied a laboratory experimental approach for determining the silica and calcium release rates from litter of typical temperate and boreal plants: pine ( Pinus laricio), birch ( Betula pubescens), larch ( Larix gmelinii), elm ( Ulmus laevis Pall.), tree fern ( Dicksonia squarrosa), and horsetail (Equisetum arvense) in 0.01 M NaCl solutions, pH of 2-10 and temperature equals to 5, 25 and 40 °C. Open system, mixed-flow reactors equipped with dialysis compartment and batch reactors were used. Comparative measurements were performed on intact larch needles and samples grounded during different time, sterilized or not and with addition or not of sodium azide in order to account for the effect of surface to mass ratio and possible microbiological activity on the litter dissolution rates. Litter degradation results suggest that the silica release rate is independent on dissolved organic carbon release (cell breakdown) which implies the presence of phytoliths in a pure "inorganic" pool not complexed with organic matter. Calcium and DOC are released at the very first stage of litter dissolution while Si concentration increases gradually suggesting the presence of Ca and Si in two different pools. The dry-weight normalized dissolution rate at circum-neutral pH range (approx. 1-10 μmol/g DW/day) is 2 orders of magnitude higher than the rates of Si release from common soil minerals (kaolinite, smectite, illite). Minimal Ca release rates evaluated from batch and mixed-flow reactors are comparable with those of most reactive soil minerals such as calcite and apatite, and several orders of magnitude higher than the dissolution rates of major rock-forming silicates (feldspars, pyroxenes). The activation energy for Si liberation from plant litter is approx. 50 kJ/mol which is comparable with that of surface-controlled mineral dissolutions. It is shown that the Si release rate from the above-ground forest biomass is capable of producing the Si concentrations observed in soil solutions of surficial horizons and contribute significantly to the Si flux from the soil to the river.

Helium refrigerator maintenance and reliability at the OPAL cold neutron source

NASA Astrophysics Data System (ADS)

Thiering, Russell; Taylor, David; Lu, Weijian

2012-06-01

Australia's first Cold Neutron Source (CNS) is a major asset to its nuclear research program. The CNS, and associated helium refrigerator, was commissioned in 2006 and is operated at the Open Pool Light Water nuclear Reactor (OPAL). The OPAL CNS operates a 20K, 5 kW Brayton cycle helium refrigerator. In this paper relevant experiences from helium refrigerator operation, maintenance and repair are presented along with the lessons learnt from a series of technical investigations. Turbine failure, due to volatile organic species, is discussed along with the related compressor oil degradation and oil separation efficiency.

Method and apparatus for controlling the flow rate of mercury in a flow system

DOEpatents

Grossman, Mark W.; Speer, Richard

1991-01-01

A method for increasing the mercury flow rate to a photochemical mercury enrichment utilizing an entrainment system comprises the steps of passing a carrier gas over a pool of mercury maintained at a first temperature T1, wherein the carrier gas entrains mercury vapor; passing said mercury vapor entrained carrier gas to a second temperature zone T2 having temperature less than T1 to condense said entrained mercury vapor, thereby producing a saturated Hg condition in the carrier gas; and passing said saturated Hg carrier gas to said photochemical enrichment reactor.

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

DOE PAGES

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel--coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

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

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain cool-able. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

The Role of Information and Research in Educational Decision-Making: Some Questions. Le Role De L'Information Et De La Recherche Dans La Prise De Decisions En Matiere D'Education: Quelques Questions.

ERIC Educational Resources Information Center

United Nations Educational, Scientific, and Cultural Organization, Paris (France).

This paper, one of a series of Unesco technical information reports, looks at the educational decision makers in developing nations and examines their access to and use of information and research results. Written in English and in French, the paper consists of five parts. Part one discusses problems encountered by educational policy-makers and…

"Bonjour History": Materials and Ideas for Teaching Louisiana's Cajun History = "Bonjour l'Histoire": Quelques Idees, Deux ou Trois Activites, et Plusieurs Materiaux pour l'Enseignement de l'Histoire des Cadiens en Louisiane.

ERIC Educational Resources Information Center

Mounier, Brenda

The goals of this teacher's guidebook and videotape are designed to incorporate Acadian (Cajun) history into the 4th grade social studies curriculum and the 4th and 5th grade Louisiana 30-minute daily French programs and French immersion programs. Another goal is to create an awareness, appreciation, and understanding of Acadian history in…

Flight Control Design - Best Practices

DTIC Science & Technology

2000-12-01

n’était pas universellement disponible à l’époque. La première partie du rapport donne quelques exemples de problèmes de commandes de vol. Ils...pitch axis. We can infer a lesson learned in the form of design guidance for control allocation or priority. Rigorous analysis is required to define...flight excitation and data gathering manoeuvres are safe and are sufficient to produce the required information. BP9.5 Time must be allocated in the

Coccomyxa actinabiotis sp. nov. (Trebouxiophyceae, Chlorophyta), a new green microalga living in the spent fuel cooling pool of a nuclear reactor.

PubMed

Rivasseau, Corinne; Farhi, Emmanuel; Compagnon, Estelle; de Gouvion Saint Cyr, Diane; van Lis, Robert; Falconet, Denis; Kuntz, Marcel; Atteia, Ariane; Couté, Alain

2016-10-01

Life can thrive in extreme environments where inhospitable conditions prevail. Organisms which resist, for example, acidity, pressure, low or high temperature, have been found in harsh environments. Most of them are bacteria and archaea. The bacterium Deinococcus radiodurans is considered to be a champion among all living organisms, surviving extreme ionizing radiation levels. We have discovered a new extremophile eukaryotic organism that possesses a resistance to ionizing radiations similar to that of D. radiodurans. This microorganism, an autotrophic freshwater green microalga, lives in a peculiar environment, namely the cooling pool of a nuclear reactor containing spent nuclear fuels, where it is continuously submitted to nutritive, metallic, and radiative stress. We investigated its morphology and its ultrastructure by light, fluorescence and electron microscopy as well as its biochemical properties. Its resistance to UV and gamma radiation was assessed. When submitted to different dose rates of the order of some tens of mGy · h -1 to several thousands of Gy · h -1 , the microalga revealed to be able to survive intense gamma-rays irradiation, up to 2,000 times the dose lethal to human. The nuclear genome region spanning the genes for small subunit ribosomal RNA-Internal Transcribed Spacer (ITS) 1-5.8S rRNA-ITS2-28S rRNA (beginning) was sequenced (4,065 bp). The phylogenetic position of the microalga was inferred from the 18S rRNA gene. All the revealed characteristics make the alga a new species of the genus Coccomyxa in the class Trebouxiophyceae, which we name Coccomyxa actinabiotis sp. nov. © 2016 Phycological Society of America.

Successful scaling-up of self-sustained pyrolysis of oil palm biomass under pool-type reactor.

PubMed

Idris, Juferi; Shirai, Yoshihito; Andou, Yoshito; Mohd Ali, Ahmad Amiruddin; Othman, Mohd Ridzuan; Ibrahim, Izzudin; Yamamoto, Akio; Yasuda, Nobuhiko; Hassan, Mohd Ali

2016-02-01

An appropriate technology for waste utilisation, especially for a large amount of abundant pressed-shredded oil palm empty fruit bunch (OFEFB), is important for the oil palm industry. Self-sustained pyrolysis, whereby oil palm biomass was combusted by itself to provide the heat for pyrolysis without an electrical heater, is more preferable owing to its simplicity, ease of operation and low energy requirement. In this study, biochar production under self-sustained pyrolysis of oil palm biomass in the form of oil palm empty fruit bunch was tested in a 3-t large-scale pool-type reactor. During the pyrolysis process, the biomass was loaded layer by layer when the smoke appeared on the top, to minimise the entrance of oxygen. This method had significantly increased the yield of biochar. In our previous report, we have tested on a 30-kg pilot-scale capacity under self-sustained pyrolysis and found that the higher heating value (HHV) obtained was 22.6-24.7 MJ kg(-1) with a 23.5%-25.0% yield. In this scaled-up study, a 3-t large-scale procedure produced HHV of 22.0-24.3 MJ kg(-1) with a 30%-34% yield based on a wet-weight basis. The maximum self-sustained pyrolysis temperature for the large-scale procedure can reach between 600 °C and 700 °C. We concluded that large-scale biochar production under self-sustained pyrolysis was successfully conducted owing to the comparable biochar produced, compared with medium-scale and other studies with an electrical heating element, making it an appropriate technology for waste utilisation, particularly for the oil palm industry. © The Author(s) 2015.

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

Pomorski, Michal; Mer-Calfati, Christine; Foulon, Francois

Diamond exhibits a combination of properties which makes it attractive for neutron detection in hostile conditions. In the particular case of detection in a nuclear reactor, it is resilient to radiation, exhibits a natural low sensitivity to gamma rays, and its small size (as compared with that of gas ionisation chambers) enables fluency monitoring with a high position resolution. We report here on the use of synthetic CVD diamond as a solid state micro-fission chamber with U-235 converting material for in-core thermal neutron monitoring. Two types of thin diamond detectors were developed for this application. The first type of detectormore » is fabricated using thin diamond membrane obtained by etching low-cost commercially available single crystal CVD intrinsic diamond, so called 'optical grade' material. Starting from a few hundred of micrometre thick samples, the sample is sliced with a laser and then plasma etched down to a few tenths of micrometre. Here we report the result obtained with a 17 μm thick device. The detection surface of this detector is equal to 1 mm{sup 2}. Detectors with surfaces up to 1 cm{sup 2} can be fabricated with this technique. The second type of detector is fabricated by growing successively two thin films of diamond, by the microwave enhanced chemical vapour deposition technique, on HPHT single crystal diamond. A first, a film of boron doped (p+) single crystal diamond, a few microns thick, is deposited. Then a second film of intrinsic diamond with a thickness of a few tens of microns is deposited. This results in a P doped, Intrinsic, Metal structure (PIM) structure in which the intrinsic volume id the active part of the detector. Here we report the results obtained with a 20 μm thick intrinsic whose detection surface is equal to 0.5 mm{sup 2}, with the possibility to enlarge the surface of the detector up to 1 cm{sup 2}. These two types of detector were tested at the VR-1 research reactor at the Czech Technical University in Prague. The Training Reactor VR-1 is a pool type (light water) reactor based on UO{sub 2} low enriched uranium. It has a nominal power of 1 kW, and can be operated for a short period up to 5 kW. The arrangement of the reactor pool reactor facilitates access to the core, setting and removal of various experimental samples and detectors, and safe and easy handling of fuel assemblies. The reactor is equipped with two horizontal channels (radial and tangential) and 10 vertical channels, of varying diameters, which can be loaded into various core positions, and one pneumatic transfer system. It is also equipped with several specifically designed educational instrumentation systems that can be used to supply complementary measurements and characterization around the reactor. The reactor is operated by the Department of Nuclear Reactors of the Faculty of Nuclear Sciences and Physical Engineering of the Czech Technical University in Prague. The two detectors were placed in-core through one of the vertical insertion channel. They were coupled to remote placed (5 m BNC cable) classical nuclear charge sensitive electronics. Detection properties of both sensors, including: pulse height spectra of U-235 fission fragments (response linearity with neutron flux, count rate, gamma background, were evaluated varying the power of the reactor from 0.005 W to 500 W. The evolution of the counting rate of the thinned optical grade detector as a function of counting rate of a gas ionization chamber used currently for reactor monitoring shows the very good linearity of the detector over the 5 decades. Similar results were obtained with the PIM detector. Additionally fast transient current signals of the detectors were recorded on a digital storage oscilloscope (DSO) using broad-band amplifier and with a simple bias-T, showing potential use of such sensors for neutron counting with no need of an amplification stage, since non-amplified signals from fission fragments exceeded 4 mV in amplitude. Therefore, one can think of simple neutron counting system by feeding diamond detectors signals directly to the low threshold discriminators. The results obtained on the VR1 will be described and discussed in detail in the paper and associated presentation. The results demonstrate that diamond micro-fission chambers can be used for in-core neutron monitoring, where robust, simple and compact devices are required.« less

PRELIMINARY EVALUATION OF FeCrAl CLADDING AND U-Si FUEL FOR ACCIDENT TOLERANT FUEL CONCEPTS

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

Hales, J. D.; Gamble, K. A.

2015-09-01

Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normalmore » operations and operational transients. This paper presents early work in developing thermal and mechanical models for two materials that may have promise: U-Si for fuel, and FeCrAl for cladding. These materials would not necessarily be used together in the same fuel system, but individually have promising characteristics. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with Zr-4/UO2 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as ductility and thermal conductivity in FeCrAl and U-Si in order to provide guidance on future experiments for these materials.« less

Investigation of saliva of patients with periodontal disease using NAA

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

Zamboni, C. B.; Metairon, S.; Medeiros, I. M. M. A.

In this study the non-stimulated whole saliva of 26 healthy subjects (mean age 33.9 {+-} 11.0 years, range: 26 to 49 years) and 11 patients with periodontal disease (mean age 41.7 {+-} 11.5 years; range 29 to 55 years) was investigated using Neutron Activation Analysis (NAA) technique. The samples were obtained from donors at Sao Paulo city (Brazil). The analyses were performed in the nuclear reactor IEA-R1 (3.5-4.5MW, pool type) at IPEN/CNEN-SP (Brazil). Considerable changes in Ca and S saliva's level were identified in patients with periodontal disease suggesting they can be used as monitors of periodontal diseases.

The shutdown reactor: Optimizing spent fuel storage cost

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

Pennington, C.W.

1995-12-31

Several studies have indicated that the most prudent way to store fuel at a shutdown reactor site safely and economically is through the use of a dry storage facility licensed under 10CFR72. While such storage is certainly safe, is it true that the dry ISFSI represents the safest and most economical approach for the utility? While no one is really able to answer that question definitely, as yet, Holtec has studied this issue for some time and believes that both an economic and safety case can be made for an optimization strategy that calls for the use of both wetmore » and dry ISFSI storage of spent fuel at some plants. For the sake of brevity, this paper summarizes some of Holtec`s findings with respect to the economics of maintaining some fuel in wet storage at a shutdown reactor. The safety issue, or more importantly the perception of safety of spent fuel in wet storage, still varies too much with the eye of the beholder, and until a more rigorous presentation of safety analyses can be made in a regulatory setting, it is not practically useful to argue about how many angels can sit on the head of a safety-related pin. Holtec is prepared to present such analyses, but this does not appear to be the proper venue. Thus, this paper simply looks at certain economic elements of a wet ISFSI at a shutdown reactor to make a prima facie case that wet storage has some attractiveness at a shutdown reactor and should not be rejected out of hand. Indeed, an optimization study at certain plants may well show the economic vitality of keeping some fuel in the pool and converting the NRC licensing coverage from 10CFR50 to 10CFR72. If the economics look attractive, then the safety issue may be confronted with a compelling interest.« less

Safety analysis

NASA Technical Reports Server (NTRS)

Knight, John C.

1995-01-01

We are engaged in a research program in safety-critical computing that is based on two case studies. We use these case studies to provide application-specific details of the various research issues, and as targets for evaluation of research ideas. The first case study is the Magnetic Stereotaxis System (MSS), an investigational device for performing human neurosurgery being developed in a joint effort between the Department of Physics at the University of Virginia and the Department of Neurosurgery at the University of Iowa. The system operates by manipulating a small permanent magnet (known as a 'seed') within the brain using an externally applied magnetic field. By varying the magnitude and gradient of the external magnetic field, the seed can be moved along a non-linear path and positioned at a site requiring therapy, e.g., a tumor. The magnetic field required for movement through brain tissue is extremely high, and is generated by a set of six superconducting magnets located in a housing surrounding the patient's head. The system uses two X-ray cameras positioned at right angles to detect in real time the locations of the seed and of X-ray opaque markers affixed to the patient's skull. the X-ray images are used to locate the objects of interest in a canonical frame of reference. the second case study is the University of Virginia Research Nuclear Reactor (UVAR). It is a 2 MW thermal, concrete-walled pool reactor. The system operates using 20 to 25 plate-type fuel assemblies placed on a rectangular grid plate. There are three scramable safety rods, and one non-scramable regulating rod that can be put in automatic mode. It was originally constructed in 1959 as a 1 MW system, and it was upgraded to 2 MW in 1973. Though only a research reactor rather than a power reactor, the issues raised are significant and can be related to the problems faced by full-scale reactor systems.

Enhancement of NRC station blackout requirements for nuclear power plants

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

McConnell, M. W.

2012-07-01

The U.S. Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) in response to Commission direction to conduct a systematic and methodical review of NRC processes and regulations to determine whether the agency should make additional improvements to its regulatory system and to make recommendations to the Commission for its policy direction, in light of the accident at the Fukushima Dai-ichi Nuclear Power Plant. The NTTF's review resulted in a set of recommendations that took a balanced approach to defense-in-depth as applied to low-likelihood, high-consequence events such as prolonged station blackout (SBO) resulting from severe natural phenomena. Part 50,more » Section 63, of Title 10 of the Code of Federal Regulations (CFR), 'Loss of All Alternating Current Power,' currently requires that each nuclear power plant must be able to cool the reactor core and maintain containment integrity for a specified duration of an SBO. The SBO duration and mitigation strategy for each nuclear power plant is site specific and is based on the robustness of the local transmission system and the transmission system operator's capability to restore offsite power to the nuclear power plant. With regard to SBO, the NTTF recommended that the NRC strengthen SBO mitigation capability at all operating and new reactors for design-basis and beyond-design-basis external events. The NTTF also recommended strengthening emergency preparedness for prolonged SBO and multi-unit events. These recommendations, taken together, are intended to clarify and strengthen US nuclear reactor safety regarding protection against and mitigation of the consequences of natural disasters and emergency preparedness during SBO. The focus of this paper is on the existing SBO requirements and NRC initiatives to strengthen SBO capability at all operating and new reactors to address prolonged SBO stemming from design-basis and beyond-design-basis external events. The NRC initiatives are intended to enhance core and spent fuel pool cooling, reactor coolant system integrity, and containment integrity. (authors)« less

Design requirements for innovative homogeneous reactor, lesson learned from Fukushima accident

NASA Astrophysics Data System (ADS)

Arbie, Bakri; Pinem, Suryan; Sembiring, Tagor; Subki, Iyos

2012-06-01

The Fukushima disaster is the largest nuclear accident since the 1986 Chernobyl disaster, but it is more complex as multiple reactors and spent fuel pools are involved. The severity of the nuclear accident is rated 7 in the International Nuclear Events Scale. Expert said that "Fukushima is the biggest industrial catastrophe in the history of mankind". According to Mitsuru Obe, in The Wall Street Journal, May 16th of 2011, TEPCO estimates the nuclear fuel was exposed to the air less than five hours after the earthquake struck. Fuel rods melted away rapidly as the temperatures inside the core reached 2800 C within six hours. In less than 16 hours, the reactor core melted and dropped to the bottom of the pressure vessel. The information should be evaluated in detail. In Germany several nuclear power plant were shutdown, Italy postponed it's nuclear power program and China reviewed their nuclear power program. Different news come from Britain, in October 11, 2011, the Safety Committee said all clear for nuclear power in Britain, because there are no risk of strong earthquake and tsunami in the region. Due to this severe fact, many nuclear scientists and engineer from all over the world are looking for a new approach, such as homogeneous reactor which was developed in Oak Ridge National Laboratory in 1960-ies, during Dr. Alvin Weinberg tenure as the Director of ORNL. The paper will describe the design requirement that will be used as the basis for innovative homogeneous reactor. Innovative Homogeneous Reactor is expected to reduce core melt by two decades (4), since the fuel is intermix homogeneously with coolant and secondly we eliminate the used fuel rod which need to be cooled for a long period of time. In order to be successful for its implementation of the innovative system, testing and validation, three phases of development will be introduced. The first phase is Low Level Goals is really the proof of concept;the Medium Level Goal is Technical Goalsand the High Level Goals which is Business Goals.

Future U.S. supply of Mo-99 production through fission based LEU/LEU technology.

PubMed

Welsh, James; Bigles, Carmen I; Valderrabano, Alejandro

Coquí RadioPharmaceuticals Corp. (Coquí) has the goal of establishing a medical isotope production facility for securing a continuous domestic supply of the radioisotope molybdenum-99 for U.S. citizens. Coquí will use an LEU/LEU proven and implemented open pool, light-water, 10 MW, reactor design. The facility is being designed with twin reactors for reliability an on-site hot lab chemical processing and a waste conditioning area and a possible generator producing radio-chemistry lab. Coquí identified a 25 acre site adjacent to an existing industrial park in northern central Florida. This land was gifted and transferred to Coquí by the University of Florida Foundation. We are in the process of developing licensing documents related to the facility. The construction permit application for submission to the U.S. Nuclear Regulatory Commission is currently being prepared. Submission is scheduled for mid to late 2015. Community reaction to the proposed development has been positive. We expect to create 220 permanent jobs and we have an anticipated to be operational by 2020.

The radioprotective properties of fungal melanin are a function of its chemical composition, stable radical presence and spatial arrangement.

PubMed

Dadachova, Ekaterina; Bryan, Ruth A; Howell, Robertha C; Schweitzer, Andrew D; Aisen, Philip; Nosanchuk, Joshua D; Casadevall, Arturo

2008-04-01

Melanized microorganisms are often found in environments with very high background radiation levels such as in nuclear reactor cooling pools and the destroyed reactor in Chernobyl. These findings and the laboratory observations of the resistance of melanized fungi to ionizing radiation suggest a role for this pigment in radioprotection. We hypothesized that the radioprotective properties of melanin in microorganisms result from a combination of physical shielding and quenching of cytotoxic free radicals. We have investigated the radioprotective properties of melanin by subjecting the human pathogenic fungi Cryptococcus neoformans and Histoplasma capsulatum in their melanized and non-melanized forms to sublethal and lethal doses of radiation of up to 8 kGy. The contribution of chemical composition, free radical presence, spatial arrangement, and Compton scattering to the radioprotective properties of melanin was investigated by high-performance liquid chromatography, electron spin resonance, transmission electron microscopy, and autoradiographic techniques. Melanin protected fungi against ionizing radiation and its radioprotective properties were a function of its chemical composition, free radical quenching, and spherical spatial arrangement.

The Upgrade of the Neutron Induced Positron Source NEPOMUC

NASA Astrophysics Data System (ADS)

Hugenschmidt, C.; Ceeh, H.; Gigl, T.; Lippert, F.; Piochacz, C.; Pikart, P.; Reiner, M.; Weber, J.; Zimnik, S.

2013-06-01

In summer 2012, the new NEutron induced POsitron Source MUniCh (NEPOMUC) was installed and put into operation at the research reactor FRM II. At NEPOMUC upgrade 80% 113Cd enriched Cd is used as neutron-gamma converter in order to ensure an operation time of 25 years. A structure of Pt foils inside the beam tube generates positrons by pair production. Moderated positrons leaving the Pt front foil are electrically extracted and magnetically guided to the outside of the reactor pool. The whole design, including Pt-foils, the electric lenses and the magnetic fields, has been improved in order to enhance both the intensity and the brightness of the positron beam. After adjusting the potentials and the magnetic guide and compensation fields an intensity of about 3·109 moderated positrons per second is expected. During the first start-up, the measured temperatures of about 90°C ensure a reliable operation of the positron source. Within this contribution the features and the status of NEPOMUC upgrade are elucidated. In addition, an overview of recent positron beam experiments and current developments at the spectrometers is given.

A multi-physics analysis for the actuation of the SSS in opal reactor

NASA Astrophysics Data System (ADS)

Ferraro, Diego; Alberto, Patricio; Villarino, Eduardo; Doval, Alicia

2018-05-01

OPAL is a 20 MWth multi-purpose open-pool type Research Reactor located at Lucas Heights, Australia. It was designed, built and commissioned by INVAP between 2000 and 2006 and it has been operated by the Australia Nuclear Science and Technology Organization (ANSTO) showing a very good overall performance. On November 2016, OPAL reached 10 years of continuous operation, becoming one of the most reliable and available in its kind worldwide, with an unbeaten record of being fully operational 307 days a year. One of the enhanced safety features present in this state-of-art reactor is the availability of an independent, diverse and redundant Second Shutdown System (SSS), which consists in the drainage of the heavy water reflector contained in the Reflector Vessel. As far as high quality experimental data is available from reactor commissioning and operation stages and even from early component design validation stages, several models both regarding neutronic and thermo-hydraulic approaches have been developed during recent years using advanced calculations tools and the novel capabilities to couple them. These advanced models were developed in order to assess the capability of such codes to simulate and predict complex behaviours and develop highly detail analysis. In this framework, INVAP developed a three-dimensional CFD model that represents the detailed hydraulic behaviour of the Second Shutdown System for an actuation scenario, where the heavy water drainage 3D temporal profiles inside the Reflector Vessel can be obtained. This model was validated, comparing the computational results with experimental measurements performed in a real-size physical model built by INVAP during early OPAL design engineering stages. Furthermore, detailed 3D Serpent Monte Carlo models are also available, which have been already validated with experimental data from reactor commissioning and operating cycles. In the present work the neutronic and thermohydraulic models, available for OPAL reactor, are coupled by means of a shared unstructured mesh geometry definition of relevant zones inside the Reflector Vessel. Several scenarios, both regarding coupled and uncoupled neutronic & thermohydraulic behavior, are presented and analyzed, showing the capabilities to develop and manage advanced modelling that allows to predict multi-physics variables observed when an in-depth performance analysis of a Research Reactor like OPAL is carried out.

Monochromatic neutron beam production at Brazilian nuclear research reactors

NASA Astrophysics Data System (ADS)

Stasiulevicius, Roberto; Rodrigues, Claudio; Parente, Carlos B. R.; Voi, Dante L.; Rogers, John D.

2000-12-01

Monochomatic beams of neutrons are obtained form a nuclear reactor polychromatic beam by the diffraction process, suing a single crystal energy selector. In Brazil, two nuclear research reactors, the swimming pool model IEA-R1 and the Argonaut type IEN-R1 have been used to carry out measurements with this technique. Neutron spectra have been measured using crystal spectrometers installed on the main beam lines of each reactor. The performance of conventional- artificial and natural selected crystals has been verified by the multipurpose neutron diffractometers installed at IEA-R1 and simple crystal spectrometer in operator at IEN- R1. A practical figure of merit formula was introduced to evaluate the performance and relative reflectivity of the selected planes of a single crystal. The total of 16 natural crystals were selected for use in the neutron monochromator, including a total of 24 families of planes. Twelve of these natural crystal types and respective best family of planes were measured directly with the multipurpose neutron diffractometers. The neutron spectrometer installed at IEN- R1 was used to confirm test results of the better specimens. The usually conventional-artificial crystal spacing distance range is limited to 3.4 angstrom. The interplane distance range has now been increased to approximately 10 angstrom by use of naturally occurring crystals. The neutron diffraction technique with conventional and natural crystals for energy selection and filtering can be utilized to obtain monochromatic sub and thermal neutrons with energies in the range of 0.001 to 10 eV. The thermal neutron is considered a good tool or probe for general applications in various fields, such as condensed matter, chemistry, biology, industrial applications and others.

New developments and prospects on COSI, the simulation software for fuel cycle analysis

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

Eschbach, R.; Meyer, M.; Coquelet-Pascal, C.

2013-07-01

COSI, software developed by the Nuclear Energy Direction of the CEA, is a code simulating a pool of nuclear power plants with its associated fuel cycle facilities. This code has been designed to study various short, medium and long term options for the introduction of various types of nuclear reactors and for the use of associated nuclear materials. In the frame of the French Act for waste management, scenario studies are carried out with COSI, to compare different options of evolution of the French reactor fleet and options of partitioning and transmutation of plutonium and minor actinides. Those studies aimmore » in particular at evaluating the sustainability of Sodium cooled Fast Reactors (SFR) deployment and the possibility to transmute minor actinides. The COSI6 version is a completely renewed software released in 2006. COSI6 is now coupled with the last version of CESAR (CESAR5.3 based on JEFF3.1.1 nuclear data) allowing the calculations on irradiated fuel with 200 fission products and 100 heavy nuclides. A new release is planned in 2013, including in particular the coupling with a recommended database of reactors. An exercise of validation of COSI6, carried out on the French PWR historic nuclear fleet, has been performed. During this exercise quantities like cumulative natural uranium consumption, or cumulative depleted uranium, or UOX/MOX spent fuel storage, or stocks of reprocessed uranium, or plutonium content in fresh MOX fuel, or the annual production of high level waste, have been computed by COSI6 and compared to industrial data. The results have allowed us to validate the essential phases of the fuel cycle computation, and reinforces the credibility of the results provided by the code.« less

Westinghouse Small Modular Reactor balance of plant and supporting systems design

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

Memmott, M. J.; Stansbury, C.; Taylor, C.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operationmore » of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)« 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

Kinetic Parameter Measurements in the MINERVE Reactor

NASA Astrophysics Data System (ADS)

Perret, Grégory; Geslot, Benoit; Gruel, Adrien; Blaise, Patrick; Di-Salvo, Jacques; De Izarra, Grégoire; Jammes, Christian; Hursin, Mathieu; Pautz, Andréas

2017-01-01

In the framework of an international collaboration, teams of the PSI and CEA research institutes measure the critical decay constant (α0 = β/A), delayed neutron fraction (β) and generation time (A) of the Minerve reactor using the Feynman-α, Power Spectral Density and Rossi-α neutron noise measurement techniques. These measurements contribute to the experimental database of kinetic parameters used to improve nuclear data files and validate modern methods in Monte Carlo codes. Minerve is a zero-power pool reactor composed of a central experimental test lattice surrounded by a large aluminum buffer and four high-enriched driver regions. Measurements are performed in three slightly subcritical configurations (-2 cents to -30 cents) using two high-efficiency 235U fission chambers in the driver regions. Measurement of α0 and β obtained by the two institutes and with the different techniques are consistent for the configurations envisaged. Slight increases of the β values are observed with the subcriticality level. Best estimate values are obtained with the Cross-Power Spectral Density technique at -2 cents, and are worth: β = 716.9±9.0 pcm, α0 = 79.0±0.6 s-1 and A = 90.7±1.4 μs. The kinetic parameters are predicted with MCNP5-v1.6 and TRIPOLI4.9 and the JEFF-3.1/3.1.1 and ENDF/B-VII.1 nuclear data libraries. The predictions for β and α0 overestimate the experimental results by 3-5% and 10-12%, respectively; that for A underestimate the experimental result by 6-7%. The discrepancies are suspected to come from the driven system nature of Minerve and the location of the detectors in the driver regions, which prevent accounting for the full reactor.

Sidération myocardique au cours d'une intoxication au monoxyde de carbone (CO) chez une femme enceinte

PubMed Central

Coulibaly, Mahamadoun; Berdai, Mohamed Adnane; Labib, Smael; Harandou, Mustapha

2015-01-01

L'intoxication au monoxyde de carbone (CO) est la première cause de décès par intoxication en France. La littérature est ancienne et peu connue. Les signes les plus fréquents de l'intoxication sont la triade: Céphalées; asthénie, faiblesse musculaire surtout des membres inférieurs. Ses conséquences sont potentiellement graves pour le fœtus quand elle survient chez la femme enceinte, il est particulièrement exposé au risque d'hypoxie en raison de la forte affinité de son hémoglobine pour le CO qui traverse aisément le placenta. Les événements cardiovasculaires ne sont pas rares et peuvent être responsable d'une morbi-mortalité assez importante qui peuvent être d'apparition rapide ou secondaire mais régressent habituellement en quelques jours. Des SCA peuvent survenir lors d'une une intoxication au CO avec à l'extrême infarctus myocardique avec surélévation du segment ST. Il paraît légitime de proposer pour toutes les patientes: l’éloignement maternel de la source de CO; l'oxygénothérapie à 100% au masque facial par les services de secours et pendant le transfert; le traitement par oxygénothérapie hyperbare pour toutes les femmes enceintes, le plus rapidement possible et quelque soit l’âge gestationnel. PMID:26405502

Radiation resistant concrete for applications in nuclear power and radioactive waste industries

NASA Astrophysics Data System (ADS)

Burnham, Steven Robert

Elemental components of ordinary concrete contain a variety of metals and rare earth elements that are susceptible to neutron activation. This activation occurs by means of radiative capture, a neutron interaction that results in formation of radioisotopes such as Co-60, Eu-152, and Eu-154. Studies have shown that these three radioisotopes are responsible for the residual radioactivity found in nuclear power plant concrete reactor dome and shielding walls. Such concrete is classified as Low Level Radioactive Waste (LLRW) and Very Low Level Waste (VLLW) by International Atomic Energy Agency (IAEA) standards and requires disposal at appropriate disposal sites. There are only three such sites in the USA, and every nuclear power plant will produce at the time of decommissioning approximately 1,500 tonnes of activated concrete classified as LLRW and VLLW. NAVA ALIGA (ancient word for a new stone) is a new concrete mixture developed mainly by research as presented in this thesis. The purpose of NAVA ALIGA is to satisfy IAEA clearance levels if used as a material for reactor dome, spent fuel pool, or radioactive waste canisters. NAVA ALIGA will never be activated above the IAEA clearance level after long-term exposure to neutron radiation when used as a material for reactor dome, spent fuel pool, and radioactive waste canisters. Components of NAVA ALIGA were identified using Instrumental Neutron Activation Analysis (INAA) and Inductively Coupled Plasma Mass Spectrometry (ISP-MS) to determine trace element composition. In addition, it was tested for compressive strength and permeability, important for nuclear infrastructure. The studied mixture had a high water to cement ratio of 0.56, which likely resulted in the high measured permeability, yet the mixture also showed a compressive strength greater than 6 000 psi after 28 days. In addition to this experimental analysis, which goal was to develop a standard approach to define the concrete mixtures in satisfying the IAEA radiation clearance levels, the NAVA ALIGA concrete was analyzed as to potentially be used together with depleted uranium. This study was purely computational (based on MCNP6 models) and was twofold: to find if this new concrete mix would enhance the radiation shielding properties when combined with depleted uranium and to find if this will be an effective and useful way of using the existing large quantities of disposed depleted uranium.

Investigation of saliva of patients with periodontal disease using NAA

NASA Astrophysics Data System (ADS)

Zamboni, C. B.; Metairon, S.; Medeiros, I. M. M. A.; Lewgoy, H. R.

2013-05-01

In this study the non-stimulated whole saliva of 26 healthy subjects (mean age 33.9 ± 11.0 years, range: 26 to 49 years) and 11 patients with periodontal disease (mean age 41.7 ± 11.5 years; range 29 to 55 years) was investigated using Neutron Activation Analysis (NAA) technique. The samples were obtained from donors at São Paulo city (Brazil). The analyses were performed in the nuclear reactor IEA-R1 (3.5-4.5MW, pool type) at IPEN/CNEN-SP (Brazil). Considerable changes in Ca and S saliva's level were identified in patients with periodontal disease suggesting they can be used as monitors of periodontal diseases.

C3I for Crisis, Emergency and Consequence Management (C3I pour la gestion des crises, des urgences et de leurs consequences)

DTIC Science & Technology

2009-05-01

Quelque soit le contexte, l’aide à la décision passe par une analyse en profondeur de trois (3) aspects importants interdépendants, à savoir le...information, including suggestions for reducing this burden to Department of Defense , Washington Headquarters Services, Directorate for Information...type de menace, nécessite en effet d’adopter une approche collective de la sécurité étendue à une coopération avec de multiples organisations civiles

Instabilités et chaos dans les oscillateurs paramétriques optiques

NASA Astrophysics Data System (ADS)

Amon, A.; Suret, P.; Bielawski, S.; Derozier, D.; Zemmouri, J.; Lefranc, M.; Nizette, M.; Erneux, T.

2004-11-01

Nous discutons quelques mécanismes d'instabilité récemment observés dans un oscillateur paramétrique optique (OPO) : d'une part des instabilités opto-thermiques où le système oscille autour des courbes de résonance d'un ou plusieurs modes, d'autre part des oscillations rapides résultant de l'interaction de plusieurs modes transverses. La première observation expérimentale de chaos déterministe dans un OPO est également présentée.

Isotope labeling to determine the dynamics of metabolic response in CHO cell perfusion bioreactors using MALDI-TOF-MS.

PubMed

Karst, Daniel J; Steinhoff, Robert F; Kopp, Marie R G; Soos, Miroslav; Zenobi, Renato; Morbidelli, Massimo

2017-11-01

The steady-state operation of Chinese hamster ovary (CHO) cells in perfusion bioreactors requires the equilibration of reactor dynamics and cell metabolism. Accordingly, in this work we investigate the transient cellular response to changes in its environment and their interactions with the bioreactor hydrodynamics. This is done in a benchtop perfusion bioreactor using MALDI-TOF MS through isotope labeling of complex intracellular nucleotides (ATP, UTP) and nucleotide sugars (UDP-Hex, UDP-HexNAc). By switching to a 13 C 6 glucose containing feed media during constant operation at 20 × 10 6 cells and a perfusion rate of 1 reactor volume per day, isotopic steady state was studied. A step change to the 13 C 6 glucose medium in spin tubes allowed the determination of characteristic times for the intracellular turnover of unlabeled metabolites pools, τST (≤0.56 days), which were confirmed in the bioreactor. On the other hand, it is shown that the reactor residence time τR (1 day) and characteristic time for glucose uptake τGlc (0.33 days), representative of the bioreactor dynamics, delayed the consumption of 13 C 6 glucose in the bioreactor and thus the intracellular 13 C enrichment. The proposed experimental approach allowed the decoupling of bioreactor hydrodynamics and intrinsic dynamics of cell metabolism in response to a change in the cell culture environment. © 2017 American Institute of Chemical Engineers Biotechnol. Prog., 33:1630-1639, 2017. © 2017 American Institute of Chemical Engineers.

Critical experiments at Sandia National Laboratories : technical meeting on low-power critical facilities and small reactors.

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

Harms, Gary A.; Ford, John T.; Barber, Allison Delo

2010-11-01

Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-IIImore » is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark reactor physics data to support validation of the reactor physics codes used to design commercial reactor fuel elements in an enrichment range above the current 5% enrichment cap. A first set of critical experiments in the 7uPCX has been completed. More experiments are planned in the 7uPCX series. The critical experiments at Sandia National Laboratories are currently funded by the US Department of Energy Nuclear Criticality Safety Program (NCSP). The NCSP has committed to maintain the critical experiment capability at Sandia and to support the development of a critical experiments training course at the facility. The training course is intended to provide hands-on experiment experience for the training of new and re-training of practicing Nuclear Criticality Safety Engineers. The current plans are for the development of the course to continue through the first part of fiscal year 2011 with the development culminating is the delivery of a prototype of the course in the latter part of the fiscal year. The course will be available in fiscal year 2012.« less

Nuclear Power - Post Fukushima

NASA Astrophysics Data System (ADS)

Reyes, Jose, Jr.

2011-10-01

The extreme events that led to the prolonged power outage at the Fukushima Daiicchi nuclear plant have highlighted the importance of assuring a means for stable long term cooling of the nuclear fuel and containment following a complete station blackout. Legislative bodies, regulatory agencies and industry are drawing lessons from those events and considering what changes, if any, are needed to nuclear power, post Fukushima. The enhanced safety of a new class of reactor designed by NuScale Power is drawing significant attention in light of the Fukushima events. During normal operation, each NuScale containment is fully immersed in a water-filled stainless steel lined concrete pool that resides underground. The pool, housed in a Seismic Category I building, is large enough to provided 30 days of core and containment cooling without adding water. After 30 days, the decay heat generations coupled with thermal radiation heat transfer is completely adequate to remove core decay heat for an unlimited period of time. These passive power systems can perform their function without requiring an external supply of water of power. An assessment of the NuScale passive systems is being performed through a comprehensive test program that includes the NuScale integral system test facility at Oregon State University

Clinical trials for treating recurrent head and neck cancer with boron neutron capture therapy using the Tsing-Hua Open Pool Reactor.

PubMed

Wang, Ling-Wei; Liu, Yen-Wan Hsueh; Chou, Fong-In; Jiang, Shiang-Huei

2018-06-19

Head and neck (HN) cancer is an endemic disease in Taiwan, China. Locally recurrent HN cancer after full-dose irradiation poses a therapeutic challenge, and boron neutron capture therapy (BNCT) may be a solution that could provide durable local control with tolerable toxicity. The Tsing-Hua Open Pool Reactor (THOR) at National Tsing-Hua University in Hsin-Chu, provides a high-quality epithermal neutron source for basic and clinical BNCT research. Our first clinical trial, entitled "A phase I/II trial of boron neutron capture therapy for recurrent head and neck cancer at THOR", was carried out between 2010 and 2013. A total of 17 patients with 23 recurrent HN tumors who had received high-dose photon irradiation were enrolled in the study. The fructose complex of L-boronophenylalanine was used as a boron carrier, and a two-fraction BNCT treatment regimen at 28-day intervals was used for each patient. Toxicity was acceptable, and although the response rate was high (12/17), re-recurrence within or near the radiation site was common. To obtain better local control, another clinical trial entitled "A phase I/II trial of boron neutron capture therapy combined with image-guided intensity-modulated radiotherapy (IG-IMRT) for locally recurrent HN cancer" was initiated in 2014. The first administration of BNCT was performed according to our previous protocol, and IG-IMRT was initiated 28 days after BNCT. As of May 2017, seven patients have been treated with this combination. The treatment-related toxicity was similar to that previously observed with two BNCT applications. Three patients had a complete response, but locoregional recurrence was the major cause of failure despite initially good responses. Future clinical trials combining BNCT with other local or systemic treatments will be carried out for recurrent HN cancer patients at THOR.

The Development of Neutron Radiography and Tomography on a SLOWPOKE-2 Reactor

NASA Astrophysics Data System (ADS)

Bennett, L. G. I.; Lewis, W. J.; Hungler, P. C.

Development of neutron radiography at the Royal Military College of Canada (RMC) started by trying to interest the Royal Canadian Air Force (RCAF) in this new non-destructive testing (NDT) technique. A Californium-252 based device was ordered and then installed at RMC for development of applicable techniques for aircraft by the first author. A second and transportable device was then designed, modified and used in trials at RCAF Bases and other locations for one year. This activity was the only foreign loan of the U.S. Californium Loan Program. Around this time, SLOWPOKE-2 reactors were being installed at four Canadian universities, while a new science and engineering building was being built at RMC. A reactor pool was incorporated and efforts to procure a reactor succeeded a decade later with a SLOWPOKE-2 reactor being installed at RMC. The only modification by the vendor for RMC was a thermal column replacing an irradiation site inside the reactor container for a later installation of a neutron beam tube (NBT). Development of a working NBT took several years, starting with the second author. A demonstration of the actual worth of neutron radiography took place with a CF-18 Hornet aircraft being neutron and X-radiographed at McClellan Air Force Base, Sacramento, CA. This inspection was followed by one of the rudders that had indications of water ingress being radiographed successfully at RMC just after the NBT became functional. The next step was to develop a neutron radioscopy system (NRS), initially employing film and then digital imaging, and is in use today for all flight control surfaces (FCS). With the third author, a technique capable of removing water from affected FCS was developed at RMC. Heating equipment and a vacuum system were utilized to carefully remove the water. This technique was proven using a sequence of near real time neutron images obtained during the drying process. The results of the drying process were correlated with a relative humidity gauge and an NDT technique that could be performed at Canadian Forces (CF) Bases was developed. In order to determine the structural integrity of the component having undergone this water removal, further research was required into the effect of water inside composite honeycomb structures. This need has led to the present development of neutron tomography on the reactor at RMC, which is capable of determining the exact location of water ingress inside composite components. This technique has been successfully applied to coupons as well as to complete rudders.

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

On the corrosion behavior of zircaloy-4 in spent fuel pools under accidental conditions

NASA Astrophysics Data System (ADS)

Lavigne, O.; Shoji, T.; Sakaguchi, K.

2012-07-01

After zircaloy cladding tubes have been subjected to irradiation in the reactor core, they are stored temporarily in spent fuel pools. In case of an accident, the integrity of the pool may be affected and the composition of the coolant may change drastically. This was the case in Fukushima Daiichi in March 2011. Successive incidents have led to an increase in the pH of the coolant and to chloride contamination. Moreover, water radiolysis may occur owing to the remnant radioactivity of the spent fuel. In this study, we propose to evaluate the corrosion behavior of oxidized Zr-4 (in autoclave at 288 °C for 32 days) in function of the pH and the presence of chloride and radical forms. The generation of radicals is achieved by the sonolysis of the solution. It appears that the increase in pH and the presence of radicals lead to an increase in current densities. However, the current densities remain quite low (depending on the conditions, between 1 and 10 μA cm-2). The critical parameter is the presence of chloride ions. The chloride ions widely decrease the passive range of the oxidized samples (the pitting potential is measured around +0.6 V (vs. SCE)). Moreover, if the oxide layer is scratched or damaged (which is likely under accidental conditions), the pitting potential of the oxidized sample reaches the pitting potential of the non-oxidized sample (around +0.16 V (vs. SCE)), leaving a shorter stable passive range for the Zr-4 cladding tubes.

Innovations for In-Pile Measurements in the Framework of the CEA-SCK•CEN Joint Instrumentation Laboratory

NASA Astrophysics Data System (ADS)

Villard, Jean-Francois; Schyns, Marc

2010-12-01

Optimizing the life cycle of nuclear systems under safety constraints requires high-performance experimental programs to reduce uncertainties on margins and limits. In addition to improvement in modeling and simulation, innovation in instrumentation is crucial for analytical and integral experiments conducted in research reactors. The quality of nuclear research programs relies obviously on an excellent knowledge of their experimental environment which constantly calls for better online determination of neutron and gamma flux. But the combination of continuously increasing scientific requirements and new experimental domains -brought for example by Generation IV programsnecessitates also major innovations for in-pile measurements of temperature, dimensions, pressure or chemical analysis in innovative mediums. At the same time, the recent arising of a European platform around the building of the Jules Horowitz Reactor offers new opportunities for research institutes and organizations to pool their resources in order to face these technical challenges. In this situation, CEA (French Nuclear Energy Commission) and SCK'CEN (Belgian Nuclear Research Centre) have combined their efforts and now share common developments through a Joint Instrumentation Laboratory. Significant progresses have thus been obtained recently in the field of in-pile measurements, on one hand by improvement of existing measurement methods, and on the other hand by introduction in research reactors of original measurement techniques. This paper highlights the state-of-the-art and the main requirements regarding in-pile measurements, particularly for the needs of current and future irradiation programs performed in material testing reactors. Some of the main on-going developments performed in the framework of the Joint Instrumentation Laboratory are also described, such as: - a unique fast neutron flux measurement system using fission chambers with 242Pu deposit and a specific online data processing, - an optical system designed to perform in-pile dimensional measurements of material samples under irradiation, - an acoustical instrumentation allowing the online characterization of fission gas release in Pressurized Water Reactor fuel rods. For each example, the obtained results, expected impacts and development status are detailed.

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

Hecker, Siegfried S.

Actions of the Government of the Democratic People's Republic of Korea have precipitated two nuclear crises in the past 10 years. The 1994 crisis was resolved through the 'Agreed Framework.' North Korea agreed to 'freeze' and eventually dismantle its nuclear program (with U.S. help to store spent fuel safely and under IAEA inspection). In return, the United States agreed (with the KEDO international consortium) to build two light-water reactors and supply North Korea with heavy-fuel oil until the reactors come on line. In addition, both sides agreed to move towards full normalization of relations, work for peace and security onmore » a nuclear-free Korean Peninsula, and work on strengthening the international nonproliferation regime. The second nuclear crisis erupted when North Korean Government officials allegedly admitted to having a clandestine uranium enrichment program when confronted with this accusation by U.S. officials in October 2002. The United States (through KEDO) suspended heavy-fuel oil shipments and North Korea responded by expelling the IAEA inspectors, withdrawing from the Nuclear Nonproliferation Treaty, and restarting its nuclear program in January 2003. The North Korean Government has invited Professor John Lewis of Stanford University, a China and North Korea scholar, for Track I1 discussions of nuclear and other key issues since 1987. In August 2003, Professor Lewis visited North Korea just before the first six-party talks, which were designed by the United States to solve the current nuclear crisis. Professor Lewis was invited back for the January 2004 visit. He asked Jack Pritchard, former U.S. special envoy for DRPK negotiations, and me to accompany him. Two Asian affairs staff specialists from the U.S. Senate Foreign Relations Committee also joined us. I will report on the visit to the Yongbyon Nuclear Scientific Research Center on January 8,2004. We toured the 5 MWe reactor, the 50 MWe reactor construction site, the spent fuel pool storage building, and the radiochemical laboratory. We concluded that North Korea has restarted its 5 MWe reactor (which produces roughly 6 kg of plutonium annually), it removed the 8000 spent fuel rods that were previously stored under IAEA safeguards from the spent fuel pool, and that it most likely extracted the 25 to 30 kg of plutonium contained in these fuel rods. Although North Korean officials showed us what they claimed was their plutonium metal product from this reprocessing campaign, we were not able to conclude definitively that it was in fact plutonium metal and that it came from the most recent reprocessing campaign. Nevertheless, our North Korean hosts demonstrated that they had the capability, the facility and requisite capacity, and the technical expertise to produce plutonium metal. We were not shown any facilities or had the opportunity to talk to technical or military experts who were able to address the issue of whether or not North Korea had a 'deterrent' as claimed - that is, we were not able to conclude that North Korea can build a nuclear device and that it can integrate nuclear devices into suitable delivery systems. On the matter of uranium enrichment programs, Vice Minister Kim Gye Gwan categorically denied that North Korea has a uranium enrichment program - he said, 'we have no program, no equipment, and no technical expertise for uranium enrichment.' Upon return to the United States, I shared my observations and analysis with U.S. Government officials in Washington, DC, including congressional testimony to the Senate Foreign Relations Committee and briefings to two House of Representative Committees.« less

Problemes theoriques et methodologiques dans l'etude des langues/dialectes en contact aux niveaux macrologique et micrologique = Theoretical and Methodological Issues in the Study of Languages/Dialects in Contact at Macro- and Micro-Logical Levels of Analysis. Proceedings of the International Conference DALE (University of London)/ICRB (Laval University, Quebec)/ICSBT (Vrije Universiteit te Brussel) (London, England, May 23-26, 1985).

ERIC Educational Resources Information Center

Blanc, Michel, Ed.; Hamers, Josiane F., Ed.

Papers from an international conference on the interaction of languages and dialects in contact are presented in this volume. Papers include: "Quelques reflexions sur la variation linguistique"; "The Investigation of 'Language Continuum' and 'Diglossia': A Macrological Case Study and Theoretical Model"; "A Survey of…

Decrire et enseigner une competence de communication: remarques sur quelques solutions de continuite. L'Enseignement de la competence de communication en langues secondes. (Describing and Teaching Communicative Competence: Some Remarks on Solutions of Continuity. The Teaching of Communicative Competence in Second Languages.) Acts of the Colloquium of the Swiss Interuniversity Commission for Applied Linguistics. CILA Bulletin.

ERIC Educational Resources Information Center

Coste, Daniel

Two projects of the Ecole Normale Superieure de Saint-Cloud (CREDIF) are described and critically analyzed in this paper: the definition of a threshold level, "Niveau-seuil," in French and a learning module, "Looking for Work," intended to teach necessary written French to migrant workers. The threshold level section is a…

Silent Ship Research Applications and Operation. Volume 2. Unclassified Papers. Proceedings of a Conference Held at SACLANTCEN on 2-4 October 1984

DTIC Science & Technology

1985-01-15

moyennes calculees sur 62 bateaux sont priesnteesdans le tableau suivznt aoy’?nno mowvuw. desma yonn des mayavw dom % aupentation genral: bate"u A bateaux 5...i coque en bois, acier ou polyester. Le decoupaqe des variables en classes pernet de bitir deux matrices *un - tableau disjonctif complet", *un...pr6sentents quelques expertises ac oustiques provenant de deux t~tudes lr~alis~es .par le G.E.R.B.A.M. .I *la premiý-re, sur 95 thoniers ligneurs

Measurement and simulation of the TRR BNCT beam parameters

NASA Astrophysics Data System (ADS)

Bavarnegin, Elham; Sadremomtaz, Alireza; Khalafi, Hossein; Kasesaz, Yaser; Golshanian, Mohadeseh; Ghods, Hossein; Ezzati, Arsalan; Keyvani, Mehdi; Haddadi, Mohammad

2016-09-01

Recently, the configuration of the Tehran Research Reactor (TRR) thermal column has been modified and a proper thermal neutron beam for preclinical Boron Neutron Capture Therapy (BNCT) has been obtained. In this study, simulations and experimental measurements have been carried out to identify the BNCT beam parameters including the beam uniformity, the distribution of the thermal neutron dose, boron dose, gamma dose in a phantom and also the Therapeutic Gain (TG). To do this, the entire TRR structure including the reactor core, pool, the thermal column and beam tubes have been modeled using MCNPX Monte Carlo code. To measure in-phantom dose distribution a special head phantom has been constructed and foil activation techniques and TLD700 dosimeter have been used. The results show that there is enough uniformity in TRR thermal BNCT beam. TG parameter has the maximum value of 5.7 at the depth of 1 cm from the surface of the phantom, confirming that TRR thermal neutron beam has potential for being used in treatment of superficial brain tumors. For the purpose of a clinical trial, more modifications need to be done at the reactor, as, for example design, and construction of a treatment room at the beam exit which is our plan for future. To date, this beam is usable for biological studies and animal trials. There is a relatively good agreement between simulation and measurement especially within a diameter of 10 cm which is the dimension of usual BNCT beam ports. This relatively good agreement enables a more precise prediction of the irradiation conditions needed for future experiments.

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

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

Murata, K.K.; Williams, D.C.; Griffith, R.O.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of themore » input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.« less

Large-Scale Weibull Analysis of H-451 Nuclear- Grade Graphite Specimen Rupture Data

NASA Technical Reports Server (NTRS)

Nemeth, Noel N.; Walker, Andrew; Baker, Eric H.; Murthy, Pappu L.; Bratton, Robert L.

2012-01-01

A Weibull analysis was performed of the strength distribution and size effects for 2000 specimens of H-451 nuclear-grade graphite. The data, generated elsewhere, measured the tensile and four-point-flexure room-temperature rupture strength of specimens excised from a single extruded graphite log. Strength variation was compared with specimen location, size, and orientation relative to the parent body. In our study, data were progressively and extensively pooled into larger data sets to discriminate overall trends from local variations and to investigate the strength distribution. The CARES/Life and WeibPar codes were used to investigate issues regarding the size effect, Weibull parameter consistency, and nonlinear stress-strain response. Overall, the Weibull distribution described the behavior of the pooled data very well. However, the issue regarding the smaller-than-expected size effect remained. This exercise illustrated that a conservative approach using a two-parameter Weibull distribution is best for designing graphite components with low probability of failure for the in-core structures in the proposed Generation IV (Gen IV) high-temperature gas-cooled nuclear reactors. This exercise also demonstrated the continuing need to better understand the mechanisms driving stochastic strength response. Extensive appendixes are provided with this report to show all aspects of the rupture data and analytical results.

Evaluation of the Pool Critical Assembly Benchmark with Explicitly-Modeled Geometry using MCNP6

DOE PAGES

Kulesza, Joel A.; Martz, Roger Lee

2017-03-01

Despite being one of the most widely used benchmarks for qualifying light water reactor (LWR) radiation transport methods and data, no benchmark calculation of the Oak Ridge National Laboratory (ORNL) Pool Critical Assembly (PCA) pressure vessel wall benchmark facility (PVWBF) using MCNP6 with explicitly modeled core geometry exists. As such, this paper provides results for such an analysis. First, a criticality calculation is used to construct the fixed source term. Next, ADVANTG-generated variance reduction parameters are used within the final MCNP6 fixed source calculations. These calculations provide unadjusted dosimetry results using three sets of dosimetry reaction cross sections of varyingmore » ages (those packaged with MCNP6, from the IRDF-2002 multi-group library, and from the ACE-formatted IRDFF v1.05 library). These results are then compared to two different sets of measured reaction rates. The comparison agrees in an overall sense within 2% and on a specific reaction- and dosimetry location-basis within 5%. Except for the neptunium dosimetry, the individual foil raw calculation-to-experiment comparisons usually agree within 10% but is typically greater than unity. Finally, in the course of developing these calculations, geometry that has previously not been completely specified is provided herein for the convenience of future analysts.« less

Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

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

Lescop, B.; Badeau, G.; Ivanovic, S.

Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. Themore » facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can also modify the settings of the electronic system, such as the high voltage and the discrimination level in order to obtain all the characteristic curves of the detectors. These curves are used to define the right setting of the electronic system and to discuss the expected degradation of the detector signal resulting from the detector damage under the integrated neutron and gamma fluxes. Moreover, in addition to the study of the neutron detection systems itself, the integration of the measurements made by these detection systems in the logic of the safety system of the nuclear reactor is also addressed. Providing the trainees with an extensive overview of each part of the neutron monitoring instrumentation apply to a nuclear reactor, hands-on measurements on the ISIS reactor play a major role in ensuring a practical and comprehensive understanding of the neutron detection system and their integration in the safety system of nuclear reactors. It also gives a solid background for the follow up and the development of the neutron detection systems. In addition to on-reactor training, Internet Reactor Laboratory capability has been implemented on the ISIS reactor in 2014. For the Internet Reactor Laboratory an extensive video conference system has been implemented on ISIS reactor. The system includes 4 cameras and the transmission of the video signal given by the supervision system of the reactor which records and processes the data of the reactor. According to the pedagogic needs during the training courses, the lecturer on the ISIS reactor chooses to broadcast the relevant information at each stage of the course. For example, graph showing the histogram of the counting and current as a function of the time, or the electrical signal observed on the oscilloscope, can be broadcasted trough internet. By interacting through the video conference, the remote classroom is able to ask for changes in the reactor power or settings of the detection systems. They can also ask for the broadcast of some particular information. At the guest institution, the information is displayed in two parts or screens, as shown in the Figure 3. Concerning the interaction with - and the feedback from - the remote classroom, the camera of the video system in the remote classroom is used to ensure the contact between the trainees and the lecturer and reactor operators. Thus, the Internet Reactor Laboratory is complementary to the on reactor training courses. It allows distant learning, reducing the overall cost of the course when this is necessary. It can efficiently be used for the development of the human resources needed by the nuclear industry and the nuclear programs in countries without research reactors.« less

ORIGAMI Automator Primer. Automated ORIGEN Source Terms and Spent Fuel Storage Pool Analysis

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

Wieselquist, William A.; Thompson, Adam B.; Bowman, Stephen M.

2016-04-01

Source terms and spent nuclear fuel (SNF) storage pool decay heat load analyses for operating nuclear power plants require a large number of Oak Ridge Isotope Generation and Depletion (ORIGEN) calculations. SNF source term calculations also require a significant amount of bookkeeping to track quantities such as core and assembly operating histories, spent fuel pool (SFP) residence times, heavy metal masses, and enrichments. The ORIGEN Assembly Isotopics (ORIGAMI) module in the SCALE code system provides a simple scheme for entering these data. However, given the large scope of the analysis, extensive scripting is necessary to convert formats and process datamore » to create thousands of ORIGAMI input files (one per assembly) and to process the results into formats readily usable by follow-on analysis tools. This primer describes a project within the SCALE Fulcrum graphical user interface (GUI) called ORIGAMI Automator that was developed to automate the scripting and bookkeeping in large-scale source term analyses. The ORIGAMI Automator enables the analyst to (1) easily create, view, and edit the reactor site and assembly information, (2) automatically create and run ORIGAMI inputs, and (3) analyze the results from ORIGAMI. ORIGAMI Automator uses the standard ORIGEN binary concentrations files produced by ORIGAMI, with concentrations available at all time points in each assembly’s life. The GUI plots results such as mass, concentration, activity, and decay heat using a powerful new ORIGEN Post-Processing Utility for SCALE (OPUS) GUI component. This document includes a description and user guide for the GUI, a step-by-step tutorial for a simplified scenario, and appendices that document the file structures used.« less

Speciation and reactivity of uranium products formed during in situ bioremediation in a shallow alluvial aquifer.

PubMed

Alessi, Daniel S; Lezama-Pacheco, Juan S; Janot, Noémie; Suvorova, Elena I; Cerrato, José M; Giammar, Daniel E; Davis, James A; Fox, Patricia M; Williams, Kenneth H; Long, Philip E; Handley, Kim M; Bernier-Latmani, Rizlan; Bargar, John R

2014-11-04

In this study, we report the results of in situ U(VI) bioreduction experiments at the Integrated Field Research Challenge site in Rifle, Colorado, USA. Columns filled with sediments were deployed into a groundwater well at the site and, after a period of conditioning with groundwater, were amended with a mixture of groundwater, soluble U(VI), and acetate to stimulate the growth of indigenous microorganisms. Individual reactors were collected as various redox regimes in the column sediments were achieved: (i) during iron reduction, (ii) just after the onset of sulfate reduction, and (iii) later into sulfate reduction. The speciation of U retained in the sediments was studied using X-ray absorption spectroscopy, electron microscopy, and chemical extractions. Circa 90% of the total uranium was reduced to U(IV) in each reactor. Noncrystalline U(IV) comprised about two-thirds of the U(IV) pool, across large changes in microbial community structure, redox regime, total uranium accumulation, and reaction time. A significant body of recent research has demonstrated that noncrystalline U(IV) species are more suceptible to remobilization and reoxidation than crystalline U(IV) phases such as uraninite. Our results highlight the importance of considering noncrystalline U(IV) formation across a wide range of aquifer parameters when designing in situ remediation plans.

Speciation and Reactivity of Uranium Products Formed during in Situ Bioremediation in a Shallow Alluvial Aquifer

PubMed Central

2015-01-01

In this study, we report the results of in situ U(VI) bioreduction experiments at the Integrated Field Research Challenge site in Rifle, Colorado, USA. Columns filled with sediments were deployed into a groundwater well at the site and, after a period of conditioning with groundwater, were amended with a mixture of groundwater, soluble U(VI), and acetate to stimulate the growth of indigenous microorganisms. Individual reactors were collected as various redox regimes in the column sediments were achieved: (i) during iron reduction, (ii) just after the onset of sulfate reduction, and (iii) later into sulfate reduction. The speciation of U retained in the sediments was studied using X-ray absorption spectroscopy, electron microscopy, and chemical extractions. Circa 90% of the total uranium was reduced to U(IV) in each reactor. Noncrystalline U(IV) comprised about two-thirds of the U(IV) pool, across large changes in microbial community structure, redox regime, total uranium accumulation, and reaction time. A significant body of recent research has demonstrated that noncrystalline U(IV) species are more suceptible to remobilization and reoxidation than crystalline U(IV) phases such as uraninite. Our results highlight the importance of considering noncrystalline U(IV) formation across a wide range of aquifer parameters when designing in situ remediation plans. PMID:25265543

A feasibility study of the Tehran research reactor as a neutron source for BNCT.

PubMed

Kasesaz, Yaser; Khalafi, Hossein; Rahmani, Faezeh; Ezati, Arsalan; Keyvani, Mehdi; Hossnirokh, Ashkan; Shamami, Mehrdad Azizi; Monshizadeh, Mahdi

2014-08-01

Investigation on the use of the Tehran Research Reactor (TRR) as a neutron source for Boron Neutron Capture Therapy (BNCT) has been performed by calculating and measuring energy spectrum and the spatial distribution of neutrons in all external irradiation facilities, including six beam tubes, thermal column, and the medical room. Activation methods with multiple foils and a copper wire have been used for the mentioned measurements. The results show that (1) the small diameter and long length beam tubes cannot provide sufficient neutron flux for BNCT; (2) in order to use the medical room, the TRR core should be placed in the open pool position, in this situation the distance between the core and patient position is about 400 cm, so neutron flux cannot be sufficient for BNCT; and (3) the best facility which can be adapted for BNCT application is the thermal column, if all graphite blocks can be removed. The epithermal and fast neutron flux at the beginning of this empty column are 4.12×10(9) and 1.21×10(9) n/cm(2)/s, respectively, which can provide an appropriate neutron beam for BNCT by designing and constructing a proper Beam Shaping Assembly (BSA) structure. Copyright © 2014 Elsevier Ltd. All rights reserved.

Supported molten-metal catalysts

DOEpatents

Datta, Ravindra; Singh, Ajeet; Halasz, Istvan; Serban, Manuela

2001-01-01

An entirely new class of catalysts called supported molten-metal catalysts, SMMC, which can replace some of the existing precious metal catalysts used in the production of fuels, commodity chemicals, and fine chemicals, as well as in combating pollution. SMMC are based on supporting ultra-thin films or micro-droplets of the relatively low-melting (<600.degree. C.), inexpensive, and abundant metals and semimetals from groups 1, 12, 13, 14, 15 and 16, of the periodic table, or their alloys and intermetallic compounds, on porous refractory supports, much like supported microcrystallites of the traditional solid metal catalysts. It thus provides orders of magnitude higher surface area than is obtainable in conventional reactors containing molten metals in pool form and also avoids corrosion. These have so far been the chief stumbling blocks in the application of molten metal catalysts.

Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

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

Ade, Brian J; Marshall, William BJ J; Martinez-Gonzalez, Jesus S

Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents themore » analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.« less

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

Bernard, D.; Fabbris, O.

Two different experiments performed in the 8 MWth MELUSINE experimental power pool reactor aimed at analyzing 1 GWd/t spent fuel pellets doped with several actinides. The goal was to measure the averaged neutron induced capture cross section in two very different neutron spectra (a PWR-like and an under-moderated one). This paper summarizes the combined deterministic APOLLO2-stochastic TRIPOLI4 analysis using the JEFF-3.1.1 European nuclear data library. A very good agreement is observed for most of neutron induced capture cross section of actinides and a clear underestimation for the {sup 241}Am(n,{gamma}) as an accurate validation of its associated isomeric ratio are emphasized.more » Finally, a possible huge resonant fluctuation (factor of 2.7 regarding to the 1=0 resonance total orbital momenta) is suggested for isomeric ratio. (authors)« less

Un cosmologiste oublié: Jean Henri Lambert

NASA Astrophysics Data System (ADS)

Débarbat, Suzanne; Lévy, Jacques

Si les travaux de Kepler ont eu une large influence sure les progrès réalisés en astronomie au cours du 17e siècle, le Siècle de lumières a vu apparaître de nouvelles conceptions. La court vie de J.H. lambert s'inscrit dans le 18e siècle. Il s'agit d'un nom bien connu dans différents domaines (photométrie, projections cartographiques, mathématiques appliquées, etc.); mais il n'est guàre mentionné en cosmologie, alors que Lambert y a fourni une contribution originale offrant quelques suprenantes anticipations...

A Selection of Test Cases for the Validation of Large-Eddy Simulations of Turbulent Flows (Quelques cas d’essai pour la validation de la simulation des gros tourbillons dans les ecoulements turbulents)

DTIC Science & Technology

1998-04-01

they approach the more useful (higher) Reynolds numbers. 8.6 SUMMARY OF COMPLEX FLOWS SQUARE DUCT CMPO00 UDOv 6.5 x 10’i E Yokosawa ei al. 164] pg...Sheets for: Chapter 8. Complex Flows 184 185 CMPOO: Flow in a square duct - Experiments Yokosawa , Fujita, Hirota, & Iwata 1. Description of the flow...These are the experiments of Yokosawa ei al (1989). Air was blown through a flow meter and a settling chamber into a square duct. Measuremsents were

Reconstruction du Flux d'Energie et Recherche de Squarks et Gluinos dans l'Experience D0 (in French)

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

Ridel, Melissa

2002-01-01

Le modèle standard décrit la matière et les interactions fondamentales qui la gouvernent (électromagnétique, faible et forte). L'analyse des données accumulées jusqu'à présent conffrme ces prédictions notamment les mesures de précision effectuées à LEP. Malgré tout, il doit se confronter à quelques dicultés théoriques qui laisseraient penser que le Modèle Standard n'est que la théorie effective d'une autre théorie à plus haute énergie....

Study of Behavior of Some Varieties of Belgian Potatoes Subjected to Gamma Irradiation; ETUDE DU COMPORTEMENT DE QUELQUES VARIETES BELGES DE POMMES DE TERRE SOUMISES A L'IRRADIATION GAMMA

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

Kirchmann, R.; De Proost, M.; Demalsy, P.

1962-07-01

Different varieties of potatoes were irradiated with doses between 5000 and 20000 rads and stored at two different temperatures. Irradiation has a grent influence on the weight loss of the potatoes during storage; the degree of sprout inhibition depends on the variety of the potatoes. The glutathione content and the oxygen consumption of potatoes are influenced by irradiation. The greatest effect of irradiation on the chemical composition concerns the starch; an increase in sugar content is observed. The culinary properties of potatoes are not changed by irradiation. (auth)

PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

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

Blakeman, Edward D; Peplow, Douglas E.; Wagner, John C

2007-09-01

The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally filesmore » and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts.« less

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

NASA Astrophysics Data System (ADS)

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

2017-11-01

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

An intrinsically safe facility for forefront research and training on nuclear technologies — General description of the system

NASA Astrophysics Data System (ADS)

Mansani, L.; Bruzzone, M.; Frambati, S.; Reale, M.

2014-04-01

In the framework of research on generation-IV reactors, it is very important to have infrastructures specifically dedicated to the study of fundamental parameters in dynamics and kinetics of future fast-neutron reactors. Among various options pursued by international groups, Italy focused on lead-cooled reactors, which guarantee minimal neutron slowdown and capture and efficient cooling. In this paper it is described the design of a the low-power prototype generator, LEADS, that could be used within research facilities such as the National Laboratory of Legnaro of the INFN. The LEADS has a high safety standard in order to be used as a training facility, but it has also a good flexibility so as to allow a wide range of measurements and experiments. A high safety standard is achieved by limiting the reactor power to less than few hundred kW and the neutron multiplication factor k eff to less than 0.95 (a limiting value for spent fuel pool), by using a pure-uranium fuel (no plutonium) and by using solid lead as a diffuser. The proposed core is therefore intrinsically subcritical and has to be driven by an external neutron source generated by a proton beam impinging in a target. Preliminary simulations, performed with the MCNPX code indicated, for a 0.75mA continuous proton beam current at 70MeV proton energy, a reactor power of about 190kW when using a beryllium converter. The enriched-uranium fuel elements are immersed in a solid-lead matrix and contained within a steel vessel. The system is cooled by helium gas, which is transparent to neutrons and does not undergo activation. The gas is pumped by a compressor through specific holes at the entrance of the active volume with a temperature which varies according to the operating conditions and a pressure of about 1.1MPa. The hot gas coming out of the vessel is cooled by an external helium-water heat exchanger. The beryllium converter is cooled by its dedicated helium gas cooling system. After shutdown, the decay is completely dissipated by conduction through the lead reflector and steel vessel, and then evacuated by irradiation from the vessel surface to the external ambient air.

Sister Lab Program Prospective Partner Nuclear Profile: Indonesia

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

Bissani, M; Tyson, S

2006-12-14

Indonesia has participated in cooperative technical programs with the IAEA since 1957, and has cooperated with regional partners in all of the traditional areas where nuclear science is employed: in medicine, public health (such as insect control and eradication programs), agriculture (e.g. development of improved varieties of rice), and the gas and oil industries. Recently, Indonesia has contributed significantly to the Reduced Enrichment Research and Training Reactor (RERTR) Program by conducting experiments to confirm the feasibility of Mo-99 production using high-density low enriched uranium (LEU) fuel, a primary goal of the RERTR Program. Indonesia's first research reactor, the TRIGA Markmore » II at Bandung, began operation in 1964 at 250 kW and was subsequently upgraded in 1971 to 1 MW and further upgraded in 2000 to 2 MW. This reactor was joined by another TRIGA Mark II, the 100-kW Kartini-PPNY at Yogyakarta, in 1979, and by the 30-MW G.A. Siwabessy multipurpose reactor in Serpong, which achieved criticality in July 1983. A 10-MW radioisotope production reactor, to be called the RPI-10, also was proposed for construction at Serpong in the late 1990s, but the project apparently was not carried out. In the five decades since its nuclear research program began, Indonesia has trained a cadre of scientific and technical staff who not only operate and conduct research with the current facilities, but also represent the nucleus of a skilled labor pool to support development of a nuclear power program. Although Indonesia's previous on-again, off-again consideration of nuclear power has not gotten very far in the past, it now appears that Indonesia again is giving serious consideration to beginning a national nuclear energy program. In June 2006, Research and Technology Minister Kusmayanto Kadiman said that his ministry was currently putting the necessary procedures in place to speed up the project to acquire a nuclear power plant, indicating that, ''We will need around five years to complete the project. If we can start the study, go to tender, and sign the contract for the project this year, the power plant could be on stream by 2011''. While this ambitious schedule may be a bit unrealistic, it suggests new momentum to move forward on the project. The favored site for the proposed plant is the Muria Peninsula, located on Java's north central coast.« less

Effects of Fuel Composition on Combustion Stability and NO X Emissions for Traditional and Alternative Jet Fuels

NASA Astrophysics Data System (ADS)

Vijlee, Shazib Z.

Synthetic jet fuels are studied to help understand their viability as alternatives to traditionally derived jet fuel. Two combustion parameters -- flame stability and NOX emissions -- are used to compare these fuels through experiments and models. At its core, this is a fuels study comparing how chemical makeup and behavior relate. Six 'real', complex fuels are studied in this work -- four are synthetic from alternative sources and two are traditional from petroleum sources. Two of the synthetic fuels are derived from natural gas and coal via the Fischer Tropsch catalytic process. The other two are derived from Camelina oil and tallow via hydroprocessing. The traditional military jet fuel, JP8, is used as a baseline as it is derived from petroleum. The sixth fuel is derived from petroleum and is used to study the effects of aromatic content on the synthetic fuels. The synthetic fuels lack aromatic compounds, which are an important class of hydrocarbons necessary for fuel handling systems to function properly. Several single-component fuels are studied (through models and/or experiments) to facilitate interpretation and understanding. The flame stability study first compares all the 'real', complex fuels for blowout. A toroidal stirred reactor is used to try and isolate temperature and chemical effects. The modeling study of blowout in the toroidal reactor is the key to understanding any fuel-based differences in blowout behavior. A detailed, reacting CFD model of methane is used to understand how the reactor stabilizes the flame and how that changes as the reactor approaches blowout. A 22 species reduced form of GRI 3.0 is used to model methane chemistry. The knowledge of the radical species role is utilized to investigate the differences between a highly aliphatic fuel (surrogated by iso-octane) and a highly aromatic fuel (surrogated by toluene). A perfectly stirred reactor model is used to study the chemical kinetic pathways for these fuels near blowout. The differences in flame stabilization can be attributed to the rate at which these fuels are attacked and destroyed by radical species. The slow disintegration of the aromatic rings reduces the radical pool available for chain-initiating and chain-branching, which ultimately leads to an earlier blowout. The NOX study compares JP8, the aromatic additive, the synthetic fuels with and without an aromatic additive, and an aromatic surrogate (1,3,5-trimethylbenzene). A jet stirred reactor is used to try and isolate temperature and chemical effects. The reactor has a volume of 15.8 mL and a residence time of approximately 2.5 ms. The fuel flow rate (hence equivalence ratio) is adjusted to achieve nominally consistent temperatures of 1800, 1850, and 1900K. Small oscillations in fuel flow rate cause the data to appear in bands, which facilitated Arrhenius-type NOX-temperature correlations for direct comparison between fuels. The fuel comparisons are somewhat inconsistent, especially when the aromatic fuel is blended into the synthetic fuels. In general, the aromatic surrogate (1,3,5-trimethylbenzene) produces the most NOX, followed by JP8. The synthetic fuels (without aromatic additive) are always in the same ranking order for NOX production (HP Camelina > FT Coal > FT Natural Gas > HP Tallow). The aromatic additive ranks differently based on the temperature, which appears to indicate that some of the differences in NOX formation are due to the Zeldovich NOX formation pathway. The aromatic additive increases NOX for the HP Tallow and decreases NOX for the FT Coal. The aromatic additive causes increased NOX at low temperatures but decreases NOX at high temperatures for the HP Camelina and FT Natural Gas. A single perfectly stirred reactor model is used with several chemical kinetic mechanisms to study the effects of fuel (and fuel class) on NO X formation. The 27 unique NOX formation reactions from GRI 3.0 are added to published mechanisms for jet fuel surrogates. The investigation first looked at iso-octane and toluene and found that toluene produces more NOX because of a larger pool of O radical. The O radical concentration was lower for iso-octane because of an increased concentration of methyl (CH 3) radical that consumes O radical readily. Several surrogate fuels (iso-octane, toluene, propylcyclohexane, n-octane, and 1,3,5-trimethylbenzene) are modeled to look for differences in NOX production. The trend (increased CH3 → decreased O → decreased NOX) is consistently true for all surrogate fuels with multiple kinetic mechanisms. It appears that the manner in which the fuel disintegrates and creates methyl radical is an extremely important aspect of how much NOX a fuel will produce. (Abstract shortened by UMI.).

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

Brandstätter, Christian, E-mail: bran.chri@gmail.com; Laner, David, E-mail: david.laner@tuwien.ac.at; Fellner, Johann, E-mail: johann.fellner@tuwien.ac.at

Graphical abstract: Display Omitted - Highlights: • 40 year old waste from an old MSW landfill was incubated in LSR experiments. • Carbon balances for anaerobic and aerobic waste degradation were established. • The transformation of carbon pools during waste degradation was investigated. • Waste aeration resulted in the formation of a new, stable organic carbon pool. • Water addition did not have a significant effect on aerobic waste degradation. - Abstract: Landfill aeration has been proven to accelerate the degradation of organic matter in landfills in comparison to anaerobic decomposition. The present study aims to evaluate pools of organicmore » matter decomposing under aerobic and anaerobic conditions using landfill simulation reactors (LSR) filled with 40 year old waste from a former MSW landfill. The LSR were operated for 27 months, whereby the waste in one pair was kept under anaerobic conditions and the four other LSRs were aerated. Two of the aerated LSR were run with leachate recirculation and water addition and two without. The organic carbon in the solid waste was characterized at the beginning and at the end of the experiments and major carbon flows (e.g. TOC in leachate, gaseous CO{sub 2} and CH{sub 4}) were monitored during operation. After the termination of the experiments, the waste from the anaerobic LSRs exhibited a long-term gas production potential of more than 20 NL kg{sup −1} dry waste, which corresponded to the mineralization of around 12% of the initial TOC (67 g kg{sup −1} dry waste). Compared to that, aeration led to threefold decrease in TOC (32–36% of the initial TOC were mineralized), without apparent differences in carbon discharge between the aerobic set ups with and without water addition. Based on the investigation of the carbon pools it could be demonstrated that a bit more than 10% of the initially present organic carbon was transformed into more recalcitrant forms, presumably due to the formation of humic substances. The source of anaerobic degradation could be identified mainly as cellulose which played a minor role during aerobic degradation in the experiment.« less

The international atom: evolution of radiation control programs.

PubMed

Bradley, F J

2002-07-01

Under the Atoms for Peace program, Turkey received a one MWt swimming pool reactor in 1962 that initiated a health physics program for the reactor and a Radiation Control Program (RCP) for the country's use of ionizing radiation. Today, over 13,000 radiation workers, concentrated in the medical field, provide improved medical care with 6,200 x-ray units, including 494 CAT scanners, 222 radioimmunoassay (RIA) labs and 42 radiotherapy centers. Industry has a large stake in the safe use of ionizing radiation with over 1,200 x-ray and gamma radiography and fluoroscopic units, 2,500 gauges in automated process control and five irradiators. A 48-person RCP staff oversees this expanded radiation use. One incident involving a spent 3.3 TBq (88 Ci) 60Co source resulted in 10 overexposures but no fatalities. Taiwan received a 1.6 MWt swimming pool reactor in 1961 and rapidly applied nuclear technology to the medical and industrial fields. Today, there are approximately 24,000 licensed radiation workers in nuclear power field, industry, medicine and academia. Four BWRs and two PWRs supply about 25% of the island's electrical power needs. One traumatic event galvanized the RCP when an undetermined amount of 60Co was accidentally incorporated into reinforcing bars, which in turn were incorporated into residential and commercial buildings. Public exposures were estimated to range up to 15 mSv (1.3 rem) per annum. There were no reported ill effects, except possibly psychological, to date. The RCP now has instituted stringent control measures to ensure radiation-free dwellings and work places. Albania's RCP is described as it evolved since 1972. Regulations were promulgated which followed the IAEA Basic Safety Standards of that era. With 525 licenses and 600 radiation workers, the problem was not in the regulations per se but in their enforcement. The IAEA helped to upgrade the RCP as the economy evolved from one that was centrally planned economy to a free market economy. As this transition takes place, public radiation exposures in the medical field will continue to be high until the old x-ray equipment is phased out. A small conscientious health physics staff works with limited resources to keep radiation exposures at acceptable levels. These three country RCPs, as they have evolved, have some commonality. Today, all radiation installations are licensed, both for radioactive material and x-ray equipment. Radiation workers are individually licensed or registered. All RCPs have, or are striving to have, their radiation regulations conform to ICRP 60 recommendations as spelled out in the Basic Safety Standard (1996). Finally, all three countries have as yet to find a permanent solution for their radioactive waste.

The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

NASA Astrophysics Data System (ADS)

Dautray, Robert

2011-06-01

The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation, etc., radioprotection measures and treatment for the "transuranic" elements. For a long period of time, France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts and this could per se constitute one of the national priorities. Nous sommes naturellement bien conscients de l'énorme problème qui se pose au Japon actuellement comme suite au tremblement de terre et au tsunami de mars 2011 et leurs conséquences, notamment sur des installations électronucléaires. Le texte que nous présentons concerne des conditions totalement générales, indépendantes des problèmes spécifiques de sûreté qu'il faudra, de toute façon, traiter dans le cadre d'un développement éventuel de l'énergie nucléaire.We are aware, of course, of the huge problem that Japan has to deal with the aftermath of the quake and tsunami of March 2011 and their consequences on electronuclear power plants. The text that we present here concerns general physical topics independent of the specific safety problems, general physical topics which will have to be solved in the case of a contingent development of electronuclear power plants.

Preliminary Concept of Operations for the Spent Fuel Management System--WM2017

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

Cumberland, Riley M; Adeniyi, Abiodun Idowu; Howard, Rob L

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the U.S. Department of Energy s Office of Nuclear Energy is tasked with identifying, planning, and conducting activities to lay the groundwork for developing interim storage and transportation capabilities in support of an integrated waste management system. The system will provide interim storage for commercial spent nuclear fuel (SNF) from reactor sites and deliver it to a repository. The system will also include multiple subsystems, potentially including; one or more interim storage facilities (ISF); one or more repositories; facilities to package and/or repackage SNF; and transportation systems. The project teammore » is analyzing options for an integrated waste management system. To support analysis, the project team has developed a Concept of Operations document that describes both the potential integrated system and inter-dependencies between system components. The goal of this work is to aid systems analysts in the development of consistent models across the project, which involves multiple investigators. The Concept of Operations document will be updated periodically as new developments emerge. At a high level, SNF is expected to travel from reactors to a repository. SNF is first unloaded from reactors and placed in spent fuel pools for wet storage at utility sites. After the SNF has cooled enough to satisfy loading limits, it is placed in a container at reactor sites for storage and/or transportation. After transportation requirements are met, the SNF is transported to an ISF to store the SNF until a repository is developed or directly to a repository if available. While the high level operation of the system is straightforward, analysts must evaluate numerous alternative options. Alternative options include the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if any), repackaging technology, the types of containers used, repository design, component sizing, and timing of events. These alternative options arise due to technological, economic, or policy considerations. As new developments regularly emerge, the operational concepts will be periodically updated. This paper gives an overview of the different potential alternatives identified in the Concept of Operations document at a conceptual level.« less

Preliminary topical report on comparison reactor disassembly calculations

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

McLaughlin, T.P.

1975-11-01

Preliminary results of comparison disassembly calculations for a representative LMFBR model (2100-l voided core) and arbitrary accident conditions are described. The analytical methods employed were the computer programs: FX2- POOL, PAD, and VENUS-II. The calculated fission energy depositions are in good agreement, as are measures of the destructive potential of the excursions, kinetic energy, and work. However, in some cases the resulting fuel temperatures are substantially divergent. Differences in the fission energy deposition appear to be attributable to residual inconsistencies in specifying the comparison cases. In contrast, temperature discrepancies probably stem from basic differences in the energy partition models inherentmore » in the codes. Although explanations of the discrepancies are being pursued, the preliminary results indicate that all three computational methods provide a consistent, global characterization of the contrived disassembly accident. (auth)« less

Characteristics and application of spherical-type activation detectors in neutron spectrum measurements at a boron neutron capture therapy (BNCT) facility

NASA Astrophysics Data System (ADS)

Lin, Heng-Xiao; Chen, Wei-Lin; Liu, Yuan-Hao; Sheu, Rong-Jiun

2016-03-01

A set of spherical-type activation detectors was developed aiming to provide better determination of the neutron spectrum at the Tsing Hua Open-pool Reactor (THOR) BNCT facility. An activation foil embedded in a specially designed spherical holder exhibits three advantages: (1) minimizing the effect of neutron angular dependence, (2) creating response functions with broadened coverage of neutron energies by introducing additional moderators or absorbers to the central activation foil, and (3) reducing irradiation time because of improved detection efficiencies to epithermal neutron beam. This paper presents the design concept and the calculated response functions of new detectors. Theoretical and experimental demonstrations of the performance of the detectors are provided through comparisons of the unfolded neutron spectra determined using this method and conventional multiple-foil activation techniques.

Rate measurements of the hydrolysis of complex organic macromolecules in cold aqueous solutions: implications for prebiotic chemistry on the early Earth and Titan.

PubMed

Neish, C D; Somogyi, A; Imanaka, H; Lunine, J I; Smith, M A

2008-04-01

Organic macromolecules ("complex tholins") were synthesized from a 0.95 N(2)/0.05 CH(4) atmosphere in a high-voltage AC flow discharge reactor. When placed in liquid water, specific water soluble compounds in the macromolecules demonstrated Arrhenius type first order kinetics between 273 and 313 K and produced oxygenated organic species with activation energies in the range of approximately 60+/-10 kJ mol(-1). These reactions displayed half lives between 0.3 and 17 days at 273 K. Oxygen incorporation into such materials--a necessary step toward the formation of biological molecules--is therefore fast compared to processes that occur on geologic timescales, which include the freezing of impact melt pools and possible cryovolcanic sites on Saturn's organic-rich moon Titan.

Rate Measurements of the Hydrolysis of Complex Organic Macromolecules in Cold Aqueous Solutions: Implications for Prebiotic Chemistry on the Early Earth and Titan

NASA Astrophysics Data System (ADS)

Neish, C. D.; Somogyi, Á.; Imanaka, H .; Lunine, J. I.; Smith, M. A.

2008-04-01

Organic macromolecules (``complex tholins'') were synthesized from a 0.95 N2 / 0.05 CH4 atmosphere in a high-voltage AC flow discharge reactor. When placed in liquid water, specific water soluble compounds in the macromolecules demonstrated Arrhenius type first order kinetics between 273 and 313 K and produced oxygenated organic species with activation energies in the range of ~60 +/- 10 kJ mol-1. These reactions displayed half lives between 0.3 and 17 days at 273 K. Oxygen incorporation into such materials-a necessary step toward the formation of biological molecules-is therefore fast compared to processes that occur on geologic timescales, which include the freezing of impact melt pools and possible cryovolcanic sites on Saturn's organic-rich moon Titan.

Numerical study of effect of the gas-coolant free surface on the droplet fragmentation behavior of coolants

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

Li, H.X.; Anh, B.V.; Dinh, T.N.

1999-07-01

This paper presents results of a numerical investigation on the behavior of melt drops falling in a gas (vapor) space and then penetrating into a liquid volume through the gas-liquid interface. The phenomenon studied here is, usually, observed when a liquid drop falls through air into a water pool and is, specially, of interest when a hypothetical severe reactor core meltdown accident is considered. The objective of this work is to study the effect of the gas-liquid interface on the dynamic evolution of the interaction area between the fragmenting melt drop and water. In the present study, the Navier-Stokes equationsmore » are solved for three phases (gas, liquid and melt-drop) using a higher-order, explicit, numerical method, called Cubic-Interpolated Pseudo-Particle (CIP) method, which is employed in combination with an advanced front-capturing scheme, named the Level Set Algorithm (LSA). By using this method, reasonable physical pictures of droplet deformation and fragmentation during movement in a stationary uniform water pool, and in a gas-liquid two-layer volume, is simulated. Effect of the gas-liquid interface on the drop deformation and fragmentation is analyzed by comparing the simulation results obtained for the two cases. Effects of the drop geometry, and of the flow conditions, on the behavior of the melt drop are also analyzed.« less

Synchronized dynamics of bacterial niche-specific functions during biofilm development in a cold seep brine pool.

PubMed

Zhang, Weipeng; Wang, Yong; Bougouffa, Salim; Tian, Renmao; Cao, Huiluo; Li, Yongxin; Cai, Lin; Wong, Yue Him; Zhang, Gen; Zhou, Guowei; Zhang, Xixiang; Bajic, Vladimir B; Al-Suwailem, Abdulaziz; Qian, Pei-Yuan

2015-10-01

The biology of biofilm in deep-sea environments is barely being explored. Here, biofilms were developed at the brine pool (characterized by limited carbon sources) and the normal bottom water adjacent to Thuwal cold seeps. Comparative metagenomics based on 50 Gb datasets identified polysaccharide degradation, nitrate reduction and proteolysis as enriched functional categories for brine biofilms. The genomes of two dominant species: a novel Deltaproteobacterium and a novel Epsilonproteobacterium in the brine biofilms were reconstructed. Despite rather small genome sizes, the Deltaproteobacterium possessed enhanced polysaccharide fermentation pathways, whereas the Epsilonproteobacterium was a versatile nitrogen reactor possessing nar, nap and nif gene clusters. These metabolic functions, together with specific regulatory and hypersaline-tolerant genes, made the two bacteria unique compared with their close relatives, including those from hydrothermal vents. Moreover, these functions were regulated by biofilm development, as both the abundance and the expression level of key functional genes were higher in later stage biofilms, and co-occurrences between the two dominant bacteria were demonstrated. Collectively, unique mechanisms were revealed: (i) polysaccharides fermentation, proteolysis interacted with nitrogen cycling to form a complex chain for energy generation, and (ii) remarkably exploiting and organizing niche-specific functions would be an important strategy for biofilm-dependent adaptation to the extreme conditions. © 2015 Society for Applied Microbiology and John Wiley & Sons Ltd.

AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

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

Gorgemans, J.; Mulhollem, L.; Glavin, J.

2012-07-01

The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first levelmore » of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all possible pool heat load conditions. - After 3 days, several different means are provided to continue spent fuel cooling using installed plant equipment as well as off-site equipment with built-in connections. Even for beyond design basis accidents with postulated pool damage and multiple failures in the passive safety-related systems and in the defense-in-depth active systems, the AP1000 multiple spent fuel pool spray and fill systems provide additional lines of defense to prevent spent fuel damage. (authors)« less

Study of the Retention of Fission Products by a Few Common Minerals. Application to the Treatment of Medium Activity Effluents; ETUDE DE LA RETENTION DES PRODUITS DE FISSION PAR QUELQUES MINERAUX USUELS. (APPLICATION AUX TRAITEMENTS D'EFFLUENTS DE MOYENNE ACTIVITE SPECIFIQUE)

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

Auchapt, J.M.

1962-01-01

The conditions in which Sr is fixed on calcite (the object of Geneva report P/395-USA-- 1958) are more closely studied and the work is extended to five fission products in the effluerts and to 17 common rocks and minerals. Although this fixation is not suitsble as a method of treating STE effluents (i.e., those from the effluent treatment plant at MIarcoule), the study shows that all the crystals considered are strongly contaminated by simple contact. (auth)

(135)Cs/(137)Cs isotopic ratio as a new tracer of radiocesium released from the Fukushima nuclear accident.

PubMed

Zheng, Jian; Tagami, Keiko; Bu, Wenting; Uchida, Shigeo; Watanabe, Yoshito; Kubota, Yoshihisa; Fuma, Shoichi; Ihara, Sadao

2014-05-20

Since the Fukushima Daiichi nuclear power plant (FDNPP) accident in 2011, intensive studies of the distribution of released fission products, in particular (134)Cs and (137)Cs, in the environment have been conducted. However, the release sources, that is, the damaged reactors or the spent fuel pools, have not been identified, which resulted in great variation in the estimated amounts of (137)Cs released. Here, we investigated heavily contaminated environmental samples (litter, lichen, and soil) collected from Fukushima forests for the long-lived (135)Cs (half-life of 2 × 10(6) years), which is usually difficult to measure using decay-counting techniques. Using a newly developed triple-quadrupole inductively coupled plasma tandem mass spectrometry method, we analyzed the (135)Cs/(137)Cs isotopic ratio of the FDNPP-released radiocesium in environmental samples. We demonstrated that radiocesium was mainly released from the Unit 2 reactor. Considering the fact that the widely used tracer for the released Fukushima accident-sourced radiocesium in the environment, the (134)Cs/(137)Cs activity ratio, will become unavailable in the near future because of the short half-life of (134)Cs (2.06 years), the (135)Cs/(137)Cs isotopic ratio can be considered as a new tracer for source identification and long-term estimation of the mobility of released radiocesium in the environment.

Nuclear Research Reactor IEA-R1 - A Study of the Preparing for Decommissioning

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

Lopes, Valdir Maciel; Filho, Tufic Madi; Ricci, Walter

2015-07-01

The Institute of Energy and Nuclear Research (IPEN), Sao Paulo, according to the assignments given by the National Commission of Nuclear Energy (CNEN), enabled the development of this study, especially operational reports about refurbishing carried out on 2013, involving the production of radioisotopes and research in the areas of Radiochemistry and Nuclear Physics. These reports are made in accordance with established standard procedures to meet the requirements of CNEN (National Nuclear Energy Commission, the regulator the nuclear area activities in Brazil) and IAEA (International Atomic Energy Agency). This study presents an assessment of the procedures and methods of treatments formore » decontamination of the refrigeration primary circuit and changes parts, equipment and tubes of the of the IEA-R1 nuclear research reactor, pool type, power between 3,5 and 4,5 MW. In order to have a sequence in the work, the well-known contaminant radioisotopes were evaluated firstly, using Geiger- Muller equipment. In the second phase, the decontamination was done manually together with the ultrasound cleaning and washing equipment. From the several water solutions of citric acid assessment, the concentration with better confidence was obtained; in order to achieve the best results for decontamination. This study intends to define the best process for decontamination with low taxes of waste and without expensive costs. (authors)« less

H-division quarterly report, October--December 1977. [Lawrence Livermore Laboratory

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

Not Available

1978-02-10

The Theoretical EOS Group develops theoretical techniques for describing material properties under extreme conditions and constructs equation-of-state (EOS) tables for specific applications. Work this quarter concentrated on a Li equation of state, equation of state for equilibrium plasma, improved ion corrections to the Thomas--Fermi--Kirzhnitz theory, and theoretical estimates of high-pressure melting in metals. The Experimental Physics Group investigates properties of materials at extreme conditions of pressure and temperature, and develops new experimental techniques. Effort this quarter concerned the following: parabolic projectile distortion in the two-state light-gas gun, construction of a ballistic range for long-rod penetrators, thermodynamics and sound velocities inmore » liquid metals, isobaric expansion measurements in Pt, and calculation of the velocity--mass profile of a jet produced by a shaped charge. Code development was concentrated on the PELE code, a multimaterial, multiphase, explicit finite-difference Eulerian code for pool suppression dynamics of a hypothetical loss-of-coolant accident in a nuclear reactor. Activities of the Fluid Dynamics Group were directed toward development of a code to compute the equations of state and transport properties of liquid metals (e.g. Li) and partially ionized dense plasmas, jet stability in the Li reactor system, and the study and problem application of fluid dynamic turbulence theory. 19 figures, 5 tables. (RWR)« less

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor

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

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

At the Atucha-I pressurized heavy water reactor in Argentina, fuel assemblies in the spent fuel pools are stored by suspending them in two vertically stacked layers. This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Since much of the fuel is very old, Cerenkov viewing devices are often not very useful even for the top layer. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 w% {sup 235}U, and has been in operation since 1974, a wide range of burnups and cooling times canmore » exist in any given pool. A spent fuel neutron counting tool consisting of a fission chamber, SFNC, has been used at the site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups to levels up 11,000 MWd/t, the existing signal processing software of the tool was found to fail due to non-linearity of the source term with burnup. A new Graphical User Interface software package based on the LabVIEW platform was developed to predict expected neutron signals covering all ranges of burnups and cooling times and establish maps of expected signals at various pool locations. The algorithm employed in the software uses a set of transfer functions in a 47-energy group structure which are coupled with a 47-energy group neutron source spectrum based on various cooling times and burnups for each of the two enrichment levels. The database of the software consists of these transfer functions for the three different inter-assembly pitches that the fuel is stored in at the site. The transfer functions were developed for a 6 by 6 matrix of fuel assemblies with the detector placed at the center surrounded by four near neighbors, eight next nearest neighbors and so on for the 36 assemblies. These calculations were performed using Monte Carlo radiation transport methods. The basic methodology consisted of starting sources in each of the assemblies and tallying the contribution to the detector by a single neutron in each of the 47 energy groups used. Thus for the single existing symmetric pitch in the pools, where the vertical and horizontal separations are equal, only 6 sets of transfer functions are required. For the two asymmetrical pitches, nine sets of transfer functions are stored. In addition, source spectra at burnups ranging from 4000 to 20000 MWd/t and cooling times up to 40 years are stored. These source terms were established based on CANDU 37-rod fuel that is very similar to the Atucha fuel. Linear interpolation is used by the software for both burnup and cooling time to establish source terms at any intermediate condition. Using the burnup, cooling time and initial enrichment of the surrounding assemblies a set of source strengths in the 47-group structure for each of the 36 assemblies is established and multiplied group-wise with the appropriate transfer function set. The grand total over the 47 groups for all 36 assemblies is the predicted signal at the detector. The software was initially calibrated against a set of typically 5-6 measurements chosen from among the measured data at each level of the six pools and calibration factors were established. The set used for calibration is chosen such that it is fairly representative of the range of spent fuel assembly characteristics present in each level. Once established, these calibration factors can be repeatedly used for verification purposes. Recalibration will be required if the hardware or pool configurations has changed. It will also be required if a long enough time has elapsed since they were established thus making a cooling time correction necessary. The objective of the inspection is to detect missing fuel from one or more nearest neighbors of the detector. During the verification mode of the software, the predicted and measured signals are compared and the inspector is alerted if the difference between the two signals is beyond a set tolerance limit. Based on the uncertainties associated with both the calculations and measurements, a lower limit of the tolerance will be 15% with an upper limit of 20%. For the most part a 20% tolerance limit will be able to detect a missing assembly since in the vast majority of cases the drop in signal due to a single missing nearest neighbor assembly will be in the range 24-27%. The software was benchmarked against an extensive set of measured data taken at the site in 2004. Overall, 326 data points were examined and the prediction of the calibrated software was compared to the measurements within a set tolerance of ±20%. Of these, 283 of the predicted signals representing 87% of the total matched the measured data within ±10%. A further 27 or 8% were in the range of ±10-15% and 8 or 2.5% were in the range of ±15-20%. Thus, 97.5% of the data matched the measurements within the set tolerance limit of 20%, with 95% matching measured data with the lowest allowed tolerance limit of ±15%. The remaining 2.5% had measured signals that were very different from those at locations with very similar surrounding assemblies and the cause of these discrepancies could not be ascertained from the measurement logs. In summary, 97.5% of the predictions matched the measurements within the set 20% tolerance limit providing proof of the robustness of the software. This software package linked to SFNC will be deployed at the site and will enhance the capability of gross defect verification for the whole range of burnup, cooling time and initial enrichments of the spent fuel being discharged into the various pools at the Atucha-I reactor site.« less

Adaptive Neuro-Fuzzy Inference System (ANFIS)-Based Models for Predicting the Weld Bead Width and Depth of Penetration from the Infrared Thermal Image of the Weld Pool

NASA Astrophysics Data System (ADS)

Subashini, L.; Vasudevan, M.

2012-02-01

Type 316 LN stainless steel is the major structural material used in the construction of nuclear reactors. Activated flux tungsten inert gas (A-TIG) welding has been developed to increase the depth of penetration because the depth of penetration achievable in single-pass TIG welding is limited. Real-time monitoring and control of weld processes is gaining importance because of the requirement of remoter welding process technologies. Hence, it is essential to develop computational methodologies based on an adaptive neuro fuzzy inference system (ANFIS) or artificial neural network (ANN) for predicting and controlling the depth of penetration and weld bead width during A-TIG welding of type 316 LN stainless steel. In the current work, A-TIG welding experiments have been carried out on 6-mm-thick plates of 316 LN stainless steel by varying the welding current. During welding, infrared (IR) thermal images of the weld pool have been acquired in real time, and the features have been extracted from the IR thermal images of the weld pool. The welding current values, along with the extracted features such as length, width of the hot spot, thermal area determined from the Gaussian fit, and thermal bead width computed from the first derivative curve were used as inputs, whereas the measured depth of penetration and weld bead width were used as output of the respective models. Accurate ANFIS models have been developed for predicting the depth of penetration and the weld bead width during TIG welding of 6-mm-thick 316 LN stainless steel plates. A good correlation between the measured and predicted values of weld bead width and depth of penetration were observed in the developed models. The performance of the ANFIS models are compared with that of the ANN models.

Preliminary Consideration of the ADS Research in China

NASA Astrophysics Data System (ADS)

Fang, Shouxian; Fu, Shinian

2002-08-01

Power supply is a key issue for China's further economic development. To meet the needs of our economic growth in the next century, the part of nuclear energy in the total newly increased power supply must become larger. However, the present nuclear power stations dominated by the PWR in the world are facing some troubles. Recently, a new concept, called ADS (Accelerator Driven Subcritical system), can avoid these troubles and it is recognized as a most prospective power system for fission energy. So during the early time of nuclear power development in our country, it is worthwhile to exploit this novel idea. In this paper, the ADS research program and a proposed verification facility are described. It consists of an 300MeV/3mA low energy accelerator, a swimming pool reactor and some basic research equipment. Beam physics, such as beam halo formation, in the intense-beam accelerator is also discussed.

Composite space charge density functions for the calculation of gamma sensitivity of self-powered neutron detectors, using Warren's model

NASA Astrophysics Data System (ADS)

Mahant, A. K.; Rao, P. S.; Misra, S. C.

1994-07-01

In the calculational model developed by Warren and Shah for the computation of the gamma sensitivity ( Sγ) it has been observed that the computed Sγ value is quite sensitive to the space charge distribution function assumed for the insulator region and the energy of the gamma photons. The Sγ of SPNDs with Pt, Co and V emitters (manufactured by Thermocoax, France) has been measured at 60Co photon energy and a good correlation between the measured and computed values has been obtained using a composite space charge density function (CSCD), the details of which are presented in this paper. The arguments are extended for evaluating the Sγ values of several SPNDs for which Warren and Shah reported the measured values for a prompt fission gamma spectrum obtained in a swimming pool reactor. These results are also discussed.

TRAC-PF1 code verification with data from the OTIS test facility. [Once-Through Intergral System

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

Childerson, M.T.; Fujita, R.K.

1985-01-01

A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code wasmore » successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer.« less

Application of a Depositional Facies Model to an Acid Mine Drainage Site▿ †

PubMed Central

Brown, Juliana F.; Jones, Daniel S.; Mills, Daniel B.; Macalady, Jennifer L.; Burgos, William D.

2011-01-01

Lower Red Eyes is an acid mine drainage site in Pennsylvania where low-pH Fe(II) oxidation has created a large, terraced iron mound downstream of an anoxic, acidic, metal-rich spring. Aqueous chemistry, mineral precipitates, microbial communities, and laboratory-based Fe(II) oxidation rates for this site were analyzed in the context of a depositional facies model. Depositional facies were defined as pools, terraces, or microterracettes based on cm-scale sediment morphology, irrespective of the distance downstream from the spring. The sediments were composed entirely of Fe precipitates and cemented organic matter. The Fe precipitates were identified as schwertmannite at all locations, regardless of facies. Microbial composition was studied with fluorescence in situ hybridization (FISH) and transitioned from a microaerophilic, Euglena-dominated community at the spring, to a Betaproteobacteria (primarily Ferrovum spp.)-dominated community at the upstream end of the iron mound, to a Gammaproteobacteria (primarily Acidithiobacillus)-dominated community at the downstream end of the iron mound. Microbial community structure was more strongly correlated with pH and geochemical conditions than depositional facies. Intact pieces of terrace and pool sediments from upstream and downstream locations were used in flowthrough laboratory reactors to measure the rate and extent of low-pH Fe(II) oxidation. No change in Fe(II) concentration was observed with 60Co-irradiated sediments or with no-sediment controls, indicating that abiotic Fe(II) oxidation was negligible. Upstream sediments attained lower effluent Fe(II) concentrations compared to downstream sediments, regardless of depositional facies. PMID:21097582

Characterization of cartridge filters from the IEA-R1 Nuclear Reactor

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

NONE

The management of radioactive waste ensures safety to human health and the environment nowadays and for the future, without overwhelming the upcoming generations. The primary characterization of radioactive waste is one of the main steps in the management of radioactive waste. This step permits to choose the best treatment for the radioactive waste before forwarding it to its final disposal. The aim of the present work is the primary characterization of cartridge filters from the IEA-R1 nuclear reactor utilizing gamma-ray spectrometry, and the method of Monte Carlo for calibration. The IEA-R1 is located in the Nuclear and Energy Research Institutemore » (IPEN - CNEN) in the city of Sao Paulo, Brazil. Cartridge filters are used for purification of the cooling water that is pumped through the core of the pool type nuclear research reactors. Once worn out, these filters are replaced and then become radioactive waste. Determination of the radioactive inventory is of paramount importance in the management of such radioactive waste, and one of the main methods for doing so is the gamma-ray spectrometry, which can identify and quantify high energy photon emitters. The technique chosen for the characterization of radioactive waste in the present work is the gamma-ray spectrometry with High purity Germanium (HPGe) detectors. From the energy identified in the experimental spectrum, three radioisotopes were identified in the cartridge filter: {sup 108m}Ag, {sup 110m}Ag, {sup 60}Co. For the estimated activity of the filter, the calibration in efficiency was made utilizing the MCNP4C code of the Monte Carlo method. Such method was chosen because there is no standard source available in the same geometry of the cartridge filter, therefore a simulation had to be developed in order to reach a calibration equation, necessary to estimate the activity of the radioactive waste. The results presented an activity value in the order of MBq for all radioisotopes. (authors)« less

Pile noise experiment in MINERVE reactor to estimate kinetic parameters using various data processing methods

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

Geslot, Benoit; Gruel, Adrien; Pepino, Alexandra

2015-07-01

MINERVE is a two-zone pool type zero power reactor operated by CEA (Cadarache, France). Kinetic parameters of the core (prompt neutron decay constant, delayed neutron fraction, generation time) have been recently measured using various pile noise experimental techniques, namely Feynman-α, Rossi-α and Cohn-α. Results are discussed and compared to each other's. The measurement campaign has been conducted in the framework of a tri-partite collaboration between CEA, SCK.CEN and PSI. Results presented in this paper were obtained thanks to a time-stamping acquisition system developed by CEA. PSI performed simultaneous measurements which are presented in a companion paper. Signals come from twomore » high efficiency fission chambers located in the graphite reflector next to the core driver zone. Experiments were conducted at critical state with a reactor power of 0.2 W. The core integral fission rate is obtained from a calibrated miniature fission chamber located at the center of the core. Other results obtained in two sub-critical configurations will be presented elsewhere. Best estimate delayed neutron fraction comes from the Cohn-α method: 747 ± 15 pcm (1σ). In this case, the prompt decay constant is 79 ± 0.5 s{sup -1} and the generation time is 94.5 ± 0.7 μs. Other methods give consistent results within the confidence intervals. Experimental results are compared to calculated values obtained from a full 3D core modeling with the CEA-developed Monte Carlo code TRIPOLI4.9 associated with its continuous energy JEFF3.1.1-based library. A very good agreement is observed for the calculated delayed neutron fraction (748.7 ± 0.4 pcm at 1σ), that is a difference of -0.3% with the experiment. On the contrary, a 10% discrepancy is observed for the calculated generation time (104.4 ± 0.1 μs at 1σ). (authors)« less

Used Nuclear Fuel: From Liability to Benefit

NASA Astrophysics Data System (ADS)

Orbach, Raymond L.

2011-03-01

Nuclear power has proven safe and reliable, with operating efficiencies in the U.S. exceeding 90%. It provides a carbon-free source of electricity (with about a 10% penalty arising from CO2 released from construction and the fuel cycle). However, used fuel from nuclear reactors is highly toxic and presents a challenge for permanent disposal -- both from technical and policy perspectives. The half-life of the ``bad actors'' is relatively short (of the order of decades) while the very long lived isotopes are relatively benign. At present, spent fuel is stored on-site in cooling ponds. Once the used fuel pools are full, the fuel is moved to dry cask storage on-site. Though the local storage is capable of handling used fuel safely and securely for many decades, the law requires DOE to assume responsibility for the used fuel and remove it from reactor sites. The nuclear industry pays a tithe to support sequestration of used fuel (but not research). However, there is currently no national policy in place to deal with the permanent disposal of nuclear fuel. This administration is opposed to underground storage at Yucca Mountain. There is no national policy for interim storage---removal of spent fuel from reactor sites and storage at a central location. And there is no national policy for liberating the energy contained in used fuel through recycling (separating out the fissionable components for subsequent use as nuclear fuel). A ``Blue Ribbon Commission'' has been formed to consider alternatives, but will not report until 2012. This paper will examine alternatives for used fuel disposition, their drawbacks (e.g. proliferation issues arising from recycling), and their benefits. For recycle options to emerge as a viable technology, research is required to develop cost effective methods for treating used nuclear fuel, with attention to policy as well as technical issues.

Development of an inconel self powered neutron detector for in-core reactor monitoring

NASA Astrophysics Data System (ADS)

Alex, M.; Ghodgaonkar, M. D.

2007-04-01

The paper describes the development and testing of an Inconel600 (2 mm diameter×21 cm long) self-powered neutron detector for in-core neutron monitoring. The detector has 3.5 mm overall diameter and 22 cm length and is integrally coupled to a 12 m long mineral insulated cable. The performance of the detector was compared with cobalt and platinum detectors of similar dimensions. Gamma sensitivity measurements performed at the 60Co irradiation facility in 14 MR/h gamma field showed values of -4.4×10 -18 A/R/h/cm (-9.3×10 -24 A/ γ/cm 2-s/cm), -5.2×10 -18 A/R/h/cm (-1.133×10 -23 A/ γ/cm 2-s/cm) and 34×10 -18 A/R/h/cm (7.14×10 -23 A/ γ/cm 2-s/cm) for the Inconel, Co and Pt detectors, respectively. The detectors together with a miniature gamma ion chamber and fission chamber were tested in the in-core Apsara Swimming Pool type reactor. The ion chambers were used to estimate the neutron and gamma fields. With an effective neutron cross-section of 4b, the Inconel detector has a total sensitivity of 6×10 -23 A/nv/cm while the corresponding sensitivities for the platinum and cobalt detectors were 1.69×10 -22 and 2.64×10 -22 A/nv/cm. The linearity of the detector responses at power levels ranging from 100 to 200 kW was within ±5%. The response of the detectors to reactor scram showed that the prompt response of the Inconel detector was 0.95 while it was 0.7 and 0.95 for the platinum and cobalt self-powered detectors, respectively. The detector was also installed in the horizontal flux unit of 540 MW Pressurised Heavy Water Reactor (PHWR). The neutron flux at the detector location was calculated by Triveni code. The detector response was measured from 0.02% to 0.07% of full power and showed good correlation between power level and detector signals. Long-term tests and the dynamic response of the detector to shut down in PHWR are in progress.

On the possible leakage of ET-RR1 liquid waste tank: hydrological and migration modes studies.

PubMed

Mahmoud, N S; El-Hemamy, S T

2005-03-20

The first Egyptian (ET-RR1) research reactor has been in operation since 1961 at the Egyptian Atomic Energy Authority (EAEA) Inshas site. Therefore, at present, it faces a serious problem due to aging equipment, especially those directly in contact with the environment such as the underground settling tanks of nuclear and radioactive waste. The possible leakage of radionuclides from these aging tanks and their migration to the aquifer was studied using instantaneous release. This study was done based on the geological and hydrological characteristics of the site, which were obtained from the hydrogeological data of 25 wells previously drilled at the site of the reactor[1]. These data were used to calculate the trend of water levels, hydraulic gradient, and formulation of water table maps from 1993-2002. This information was utilized to determine water velocity in the unsaturated zone. Radionuclides released from the settling tank to the aquifer were screened according to the radionuclides that have high migration ability and high activity. The amount of fission and activation products of the burned fuels that contaminated the water content of the reactor pool were considered as 10% of the original spent fuel. The radionuclides considered in this case were H-3, Sr-90, Zr-93, Tc-99, Cd-113, Cs-135, Cs-137, Sm-151, Pu-238, Pu-240, Pu-241, and Am-241. The instantaneous release was analyzed by theoretical calculations, taking into consideration the migration mechanism of the various radionuclides through the soil space between the tank bottom and the aquifer. The migration mechanism through the unsaturated zone was considered depending on soil type, thickness of the unsaturated zone, water velocity, and other factors that are specific for each radionuclide, namely retardation factor, which is the function of the specific distribution coefficient of each radionuclide. This was considered collectively as delay time. Meanwhile, the mechanism of radionuclide migration during their passage in the water body of the aquifer was the main focus of this study. The degree of water pollution in the aquifer at a point of contact with the main water body of Ismailia Canal 1000 m from the reactor site was assessed for the instantaneous release by comparing the results obtained with the regulations of the standard limit of radionuclides in drinking water.

On The Possible Leakage of ET-RR1 Liquid Waste Tank: Hydrological and Migration Modes Studies

PubMed Central

Mahmoud, N. S.; EL-Hemamy, S. T.

2005-01-01

The first Egyptian (ET-RR1) research reactor has been in operation since 1961 at the Egyptian Atomic Energy Authority (EAEA) Inshas site. Therefore, at present, it faces a serious problem due to aging equipment, especially those directly in contact with the environment such as the underground settling tanks of nuclear and radioactive waste. The possible leakage of radionuclides from these aging tanks and their migration to the aquifer was studied using instantaneous release.This study was done based on the geological and hydrological characteristics of the site, which were obtained from the hydrogeological data of 25 wells previously drilled at the site of the reactor[1]. These data were used to calculate the trend of water levels, hydraulic gradient, and formulation of water table maps from 1993–2002. This information was utilized to determine water velocity in the unsaturated zone.Radionuclides released from the settling tank to the aquifer were screened according to the radionuclides that have high migration ability and high activity. The amount of fission and activation products of the burned fuels that contaminated the water content of the reactor pool were considered as 10% of the original spent fuel. The radionuclides considered in this case were H-3, Sr-90, Zr-93, Tc-99, Cd-113, Cs-135, Cs-137, Sm-151, Pu-238, Pu-240, Pu-241, and Am-241.The instantaneous release was analyzed by theoretical calculations, taking into consideration the migration mechanism of the various radionuclides through the soil space between the tank bottom and the aquifer. The migration mechanism through the unsaturated zone was considered depending on soil type, thickness of the unsaturated zone, water velocity, and other factors that are specific for each radionuclide, namely retardation factor, which is the function of the specific distribution coefficient of each radionuclide. This was considered collectively as delay time. Meanwhile, the mechanism of radionuclide migration during their passage in the water body of the aquifer was the main focus of this study.The degree of water pollution in the aquifer at a point of contact with the main water body of Ismailia Canal 1000 m from the reactor site was assessed for the instantaneous release by comparing the results obtained with the regulations of the standard limit of radionuclides in drinking water[2,3]. PMID:15798884

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

Buitrago, Paula A.; Morrill, Mike; Lighty, JoAnn S.

This report presents experimental and modeling mercury oxidation and adsorption data. Fixed-bed and single-particle models of mercury adsorption were developed. The experimental data were obtained with two reactors: a 300-W, methane-fired, tubular, quartz-lined reactor for studying homogeneous oxidation reactions and a fixed-bed reactor, also of quartz, for studying heterogeneous reactions. The latter was attached to the exit of the former to provide realistic combustion gases. The fixed-bed reactor contained one gram of coconut-shell carbon and remained at a temperature of 150°C. All methane, air, SO 2, and halogen species were introduced through the burner to produce a radical pool representativemore » of real combustion systems. A Tekran 2537A Analyzer coupled with a wet conditioning system provided speciated mercury concentrations. At 150°C and in the absence of HCl or HBr, the mercury uptake was about 20%. The addition of 50 ppm HCl caused complete capture of all elemental and oxidized mercury species. In the absence of halogens, SO 2 increased the mercury adsorption efficiency to up to 30 percent. The extent of adsorption decreased with increasing SO 2 concentration when halogens were present. Increasing the HCl concentration to 100 ppm lessened the effect of SO 2. The fixed-bed model incorporates Langmuir adsorption kinetics and was developed to predict adsorption of elemental mercury and the effect of multiple flue gas components. This model neglects intraparticle diffusional resistances and is only applicable to pulverized carbon sorbents. It roughly describes experimental data from the literature. The current version includes the ability to account for competitive adsorption between mercury, SO 2, and NO 2. The single particle model simulates in-flight sorbent capture of elemental mercury. This model was developed to include Langmuir and Freundlich isotherms, rate equations, sorbent feed rate, and intraparticle diffusion. The Freundlich isotherm more accurately described in-flight mercury capture. Using these parameters, very little intraparticle diffusion was evident. Consistent with other data, smaller particles resulted in higher mercury uptake due to available surface area. Therefore, it is important to capture the particle size distribution in the model. At typical full-scale sorbent feed rates, the calculations under-predicted adsorption, suggesting that wall effects can account for as much as 50 percent of the removal, making it an important factor in entrained-mercury adsorption models.« less

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

Buitrago, Paula A.; Morrill, Mike; Lighty, JoAnn S.

This report presents experimental and modeling mercury oxidation and adsorption data. Fixed-bed and single-particle models of mercury adsorption were developed. The experimental data were obtained with two reactors: a 300-W, methane-fired, tubular, quartz-lined reactor for studying homogeneous oxidation reactions and a fixed-bed reactor, also of quartz, for studying heterogeneous reactions. The latter was attached to the exit of the former to provide realistic combustion gases. The fixed-bed reactor contained one gram of coconut-shell carbon and remained at a temperature of 150°C. All methane, air, SO 2, and halogen species were introduced through the burner to produce a radical pool representativemore » of real combustion systems. A Tekran 2537A Analyzer coupled with a wet conditioning system provided speciated mercury concentrations. At 150°C and in the absence of HCl or HBr, the mercury uptake was about 20%. The addition of 50 ppm HCl caused complete capture of all elemental and oxidized mercury species. In the absence of halogens, SO 2 increased the mercury adsorption efficiency to up to 30 percent. The extent of adsorption decreased with increasing SO 2 concentration when halogens were present. Increasing the HCl concentration to 100 ppm lessened the effect of SO 2. The fixed-bed model incorporates Langmuir adsorption kinetics and was developed to predict adsorption of elemental mercury and the effect of multiple flue gas components. This model neglects intraparticle diffusional resistances and is only applicable to pulverized carbon sorbents. It roughly describes experimental data from the literature. The current version includes the ability to account for competitive adsorption between mercury, SO 2, and NO 2. The single particle model simulates in-flight sorbent capture of elemental mercury. This model was developed to include Langmuir and Freundlich isotherms, rate equations, sorbent feed rate, and intraparticle diffusion. The Freundlich isotherm more accurately described in-flight mercury capture. Using these parameters, very little intraparticle diffusion was evident. Consistent with other data, smaller particles resulted in higher mercury uptake due to available surface area. Therefore, it is important to capture the particle size distribution in the model. At typical full-scale sorbent feed rates, the calculations underpredicted adsorption, suggesting that wall effects can account for as much as 50 percent of the removal, making it an important factor in entrained-mercury adsorption models.« less

Absorption of Some Mineral Salts by Root System of Different Woody Species and Accumulation over a Whole Vegetative Cycle; ABSORPTION DE QUELQUES SELS PAR L'APPAREIL RADICULAIRE DE DIFFERENTES ESPECES LIGNEUSES ET ACCUMULATION AU COURS D'UN CYCLE VEGETATIF COMPLET

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

Gagnaire, J.

1963-01-01

The concentration power of plant tissues and the translocation speed of mineral salts vary considerably with the absorbed salt, the botanical species, the considered tissue, and the part of the vegetative cycle. In Grenoble, with Picea excelsa, the true dormance is short and is accompanied by a pre-dormance period and a post dormance period. In the vegetative period, Picea excelsa leaves concentrate less mineral salt than Acer campestris leaves (coefficient 2 for Ca--3 for phosphates) and Populus nigra leaves (coefficient 3 for Ca, coefficient 5 for phosphates). Results of tracer studies are tabulated. (C.H.)

Measurement of Key Pool BOiling Parameters in nanofluids for Nuclerar Applications

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

Bang, In C; Buongiorno, Jdacopo; Hu, Lin-wen

Nanofluids, colloidal dispersions of nanoparticles in a base fluid such as water, can afford very significant Critical Heat Flux (CHF) enhancement. Such engineered fluids potentially could be employed in reactors as advanced coolants in safety systems with significant safety and economic advantages. However, a satisfactory explanation of the CHF enhancement mechanism in nanofluids is lacking. To close this gap, we have identified the important boiling parameters to be measured. These are the properties (e.g., density, viscosity, thermal conductivity, specific heat, vaporization enthalpy, surface tension), hydrodynamic parameters (i.e., bubble size, bubble velocity, departure frequency, hot/dry spot dynamics) and surface conditions (i.e.,more » contact angle, nucleation site density). We have also deployed a pool boiling facility in which many such parameters can be measured. The facility is equipped with a thin indium-tin-oxide heater deposited over a sapphire substrate. An infra-red high-speed camera and an optical probe are used to measure the temperature distribution on the heater and the hydrodynamics above the heater, respectively. The first data generated with this facility already provide some clue on the CHF enhancement mechanism in nanofluids. Specifically, the progression to burnout in a pure fluid (ethanol in this case) is characterized by a smoothly-shaped and steadily-expanding hot spot. By contrast, in the ethanol-based nanofluid the hot spot pulsates and the progression to burnout lasts longer, although the nanofluid CHF is higher than the pure fluid CHF. The presence of a nanoparticle deposition layer on the heater surface seems to enhance wettability and aid hot spot dissipation, thus delaying burnout.« less

Nutrient fluxes and stoichiometry in a large impounded river-bay system

NASA Astrophysics Data System (ADS)

Klump, J. V.; Waples, J. T.; Able, L. M.; Anderson, P. D.; Weckerly, K.; Szmania, D. C.

2003-04-01

Reservoir-induced aging of continental runoff has been shown to an anthropogenically induced global phenomenon with estimates that the mean age of river water reaching the coastal ocean has likely tripled historically. This aging is hypothesized to have a significant biogeochemical impact on land-margin systems by altering flow regimes, net water balances and residence times, reaeration of surface waters, carbon cycling processes, and sediment storage and transport. The Fox-Wolf watershed system contains more than 20 reservoirs, impoundments and lakes on the main stems of the two principal rivers that feed Green Bay and Lake Michigan. Consequently, this hydrologic system can be conceived as functioning as a series of linked biogeochemical reactors which retard flow, retain particles, significantly attenuate the flux of materials into sequential downstream "pools", and both process and repackage nutrients via tightly coupled benthic-pelagic biotic interactions. This successional transformation process results in a poorly understood delivery of nutrients, soils and contaminants from upstream sources to downstream receptors in Green Bay and ultimately -- Lake Michigan. Nutrient reprocessing (defined as the sum of all processes affecting nutrients, i.e. fixation, remineralization, repackaging, sedimentation, etc.) within each pool is hypothesized to be primarily a function of: (1) particle-solute and hydraulic residence times, (2) the quality and quantity of inputs, and (3) the food web structure. Overlaid on these dynamics are very strong seasonal forcing factors, including annual temperature cycles that induce order of magnitude variations in temperature dependent reaction rates, and winter ice cover on the upper pool lakes, reservoirs and Green Bay, that halts run off from the land and reduces within-basin mixing. These short term and seasonal loading dynamics result in considerable temporal stochasticity in the capacity of the biotic component of the ecosystem to assimilate, transform, and attenuate the flux of materials through the land margin system. We report here preliminary results on the nature of elemental riverine fluxes (e.g. carbon, nitrogen, phosphorus, silica), the shift in their composition and stoichiometry as these materials move downstream, and on the role of impoundments as nutrient traps.

Construction of a Liposome Dialyzer for preparation of high-value, small-volume liposome formulations

PubMed Central

Adamala, Katarzyna; Engelhart, Aaron E.; Kamat, Neha P.; Jin, Lin; Szostak, Jack W.

2016-01-01

The liposome dialyzer is a small-volume equilibrium dialysis device, built from commercially available materials, that is designed for rapid exchange of small volumes of an extraliposomal reagent pool against a liposome preparation. The dialyzer is prepared by modification of commercially available dialysis cartridges and consists of a reactor with two 300 µL chambers and a 1.56 cm2 dialysis surface area. The dialyzer is prepared in three stages: 1) disassembly of dialysis cartridges to obtain required parts; 2) assembly of the dialyzer; and 3) sealing the dialyzer with epoxy. Preparation of the dialyser takes about 1.5 h, not including overnight epoxy curing. Each round of dialysis takes 1–24 h, depending on the analyte and membrane employed. We previously used the dialyzer for small-volume nonenzymatic RNA synthesis reactions inside fatty acid vesicles. In this protocol, we demonstrate other applications, including removal of unencapsulated calcein from vesicles, remote loading, and vesicle microscopy. PMID:26020615

Construction of a liposome dialyzer for the preparation of high-value, small-volume liposome formulations.

PubMed

Adamala, Katarzyna; Engelhart, Aaron E; Kamat, Neha P; Jin, Lin; Szostak, Jack W

2015-06-01

The liposome dialyzer is a small-volume equilibrium dialysis device, built from commercially available materials, that is designed for the rapid exchange of small volumes of an extraliposomal reagent pool against a liposome preparation. The dialyzer is prepared by modification of commercially available dialysis cartridges (Slide-A-Lyzer cassettes), and it consists of a reactor with two 300-μl chambers and a 1.56-cm(2) dialysis surface area. The dialyzer is prepared in three stages: (i) disassembling the dialysis cartridges to obtain the required parts, (ii) assembling the dialyzer and (iii) sealing the dialyzer with epoxy. Preparation of the dialyzer takes ∼1.5 h, not including overnight epoxy curing. Each round of dialysis takes 1-24 h, depending on the analyte and membrane used. We previously used the dialyzer for small-volume non-enzymatic RNA synthesis reactions inside fatty acid vesicles. In this protocol, we demonstrate other applications, including removal of unencapsulated calcein from vesicles, remote loading and vesicle microscopy.

Sulfur status in long distance runners

NASA Astrophysics Data System (ADS)

Kovacs, L.; Zamboni, C.; Lourenço, T.; Macedo, D.

2015-07-01

In sports medicine, sulfur plays an important role and its deficiency can cause muscle injury affecting the performance of the athletes. However, its evaluation is unusual in conventional clinical practice. In this study the sulfur levels were determined in Brazilian amateur athlete's blood using Neutron Activation Analyses (NAA) technique. Twenty six male amateur runners, age 18 to 36 years, participated of this study. The athletes had a balanced diet, without multivitamin/mineral supplements. The blood collection was performed at LABEX (Laboratoriode Bioquimica do Exercicio, UNICAMP-SP) and the samples were irradiated for 300 seconds in a pneumatic station in the nuclear reactor (IEA-R1, 3-4.5MW, pool type) at IPEN/CNEN-SP. The results were compared with the control group (subjects of same age but not involved with physical activities) and showed that the sulfur concentration was 44% higher in amateurs athletes than control group. These data can be considered for preparation of balanced diet, as well as contributing for proposing new protocols of clinical evaluation.

ITER Status and Plans

NASA Astrophysics Data System (ADS)

Greenfield, Charles M.

2017-10-01

The US Burning Plasma Organization is pleased to welcome Dr. Bernard Bigot, who will give an update on progress in the ITER Project. Dr. Bigot took over as Director General of the ITER Organization in early 2015 following a distinguished career that included serving as Chairman and CEO of the French Alternative Energies and Atomic Energy Commission and as High Commissioner for ITER in France. During his tenure at ITER the project has moved into high gear, with rapid progress evident on the construction site and preparation of a staged schedule and a research plan leading from where we are today through all the way to full DT operation. In an unprecedented international effort, seven partners ``China, the European Union, India, Japan, Korea, Russia and the United States'' have pooled their financial and scientific resources to build the biggest fusion reactor in history. ITER will open the way to the next step: a demonstration fusion power plant. All DPP attendees are welcome to attend this ITER town meeting.

Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

DOE PAGES

Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; ...

2014-11-01

This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application ofmore » the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.« less

Geochemical and microbiological responses to oxidant introduction into reduced subsurface sediment from the Hanford 300 Area, Washington.

PubMed

Percak-Dennett, Elizabeth M; Roden, Eric E

2014-08-19

Pliocene-aged reduced lacustrine sediment from below a subsurface redox transition zone at the 300 Area of the Hanford site (southeastern Washington) was used in a study of the geochemical response to introduction of oxygen or nitrate in the presence or absence of microbial activity. The sediments contained large quantities of reduced Fe in the form of Fe(II)-bearing phyllosilicates, together with smaller quantities of siderite and pyrite. A loss of ca. 50% of 0.5 M HCl-extractable Fe(II) [5-10 mmol Fe(II) L(-1)] and detectable generation of sulfate (ca. 0.2 mM, equivalent to 10% of the reduced inorganic sulfur pool) occurred in sterile aerobic reactors. In contrast, no systematic loss of Fe(II) or production of sulfate was observed in any of the other oxidant-amended sediment suspensions. Detectable Fe(II) accumulation and sulfate consumption occurred in non-sterile oxidant-free reactors. Together, these results indicate the potential for heterotrophic carbon metabolism in the reduced sediments, consistent with the proliferation of known heterotrophic taxa (e.g., Pseudomonadaceae, Burkholderiaceae, and Clostridiaceae) inferred from 16S rRNA gene pyrosequencing. Microbial carbon oxidation by heterotrophic communities is likely to play an important role in maintaining the redox boundary in situ, i.e., by modulating the impact of downward oxidant transport on Fe/S redox speciation. Diffusion-reaction simulations of oxygen and nitrate consumption coupled to solid-phase organic carbon oxidation indicate that heterotrophic consumption of oxidants could maintain the redox boundary at its current position over millennial time scales.

Modeling of Kerena Emergency Condenser

NASA Astrophysics Data System (ADS)

Bryk, Rafał; Schmidt, Holger; Mull, Thomas; Wagner, Thomas; Ganzmann, Ingo; Herbst, Oliver

2017-12-01

KERENA is an innovative boiling water reactor concept equipped with several passive safety systems. For the experimental verification of performance of the systems and for codes validation, the Integral Test Stand Karlstein (INKA) was built in Karlstein, Germany. The emergency condenser (EC) system transfers heat from the reactor pressure vessel (RPV) to the core flooding pool in case of water level decrease in the RPV. EC is composed of a large number of slightly inclined tubes. During accident conditions, steam enters into the tubes and condenses due to the contact of the tubes with cold water at the secondary side. The condensed water flows then back to the RPV due to gravity. In this paper two approaches for modeling of condensation in slightly inclined tubes are compared and verified against experiments. The first approach is based on the flow regime map. Depending on the regime, heat transfer coefficient is calculated according to specific semi-empirical correlation. The second approach uses a general, fully-empirical correlation. The models are developed with utilization of the object-oriented Modelica language and the open-source OpenModelica environment. The results are compared with data obtained during a large scale integral test, simulating loss of coolant accident performed at Integral Test Stand Karlstein (INKA). The comparison shows a good agreement.Due to the modularity of models, both of them may be used in the future in systems incorporating condensation in horizontal or slightly inclined tubes. Depending on his preferences, the modeller may choose one-equation based approach or more sophisticated model composed of several exchangeable semi-empirical correlations.

Andrei Sakharov

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

None

2006-07-07

Ce discours donné par Mons.Jonauch qui est né en Tchécoslovaquie et a fait ses études à Leningrad, Moscou et Prague, est organisé par le comité Youri Orlov. Le conférencier parle de Andrei Sakharov, ce physicien et homme soviétique qui fit ses études à Moscou, effectua des recherches sur les armes thermonucléaires et entra à l'Académie des Sciences d'URSS en 1953. Il participa à la mise au point de la bombe à hydrogène, mais s'opposa quelques années plus tard à la poursuite des expériences nucléaires. Il créa en 1970 le comité pour la défense des droits de l'homme ce que luimore » valut le prix Nobel de la paix en 1975.« less

Dislocations et propriétés mécaniques des matériaux céramiques: Quelques problèmes

NASA Astrophysics Data System (ADS)

Castaing, J.; Dominguez Rodriguez, A.

1995-11-01

The study of plastic deformation of ceramic materials raised new problems on low temperature dislocation glide and high temperature dislocation climb. Mechanical behaviour can be explained. In this paper, we review some examples related to oxides which are linked to the activity of J. Philibert. L'étude de la déformation plastique de matériaux céramiques monocristallins a donné l'occasion de poser des nouveaux problèmes sur le glissement des dislocations à basse température et sur leur montée à haute température. Le comportement mécanique peut ainsi être expliqué. Nous passons en revue des cas concernant les oxydes dans lesquels J. Philibert a joué un rôle important.

FHR Process Instruments

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

Holcomb, David Eugene

2015-01-01

Fluoride salt-cooled High temperature Reactors (FHRs) are entering into early phase engineering development. Initial candidate technologies have been identified to measure all of the required process variables. The purpose of this paper is to describe the proposed measurement techniques in sufficient detail to enable assessment of the proposed instrumentation suite and to support development of the component technologies. This paper builds upon the instrumentation chapter of the recently published FHR technology development roadmap. Locating instruments outside of the intense core radiation and high-temperature fluoride salt environment significantly decreases their environmental tolerance requirements. Under operating conditions, FHR primary coolant salt ismore » a transparent, low-vapor-pressure liquid. Consequently, FHRs can employ standoff optical measurements from above the salt pool to assess in-vessel conditions. For example, the core outlet temperature can be measured by observing the fuel s blackbody emission. Similarly, the intensity of the core s Cerenkov glow indicates the fission power level. Short-lived activation of the primary coolant provides another means for standoff measurements of process variables. The primary coolant flow and neutron flux can be measured using gamma spectroscopy along the primary coolant piping. FHR operation entails a number of process measurements. Reactor thermal power and core reactivity are the most significant variables for process control. Thermal power can be determined by measuring the primary coolant mass flow rate and temperature rise across the core. The leading candidate technologies for primary coolant temperature measurement are Au-Pt thermocouples and Johnson noise thermometry. Clamp-on ultrasonic flow measurement, that includes high-temperature tolerant standoffs, is a potential coolant flow measurement technique. Also, the salt redox condition will be monitored as an indicator of its corrosiveness. Both electrochemical techniques and optical spectroscopy are candidate fluoride salt redox measurement methods. Coolant level measurement can be performed using radar-level gauges located in standpipes above the reactor vessel. While substantial technical development remains for most of the instruments, industrially compatible instruments based upon proven technology can be reasonably extrapolated from the current state of the art.« less

The alanine detector in BNCT dosimetry: Dose response in thermal and epithermal neutron fields

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

Schmitz, T., E-mail: schmito@uni-mainz.de; Bassler, N.; Blaickner, M.

Purpose: The response of alanine solid state dosimeters to ionizing radiation strongly depends on particle type and energy. Due to nuclear interactions, neutron fields usually also consist of secondary particles such as photons and protons of diverse energies. Various experiments have been carried out in three different neutron beams to explore the alanine dose response behavior and to validate model predictions. Additionally, application in medical neutron fields for boron neutron capture therapy is discussed. Methods: Alanine detectors have been irradiated in the thermal neutron field of the research reactor TRIGA Mainz, Germany, in five experimental conditions, generating different secondary particlemore » spectra. Further irradiations have been made in the epithermal neutron beams at the research reactors FiR 1 in Helsinki, Finland, and Tsing Hua open pool reactor in HsinChu, Taiwan ROC. Readout has been performed with electron spin resonance spectrometry with reference to an absorbed dose standard in a {sup 60}Co gamma ray beam. Absorbed doses and dose components have been calculated using the Monte Carlo codes FLUKA and MCNP. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using the Hansen and Olsen alanine response model. Results: The measured dose response of the alanine detector in the different experiments has been evaluated and compared to model predictions. Therefore, a relative effectiveness has been calculated for each dose component, accounting for its dependence on particle type and energy. Agreement within 5% between model and measurement has been achieved for most irradiated detectors. Significant differences have been observed in response behavior between thermal and epithermal neutron fields, especially regarding dose composition and depth dose curves. The calculated dose components could be verified with the experimental results in the different primary and secondary particle fields. Conclusions: The alanine detector can be used without difficulty in neutron fields. The response has been understood with the model used which includes the relative effectiveness. Results and the corresponding discussion lead to the conclusion that application in neutron fields for medical purpose is limited by its sensitivity but that it is a useful tool as supplement to other detectors and verification of neutron source descriptions.« less

ESBWR response to an extended station blackout/loss of all AC power

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

Barrett, A. J.; Marquino, W.

2012-07-01

U.S. federal regulations require light water cooled nuclear power plants to cope with Station Blackouts for a predetermined amount of time based on design factors for the plant. U.S. regulations define Station Blackout (SBO) as a loss of the offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency AC power system. According to U.S. regulations, typically the coping period for an SBO is 4 hours and can be as long as 16 hours for currently operating BWR plants. Being able to cope with an SBO and loss of all AC power is required by internationalmore » regulators as well. The U.S. licensing basis for the ESBWR is a coping period of 72 hours for an SBO based on U.S. NRC requirements for passive safety plants. In the event of an extended SBO (viz., greater than 72 hours), the ESBWR response shows that the design is able to cope with the event for at least 7 days without AC electrical power or operator action. ESBWR is a Generation III+ reactor design with an array of passive safety systems. The ESBWR primary success path for mitigation of an SBO event is the Isolation Condenser System (ICS). The ICS is a passive, closed loop, safety system that initiates automatically on a loss of power. Upon Station Blackout or loss of all AC power, the ICS begins removing decay heat from the Reactor Pressure Vessel (RPV) by (i) condensing the steam into water in heat exchangers located in pools of water above the containment, and (ii) transferring the decay heat to the atmosphere. The condensed water is then returned by gravity to cool the reactor again. The ICS alone is capable of maintaining the ESBWR in a safe shutdown condition after an SBO for an extended period. The fuel remains covered throughout the SBO event. The ICS is able to remove decay heat from the RPV for at least 7 days and maintains the reactor in a safe shutdown condition. The water level in the RPV remains well above the top of active fuel for the duration of the SBO event. Beyond 7 days, only a few simple actions are needed to cope with the SBO for an indefinite amount of time. The operation of the ICS as the primary success path for mitigation of an SBO, allows for near immediate plant restart once power is restored. (authors)« less

13 CFR 120.611 - Pools backing Pool Certificates.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 13 Business Credit and Assistance 1 2011-01-01 2011-01-01 false Pools backing Pool Certificates. 120.611 Section 120.611 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION BUSINESS LOANS Secondary Market Certificates § 120.611 Pools backing Pool Certificates. (a) Pool characteristics. As set...

13 CFR 120.611 - Pools backing Pool Certificates.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Pools backing Pool Certificates. 120.611 Section 120.611 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION BUSINESS LOANS Secondary Market Certificates § 120.611 Pools backing Pool Certificates. (a) Pool characteristics. As set...

POOL WATER TREATMENT AND COOLING SYSTEM DESCRIPTION DOCUMENT

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

V. King

2000-06-19

The Pool Water Treatment and Cooling System is located in the Waste Handling Building (WHB), and is comprised of various process subsystems designed to support waste handling operations. This system maintains the pool water temperature within an acceptable range, maintains water quality standards that support remote underwater operations and prevent corrosion, detects leakage from the pool liner, provides the capability to remove debris from the pool, controls the pool water level, and helps limit radiological exposure to personnel. The pool structure and liner, pool lighting, and the fuel staging racks in the pool are not within the scope of themore » Pool Water Treatment and Cooling System. Pool water temperature control is accomplished by circulating the pool water through heat exchangers. Adequate circulation and mixing of the pool water is provided to prevent localized thermal hotspots in the pool. Treatment of the pool water is accomplished by a water treatment system that circulates the pool water through filters, and ion exchange units. These water treatment units remove radioactive and non-radioactive particulate and dissolved solids from the water, thereby providing the water clarity needed to conduct waste handling operations. The system also controls pool water chemistry to prevent advanced corrosion of the pool liner, pool components, and fuel assemblies. Removal of radioactivity from the pool water contributes to the project ALARA (as low as is reasonably achievable) goals. A leak detection system is provided to detect and alarm leaks through the pool liner. The pool level control system monitors the water level to ensure that the minimum water level required for adequate radiological shielding is maintained. Through interface with a demineralized water system, adequate makeup is provided to compensate for loss of water inventory through evaporation and waste handling operations. Interface with the Site Radiological Monitoring System provides continuous radiological monitoring of the pool water. The Pool Water Treatment and Cooling System interfaces with the Waste Handling Building System, Site-Generated Radiological Waste Handling System, Site Radiological Monitoring System, Waste Handling Building Electrical System, Site Water System, and the Monitored Geologic Repository Operations Monitoring and Control System.« less

Rank-based pooling for deep convolutional neural networks.

PubMed

Shi, Zenglin; Ye, Yangdong; Wu, Yunpeng

2016-11-01

Pooling is a key mechanism in deep convolutional neural networks (CNNs) which helps to achieve translation invariance. Numerous studies, both empirically and theoretically, show that pooling consistently boosts the performance of the CNNs. The conventional pooling methods are operated on activation values. In this work, we alternatively propose rank-based pooling. It is derived from the observations that ranking list is invariant under changes of activation values in a pooling region, and thus rank-based pooling operation may achieve more robust performance. In addition, the reasonable usage of rank can avoid the scale problems encountered by value-based methods. The novel pooling mechanism can be regarded as an instance of weighted pooling where a weighted sum of activations is used to generate the pooling output. This pooling mechanism can also be realized as rank-based average pooling (RAP), rank-based weighted pooling (RWP) and rank-based stochastic pooling (RSP) according to different weighting strategies. As another major contribution, we present a novel criterion to analyze the discriminant ability of various pooling methods, which is heavily under-researched in machine learning and computer vision community. Experimental results on several image benchmarks show that rank-based pooling outperforms the existing pooling methods in classification performance. We further demonstrate better performance on CIFAR datasets by integrating RSP into Network-in-Network. Copyright © 2016 Elsevier Ltd. All rights reserved.

Interactions between pool geometry and hydraulics

USGS Publications Warehouse

Thompson, Douglas M.; Nelson, Jonathan M.; Wohl, Ellen E.

1998-01-01

An experimental and computational research approach was used to determine interactions between pool geometry and hydraulics. A 20-m-long, 1.8-m-wide flume was used to investigate the effect of four different geometric aspects of pool shape on flow velocity. Plywood sections were used to systematically alter constriction width, pool depth, pool length, and pool exit-slope gradient, each at two separate levels. Using the resulting 16 unique geometries with measured pool velocities in four-way factorial analyses produced an empirical assessment of the role of the four geometric aspects on the pool flow patterns and hence the stability of the pool. To complement the conclusions of these analyses, a two-dimensional computational flow model was used to investigate the relationships between pool geometry and flow patterns over a wider range of conditions. Both experimental and computational results show that constriction and depth effects dominate in the jet section of the pool and that pool length exhibits an increasing effect within the recirculating-eddy system. The pool exit slope appears to force flow reattachment. Pool length controls recirculating-eddy length and vena contracta strength. In turn, the vena contracta and recirculating eddy control velocities throughout the pool.

MELCOR Model of the Spent Fuel Pool of Fukushima Dai-ichi Unit 4

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

Carbajo, Juan J

2012-01-01

Unit 4 of the Fukushima Dai-ichi Nuclear Power Plant suffered a hydrogen explosion at 6:00 am on March 15, 2011, exactly 3.64 days after the earthquake hit the plant and the off-site power was lost. The earthquake occurred on March 11 at 2:47 pm. Since the reactor of this Unit 4 was defueled on November 29, 2010, and all its fuel was stored in the spent fuel pool (SFP4), it was first believed that the explosion was caused by hydrogen generated by the spent fuel, in particular, by the recently discharged core. The hypothetical scenario was: power was lost, coolingmore » to the SFP4 water was lost, pool water heated/boiled, water level decreased, fuel was uncovered, hot Zircaloy reacted with steam, hydrogen was generated and accumulated above the pool, and the explosion occurred. Recent analyses of the radioisotopes present in the water of the SFP4 and underwater video indicated that this scenario did not occur - the fuel in this pool was not damaged and was never uncovered the hydrogen of the explosion was apparently generated in Unit 3 and transported through exhaust ducts that shared the same chimney with Unit 4. This paper will try to answer the following questions: Could that hypothetical scenario in the SFP4 had occurred? Could the spent fuel in the SPF4 generate enough hydrogen to produce the explosion that occurred 3.64 days after the earthquake? Given the magnitude of the explosion, it was estimated that at least 150 kg of hydrogen had to be generated. As part of the investigations of this accident, MELCOR models of the SFP4 were prepared and a series of calculations were completed. The latest version of MELCOR, version 2.1 (Ref. 1), was employed in these calculations. The spent fuel pool option for BWR fuel was selected in MELCOR. The MELCOR model of the SFP4 consists of a total of 1535 fuel assemblies out of which 548 assemblies are from the core defueled on Nov. 29, 2010, 783 assemblies are older assemblies, and 204 are new/fresh assemblies. The total decay heat of the fuel in the pool was, at the time of the accident, 2.284 MWt, of which 1.872 MWt were from the 548 assemblies of the last core discharged and 0.412 MWt were from the older 783 assemblies. These decay heat values were calculated at Oak Ridge National Laboratory using the ORIGEN2.2 code (Ref. 2) - they agree with values reported elsewhere (Ref. 3). The pool dimensions are 9.9 m x 12.2 m x 11.8 m (height), and with the water level at 11.5 m, the pool volume is 1389 m3, of which only 1240 m3 is water, as some volume is taken by the fuel and by the fuel racks. The initial water temperature of the SFP4 was assumed to be 301 K. The fuel racks are made of an aluminum alloy but are modeled in MELCOR with stainless steel and B4C. MELCOR calculations were completed for different initial water levels: 11.5 m (pool almost full, water is only 0.3 m below the top rim), 4.4577 m (top of the racks), 4.2 m, and 4.026 m (top of the active fuel). A calculation was also completed for a rapid loss of water due to a leak at the bottom of the pool, with the fuel rapidly uncovered and oxidized in air. Results of these calculations are shown in the enclosed Table I. The calculation with the initial water level at 11.5 m (full pool) takes 11 days for the water to boil down to the top of the fuel racks, 11.5 days for the fuel to be uncovered, 14.65 days to generate 150 kg of hydrogen and 19 days for the pool to be completely dry. The calculation with the initial water level at 4.4577 m, takes 1.1 days to uncover the fuel and 4.17 days to generate 150 kg of hydrogen. The calculation with the initial water level at 4.02 m takes 3.63 days to generate 150 kg of hydrogen this is exactly the time when the actual explosion occurred in Unit 4. Finally, fuel oxidation in air after the pool drained the water in 20 minutes, generates only 10 kg of hydrogen this is because very little steam is available and Zircaloy (Zr) oxidation with the oxygen of the air does not generate hydrogen. MELCOR calculated water levels and hydrogen generated in the SFP4 as a function of time for initial water levels of 4.457 m, 4.2 m and 4.02 m are shown in Figs. 1 and 2. Water levels increase at the beginning due to the expansion of the water during the heat-up from 301 K to 373 K. Boiling occurs after the water temperature reaches 373 K. The total amount of hydrogen generated is ~2000 kg, this amount includes hydrogen generated from Zr, which is the largest amount (~1580 kg), from stainless steel (~360 kg), and from B4C (~60 kg). In theory, it is possible to generate up to 3.4 kg of hydrogen per assembly (from oxidation of Zr in the fuel cladding and box), or a total of 4,525 kg from the hot 1331 assemblies stored in the SFP4. The hydrogen generated from oxidation of steel and B4C will be additional. So the answers to the questions are YES according to these MELCOR calculations, enough hydrogen (150 kg) could be generated in the SFP4 3.64 days after the earthquake to produce ...« less

Self-formed waterfall plunge pools in homogeneous rock

NASA Astrophysics Data System (ADS)

Scheingross, Joel S.; Lo, Daniel Y.; Lamb, Michael P.

2017-01-01

Waterfalls are ubiquitous, and their upstream propagation can set the pace of landscape evolution, yet no experimental studies have examined waterfall plunge pool erosion in homogeneous rock. We performed laboratory experiments, using synthetic foam as a bedrock simulant, to produce self-formed waterfall plunge pools via particle impact abrasion. Plunge pool vertical incision exceeded lateral erosion by approximately tenfold until pools deepened to the point that the supplied sediment could not be evacuated and deposition armored the pool bedrock floor. Lateral erosion of plunge pool sidewalls continued after sediment deposition, but primarily at the downstream pool wall, which might lead to undermining of the plunge pool lip, sediment evacuation, and continued vertical pool floor incision in natural streams. Undercutting of the upstream pool wall was absent, and our results suggest that vertical drilling of successive plunge pools is a more efficient waterfall retreat mechanism than the classic model of headwall undercutting and collapse in homogeneous rock.

Controls on Filling and Evacuation of Sediment in Waterfall Plunge Pools

NASA Astrophysics Data System (ADS)

Scheingross, J. S.; Lamb, M. P.

2014-12-01

Many waterfalls are characterized by the presence of deep plunge pools that experience periods of sediment fill and evacuation. These cycles of sediment fill are a first order control on the relative magnitude of lateral versus vertical erosion at the base of waterfalls, as vertical incision requires cover-free plunge pools to expose the bedrock floor, while lateral erosion can occur when pools are partially filled and plunge-pool walls are exposed. Currently, there exists no mechanistic model describing sediment transport through waterfall plunge pools, limiting our ability to predict waterfall retreat. To address this knowledge gap, we performed detailed laboratory experiments measuring plunge-pool sediment transport capacity (Qsc_pool) under varying waterfall and plunge-pool geometries, flow hydraulics, and sediment size. Our experimental plunge-pool sediment transport capacity measurements match well with a mechanistic model we developed which combines existing waterfall jet theory with a modified Rouse profile to predict sediment transport capacity as a function of water discharge and suspended sediment concentration at the plunge-pool lip. Comparing the transport capacity of plunge pools to lower gradient portions of rivers (Qsc_river) shows that, for transport limited conditions, plunge pools fill with sediment under modest water discharges when Qsc_river > Qsc_pool, and empty to bedrock under high discharges when Qsc_pool > Qsc_river. These results are consistent with field observations of sand-filled plunge pools with downstream boulder rims, implying filling and excavation of plunge pools over single-storm timescales. Thus, partial filling of waterfall plunge pools may provide a mechanism to promote lateral undercutting and retreat of waterfalls in homogeneous rock in which plunge-pool vertical incision occurs during brief large floods that expose bedrock, whereas lateral erosion may prevail during smaller events.

Emergency cooling system and method

DOEpatents

Oosterkamp, W.J.; Cheung, Y.K.

1994-01-04

An improved emergency cooling system and method are disclosed that may be adapted for incorporation into or use with a nuclear BWR wherein a reactor pressure vessel (RPV) containing a nuclear core and a heat transfer fluid for circulation in a heat transfer relationship with the core is housed within an annular sealed drywell and is fluid communicable therewith for passage thereto in an emergency situation the heat transfer fluid in a gaseous phase and any noncondensibles present in the RPV, an annular sealed wetwell houses the drywell, and a pressure suppression pool of liquid is disposed in the wetwell and is connected to the drywell by submerged vents. The improved emergency cooling system and method has a containment condenser for receiving condensible heat transfer fluid in a gaseous phase and noncondensibles for condensing at least a portion of the heat transfer fluid. The containment condenser has an inlet in fluid communication with the drywell for receiving heat transfer fluid and noncondensibles, a first outlet in fluid communication with the RPV for the return to the RPV of the condensed portion of the heat transfer fluid and a second outlet in fluid communication with the drywell for passage of the noncondensed balance of the heat transfer fluid and the noncondensibles. The noncondensed balance of the heat transfer fluid and the noncondensibles passed to the drywell from the containment condenser are mixed with the heat transfer fluid and the noncondensibles from the RPV for passage into the containment condenser. A water pool is provided in heat transfer relationship with the containment condenser and is thermally communicable in an emergency situation with an environment outside of the drywell and the wetwell for conducting heat transferred from the containment condenser away from the wetwell and the drywell. 5 figs.

Characterizing convective cold pools: Characterizing Convective Cold Pools

DOE PAGES

Drager, Aryeh J.; van den Heever, Susan C.

2017-05-09

Cold pools produced by convective storms play an important role in Earth's climate system. However, a common framework does not exist for objectively identifying convective cold pools in observations and models. The present study investigates convective cold pools within a simulation of tropical continental convection that uses a cloud-resolving model with a coupled land-surface model. Multiple variables are assessed for their potential in identifying convective cold pool boundaries, and a novel technique is developed and tested for identifying and tracking cold pools in numerical model simulations. This algorithm is based on surface rainfall rates and radial gradients in the densitymore » potential temperature field. The algorithm successfully identifies near-surface cold pool boundaries and is able to distinguish between connected cold pools. Once cold pools have been identified and tracked, composites of cold pool evolution are then constructed, and average cold pool properties are investigated. Wet patches are found to develop within the centers of cold pools where the ground has been soaked with rainwater. These wet patches help to maintain cool surface temperatures and reduce cold pool dissipation, which has implications for the development of subsequent convection.« less

Characterizing convective cold pools: Characterizing Convective Cold Pools

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

Drager, Aryeh J.; van den Heever, Susan C.

Cold pools produced by convective storms play an important role in Earth's climate system. However, a common framework does not exist for objectively identifying convective cold pools in observations and models. The present study investigates convective cold pools within a simulation of tropical continental convection that uses a cloud-resolving model with a coupled land-surface model. Multiple variables are assessed for their potential in identifying convective cold pool boundaries, and a novel technique is developed and tested for identifying and tracking cold pools in numerical model simulations. This algorithm is based on surface rainfall rates and radial gradients in the densitymore » potential temperature field. The algorithm successfully identifies near-surface cold pool boundaries and is able to distinguish between connected cold pools. Once cold pools have been identified and tracked, composites of cold pool evolution are then constructed, and average cold pool properties are investigated. Wet patches are found to develop within the centers of cold pools where the ground has been soaked with rainwater. These wet patches help to maintain cool surface temperatures and reduce cold pool dissipation, which has implications for the development of subsequent convection.« less

The spatial variability of water chemistry and DOC in bog pools: the importance of slope position, diurnal turnover and pool type

NASA Astrophysics Data System (ADS)

Holden, Joseph; Turner, Ed; Baird, Andy; Beadle, Jeannie; Billett, Mike; Brown, Lee; Chapman, Pippa; Dinsmore, Kerry; Dooling, Gemma; Grayson, Richard; Moody, Catherine; Gee, Clare

2017-04-01

We have previously shown that marine influence is an important factor controlling regional variability of pool water chemistry in blanket peatlands. Here we examine within-site controls on pool water chemistry. We surveyed natural and artificial (restoration sites) bog pools at blanket peatland sites in northern Scotland and Sweden. DOC, pH, conductivity, dissolved oxygen, temperature, cations, anions and absorbance spectra from 220-750nm were sampled. We sampled changes over time but also conducted intensive spatial surveys within individual pools and between pools on the same sampling days at individual study sites. Artificial pools had significantly greater DOC concentrations and different spectral absorbance characteristics when compared to natural pools at all sites studied. Within-pool variability in water chemistry tended to be small, even for very large pools ( 400 m2), except where pools had a layer of loose, mobile detritus on their beds. In these instances rapid changes took place between the overlying water column and the mobile sediment layer wherein dissolved oxygen concentrations dropped from values of around 12-10 mg/L to values less than 0.5 mg/L over just 2-3 cm of the depth profile. Such strong contrasts were not observed for pools which had a hard peat floor and which lacked a significant detritus layer. Strong diurnal turnover occurred within the pools on summer days, including within small, shallow pools (e.g. < 30 cm deep, 1 m2 area). For many pools on these summer days there was an evening spike in dissolved oxygen concentrations which originated at the surface and was then cycled downwards as the pool surface waters cooled. Slope location was a significant control on several pool water chemistry variables including pH and DOC concentration with accumulation (higher concentrations) in pools that were located further downslope in both natural and artificial pool systems. These processes have important implications for our interpretation of water chemistry and gas flux data from pool systems, how we design our sampling strategies and how we upscale results.

Animal and vegetation patterns in natural and man-made bog pools: implications for restoration

USGS Publications Warehouse

Mazerolle, M.J.; Poulin, M.; Lavoie, C.; Rochefort, L.; Desrochers, A.; Drolet, B.

2006-01-01

1. Peatlands have suffered great losses following drainage for agriculture, forestry, urbanisation, or peat mining, near inhabited areas. We evaluated the faunal and vegetation patterns after restoration of a peatland formerly mined for peat. We assessed whether bog pools created during restoration are similar to natural bog pools in terms of water chemistry, vegetation structure and composition, as well as amphibian and arthropod occurrence patterns. 2. Both avian species richness and peatland vegetation cover at the site increased following restoration. Within bog pools, however, the vegetation composition differed between natural and man-made pools. The cover of low shrubs, Sphagnum moss, submerged, emergent and floating vegetation in man-made pools was lower than in natural pools, whereas pH was higher than in typical bog pools. Dominant plant species also differed between man-made and natural pools. 3. Amphibian tadpoles, juveniles and adults occurred more often in man-made pools than natural bog pools. Although some arthropods, including Coleoptera bog specialists, readily colonised the pools, their abundance was two to 26 times lower than in natural bog pools. Plant introduction in bog pools, at the stocking densities we applied, had no effect on the occurrence of most groups. 4. We conclude that our restoration efforts were partially successful. Peatland-wide vegetation patterns following restoration mimicked those of natural peatlands, but 4 years were not sufficient for man-made pools to fully emulate the characteristics of natural bog pools.

Environmental controls of C, N and P biogeochemistry in peatland pools.

PubMed

Arsenault, Julien; Talbot, Julie; Moore, Tim R

2018-08-01

Pools are common in northern peatlands but studies have seldom focused on their nutrient biogeochemistry, especially in relation to their morphological characteristics and through seasons. We determined the environmental characteristics controlling carbon (C), nitrogen (N) and phosphorus (P) biogeochemistry in pools and assessed their evolution over the course of the 2016 growing season in a subboreal ombrotrophic peatland of eastern Canada. We showed that water chemistry variations in 62 pools were significantly explained by depth (81.9%) and the surrounding vegetation type (14.8%), but not by pool area or shape. Shallow pools had larger dissolved organic carbon (DOC) and total nitrogen (TN) concentrations and lower pH than deep pools, while pools surrounded by coniferous trees had more recalcitrant DOC than pools where vegetation was dominated by mosses. The influence of depth on pool biogeochemistry was confirmed by the seasonal survey of pools of different sizes with 47.1% of the variation in pool water chemistry over time significantly explained. Of this, 67.3% was explained by the interaction between time and pool size and 32.7% by pool size alone. P concentrations were small in all pools all summer long and combined with high N:P ratios, are indicative of P-limitation. Our results show that pool biogeochemistry is influenced by internal processes and highlight the spatial and temporal heterogeneity of nutrient biogeochemistry in ombrotrophic peatlands. Copyright © 2018 Elsevier B.V. All rights reserved.

75 FR 19964 - Notice of Staff Attendance at Southwest Power Pool Regional State Committee Meeting and Southwest...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-16

... Power Pool, Inc. Docket No. ER10-696, Southwest Power Pool, Inc. Docket No. ER10-697, Southwest Power Pool, Inc. Docket No. ER10-698, Southwest Power Pool, Inc. Docket No. ER10-700, Southwest Power Pool...

Paralysie flasque en début de grossesse: penser à l'hypokaliémie due aux vomissements gravidiques, à propos de deux observations dans un pays en voie de développement

PubMed Central

Fall, Maouly

2015-01-01

Les vomissements gravidiques peuvent être responsables de rares complications neuromusculaires mais graves notamment la paralysie hypokaliémique. Nous rapportons des observations de deux jeunes femmes d'origine guinéenne, sans histoires familiales ni antécédents particuliers, admises pour paralysie flasque des quatre membres dans un contexte de vomissements incoercibles en début de grossesse. Le bilan retrouvait une hypokaliémie majeure associée à quelques troubles électro-cardiographiques. L'apport de potassium par voie parentérale avait permis une récupération motrice totale. La paralysie hypokaliémique est une complication rare des vomissements gravidiques. Une supplémentation potassique prudente avec une surveillance électro-cardiographique et biologique permet une disparition spectaculaire des troubles neuromusculaires. PMID:26327956

ÉTUDE de la Capture ÉLECTRONIQUE dans la DÉSINTÉGRATION du Nuclide 23Na

NASA Astrophysics Data System (ADS)

Charpak, G.

L'étude de 22Na est faite avec un compteur Geiger 4π. On met en évidence l'émission d'un rayonnement de très basse énergle, indépendant des rayons β+, complètement absorbé dans un film de quelques microgrammes par centimètre carré d'aluminium on de matière plastique LC 600 et attribué aux électrons Auger d'énergie maximum 0,85 keV, qui suivent la capture électronique. En raison du très faible parcours de ces électrons , nous sommes amené à discuter particulièrement une méthode simple de préparation de sources radloactives minces et uniformes. Nous obtenons la valeta du rapport [ Capture Λ/Emission β+ = (6,5±0,9) pour 100.

Beta decay heat following U-235, U-238 and Pu-239 neutron fission

NASA Astrophysics Data System (ADS)

Li, Shengjie

1997-09-01

This is an experimental study of beta-particle decay heat from 235U, 239Pu and 238U aggregate fission products over delay times 0.4-40,000 seconds. The experimental results below 2s for 235U and 239Pu, and below 20s for 238U, are the first such results reported. The experiments were conducted at the UMASS Lowell 5.5-MV Van de Graaff accelerator and 1-MW swimming-pool research reactor. Thermalized neutrons from the 7Li(p,n)7Be reaction induced fission in 238U and 239Pu, and fast neutrons produced in the reactor initiated fission in 238U. A helium-jet/tape-transport system rapidly transferred fission fragments from a fission chamber to a low background counting area. Delay times after fission were selected by varying the tape speed or the position of the spray point relative to the beta spectrometer that employed a thin-scintillator-disk gating technique to separate beta-particles from accompanying gamma-rays. Beta and gamma sources were both used in energy calibration. Based on low-energy(<1 MeV) internal-conversion electron studies, a set of trial responses for the spectrometer was established and spanned electron energies 0-10 MeV. Measured beta spectra were unfolded for their energy distributions by the program FERD, and then compared to other measurements and summation calculations based on ENDF/B-VI fission-product data performed on the LANL Cray computer. Measurements of the beta activity as a function of decay time furnished a relative normalization. Results for the beta decay heat are presented and compared with other experimental data and the summation calculations.

Rapid depressurization event analysis in BWR/6 using RELAP5 and contain

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

Mueftueoglu, A.K.; Feltus, M.A.

1995-09-01

Noncondensable gases may become dissolved in Boiling Water Reactor (BWR) water level instrumentation during normal operations. Any dissolved noncondensable gases inside these water columns may come out of solution during rapid depressurization events, and displace water from the reference leg piping resulting in a false high level. These water level errors may cause a delay or failure in actuation, or premature shutdown of the Emergency Core Cooling System. (ECCS). If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response and othermore » signals for automatic actuation such as high drywell pressure. It is also important to determine the effect of the level signal on ECCS operation after it is being actuated. The objective of this study is to determine the detailed coupled containment/NSSS response during this rapid depressurization events in BWR/6. The selected scenarios involve: (a) inadvertent opening of all ADS valves, (b) design basis (DB) large break loss of coolant accident (LOCA), and (c) main steam line break (MSLB). The transient behaviors are evaluated in terms of: (a) vessel pressure and collapsed water level response, (b) specific transient boundary conditions, (e.g., scram, MSIV closure timing, feedwater flow, and break blowdown rates), (c) ECCS initiation timing, (d) impact of operator actions, (e) whether indications besides low-low water level were available. The results of the analysis had shown that there would be signals to actuate ECCS other than low reactor level, such as high drywell pressure, low vessel pressure, high suppression pool temperature, and that the plant operators would have significant indications to actuate ECCS.« less

In-vessel coolability and retention of a core melt

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

Theofanous, T.G.; Liu, C.; Additon, S.

1997-02-01

The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. The technical treatment in this assessment includes: (a) new data on energy flow from either volumetrically heated pools or non-heated layers on top, boiling and critical heat flux in inverted, curved geometries, emissivity of molten (superheated) samples of steel, andmore » chemical reactivity proof tests, (b) a simple but accurate mathematical formulation that allows prediction of thermal loads by means of convenient hand calculations, (c) a detailed model programmed on the computer to sample input parameters over the uncertainty ranges, and to produce probability distributions of thermal loads and margins for departure from nucleate boiling at each angular position on the lower head, and (d) detailed structural evaluations that demonstrate that departure from nucleate boiling is a necessary and sufficient criterion for failure. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is {open_quotes}physically unreasonable.{close_quotes} Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings.« less

HIGH-FIDELITY SIMULATION-DRIVEN MODEL DEVELOPMENT FOR COARSE-GRAINED COMPUTATIONAL FLUID DYNAMICS

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

Hanna, Botros N.; Dinh, Nam T.; Bolotnov, Igor A.

Nuclear reactor safety analysis requires identifying various credible accident scenarios and determining their consequences. For a full-scale nuclear power plant system behavior, it is impossible to obtain sufficient experimental data for a broad range of risk-significant accident scenarios. In single-phase flow convective problems, Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) can provide us with high fidelity results when physical data are unavailable. However, these methods are computationally expensive and cannot be afforded for simulation of long transient scenarios in nuclear accidents despite extraordinary advances in high performance scientific computing over the past decades. The major issue is themore » inability to make the transient computation parallel, thus making number of time steps required in high-fidelity methods unaffordable for long transients. In this work, we propose to apply a high fidelity simulation-driven approach to model sub-grid scale (SGS) effect in Coarse Grained Computational Fluid Dynamics CG-CFD. This approach aims to develop a statistical surrogate model instead of the deterministic SGS model. We chose to start with a turbulent natural convection case with volumetric heating in a horizontal fluid layer with a rigid, insulated lower boundary and isothermal (cold) upper boundary. This scenario of unstable stratification is relevant to turbulent natural convection in a molten corium pool during a severe nuclear reactor accident, as well as in containment mixing and passive cooling. The presented approach demonstrates how to create a correction for the CG-CFD solution by modifying the energy balance equation. A global correction for the temperature equation proves to achieve a significant improvement to the prediction of steady state temperature distribution through the fluid layer.« less

10 CFR 36.53 - Operating and emergency procedures.

Code of Federal Regulations, 2011 CFR

2011-01-01

... pool, or another alarmed area; (9) Natural phenomena, including an earthquake, a tornado, flooding, or...) Monitoring pool water for contamination while the water is in the pool and before release of pool water to... or pool monitor; (4) Detection of leaking sources, pool contamination, or alarm caused by...

Swimming Pools.

ERIC Educational Resources Information Center

Ministry of Housing and Local Government, London (England).

Technical and engineering data are set forth on the design and construction of swimming pools. Consideration is given to site selection, pool construction, the comparative merits of combining open air and enclosed pools, and alternative uses of the pool. Guidelines are presented regarding--(1) pool size and use, (2) locker and changing rooms, (3)…

Swimming Pools. A Guide to Their Planning, Design and Operation.

ERIC Educational Resources Information Center

Gabrielsen, M. Alexander, Ed.

Information is presented regarding all phases of swimming pool development and operation from earliest planning considerations to final programing. This comprehensive book covers--(1) the steps involved in planning a pool, (2) designing the pool, (3) water circulation, filtration, and treatment, (4) community pools, school and agency pools, and…

13 CFR 120.1709 - Transfers of Pool Certificates.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 13 Business Credit and Assistance 1 2012-01-01 2012-01-01 false Transfers of Pool Certificates... Transfers of Pool Certificates. (a) Transfer of Pool Certificates. A Pool Certificate is transferable. A transfer of a Pool Certificate must comply with Article 8 of the Uniform Commercial Code of the State of...

13 CFR 120.1709 - Transfers of Pool Certificates.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Transfers of Pool Certificates... Transfers of Pool Certificates. (a) Transfer of Pool Certificates. A Pool Certificate is transferable. A transfer of a Pool Certificate must comply with Article 8 of the Uniform Commercial Code of the State of...

13 CFR 120.1709 - Transfers of Pool Certificates.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 13 Business Credit and Assistance 1 2014-01-01 2014-01-01 false Transfers of Pool Certificates... Transfers of Pool Certificates. (a) Transfer of Pool Certificates. A Pool Certificate is transferable. A transfer of a Pool Certificate must comply with Article 8 of the Uniform Commercial Code of the State of...

13 CFR 120.1709 - Transfers of Pool Certificates.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 13 Business Credit and Assistance 1 2011-01-01 2011-01-01 false Transfers of Pool Certificates... Transfers of Pool Certificates. (a) Transfer of Pool Certificates. A Pool Certificate is transferable. A transfer of a Pool Certificate must comply with Article 8 of the Uniform Commercial Code of the State of...

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

Barker, J. Jr.; Tenenbaum, B.; Woolf, F.

This paper focuses on the governance and regulation of power pools outside the United States. The current governance and regulatory arrangements for four power pools, as developed in pool documents and government regulations and laws, are compared and contrasted. The power pools analyzed are located in England and Wales, Australia, Canada, and Scandinavia. Topics discussed in relation to these pools are the effects of structure on governance, how each pool has dealt with a number of basic governance decisions, how the pools monitor the markets, ways in which regulators and other institutions control pools, and self-governance issues.

Nuclear fuel electrorefiner

DOEpatents

Ahluwalia, Rajesh K.; Hua, Thanh Q.

2004-02-10

The present invention relates to a nuclear fuel electrorefiner having a vessel containing a molten electrolyte pool floating on top of a cadmium pool. An anodic fuel dissolution basket and a high-efficiency cathode are suspended in the molten electrolyte pool. A shroud surrounds the fuel dissolution basket and the shroud is positioned so as to separate the electrolyte pool into an isolated electrolyte pool within the shroud and a bulk electrolyte pool outside the shroud. In operation, unwanted noble-metal fission products migrate downward into the cadmium pool and form precipitates where they are removed by a filter and separator assembly. Uranium values are transported by the cadmium pool from the isolated electrolyte pool to the bulk electrolyte pool, and then pass to the high-efficiency cathode where they are electrolytically deposited thereto.

10 CFR 36.63 - Pool water purity.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Pool water purity. 36.63 Section 36.63 Energy NUCLEAR... § 36.63 Pool water purity. (a) Pool water purification system must be run sufficiently to maintain the conductivity of the pool water below 20 microsiemens per centimeter under normal circumstances. If pool water...

10 CFR 36.63 - Pool water purity.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Pool water purity. 36.63 Section 36.63 Energy NUCLEAR... § 36.63 Pool water purity. (a) Pool water purification system must be run sufficiently to maintain the conductivity of the pool water below 20 microsiemens per centimeter under normal circumstances. If pool water...

10 CFR 36.63 - Pool water purity.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Pool water purity. 36.63 Section 36.63 Energy NUCLEAR... § 36.63 Pool water purity. (a) Pool water purification system must be run sufficiently to maintain the conductivity of the pool water below 20 microsiemens per centimeter under normal circumstances. If pool water...

10 CFR 36.63 - Pool water purity.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Pool water purity. 36.63 Section 36.63 Energy NUCLEAR... § 36.63 Pool water purity. (a) Pool water purification system must be run sufficiently to maintain the conductivity of the pool water below 20 microsiemens per centimeter under normal circumstances. If pool water...

10 CFR 36.63 - Pool water purity.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Pool water purity. 36.63 Section 36.63 Energy NUCLEAR... § 36.63 Pool water purity. (a) Pool water purification system must be run sufficiently to maintain the conductivity of the pool water below 20 microsiemens per centimeter under normal circumstances. If pool water...

Improving Pool Fencing Legislation in Queensland, Australia: Attitudes and Impact on Child Drowning Fatalities

PubMed Central

2017-01-01

Four-sided, non-climbable pool fencing is an effective strategy for preventing children from drowning in home swimming pools. In 2009, the Queensland Government introduced legislation to improve the effectiveness of pool fencing. This study explores community attitudes towards the effectiveness of these legislative changes and examines child (<5 years) drowning deaths in pools. Data from the 2011 Queensland Computer-Assisted Telephone Interviewing (CATI) Social Survey include results from questions related to pool ownership and pool fencing legislation. Fatal child drowning cases between 1 January 2005 and 31 December 2015 were sourced from coronial data. Of the 1263 respondents, 26/100 households had a pool. A total of 58% believed tightening legislation would be effective in reducing child drowning deaths. Pool owners were more likely to doubt the effectiveness of legislation (p < 0.001) when compared to non-pool owners. Perceptions of effectiveness did not differ by presence of children under the age of five. There were 46 children who drowned in Queensland home pools (7.8/100,000 pools with children residing in the residence/annum) between 2005 and 2015. While pool owners were less likely to think that tightening the legislation would be effective, the number of children drowning in home swimming pools declined over the study period. Drowning prevention agencies have more work to do to ensure that the most vulnerable (young children in houses with swimming pools) are protected. PMID:29186787

New England salt marsh pools: A quantitative analysis of geomorphic and geographic features

USGS Publications Warehouse

Adamowicz, S.C.; Roman, C.T.

2005-01-01

New England salt marsh pools provide important wildlife habitat and are the object of on-going salt marsh restoration projects; however, they have not been quantified in terms of their basic geomorphic and geographic traits. An examination of 32 ditched and unditched salt marshes from the Connecticut shore of Long Island Sound to southern Maine, USA, revealed that pools from ditched and unditched marshes had similar average sizes of about 200 m2, averaged 29 cm in depth, and were located about 11 m from the nearest tidal flow. Unditched marshes had 3 times the density (13 pools/ha), 2.5 times the pool coverage (83 m pool/km transect), and 4 times the total pool surface area per hectare (913 m2 pool/ha salt marsh) of ditched sites. Linear regression analysis demonstrated that an increasing density of ditches (m ditch/ha salt marsh) was negatively correlated with pool density and total pool surface area per hectare. Creek density was positively correlated with these variables. Thus, it was not the mere presence of drainage channels that were associated with low numbers of pools, but their type (ditch versus creek) and abundance. Tidal range was not correlated with pool density or total pool surface area, while marsh latitude had only a weak relationship to total pool surface area per hectare. Pools should be incorporated into salt marsh restoration planning, and the parameters quantified here may be used as initial design targets.

Classification of upper Mississippi River pools based on contiguous aquatic/geomorphic habitats

USGS Publications Warehouse

Koel, Todd M.

2001-01-01

Navigation pools of the upper Mississippi River (UMR) vary greatly in terms of available contiguous aquatic/geomorphic habitats. These habitats are critical for the biotic diversity and overall productivity of the floodplain corridor of each pool. In this study, similarities among pools 4-26 and an open river reach (river kilometer 47-129) of the UMR were determined from multivariate analysis of eleven habitat types that were hydrologically-contiguous (non-leveed). Isolated floodplain habitats were not included in final analyses because this isolation limits their contribution to overall riverine productivity, in part due to a lack of hydrological connectivity to the main channel during the flood pulse. Cluster analysis based on simple Euclidean distance was used to produce two major pool groups and five pool subgroups. Important habitat variables in defining pool groups, as interpreted from principal components analysis (PCA) axis 1, were contiguous floodplain shallow aquatic area and contiguous impounded area. The habitat variable most important in defining pool subgroups, as interpreted from PCA axis 2, was tertiary channel. Most notably, pool 6 was more similar to pools 14-24 than other upper pools, and pools 19 and 25 were more similar to pools 4-13 than other lower pools. These results were quite different from those of two previous investigators, primarily because only areas of non-isolated aquatic habitat were considered.

Pool spacing in forest channels

Treesearch

David R. Montgomery; John M. Buffington; Richard D. Smith; Kevin M. Schmidt; George Pess

1995-01-01

Field surveys of stream channels in forested mountain drainage basins in southeast Alaska and Washington reveal that pool spacing depends on large woody debris (LWD) loading and channel type, slope, and width. Mean pool spacing in pool-riffle, plane-bed, and forced pool-riffle channels systematically decreases from greater than 13 channel widths per pool to less than 1...

76 FR 7528 - Endangered and Threatened Wildlife and Plants; 90-Day Finding on a Petition To Revise Critical...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-02-10

... Revise Critical Habitat for Vernal Pool Fairy Shrimp and Vernal Pool Tadpole Shrimp AGENCY: Fish and... critical habitat for vernal pool fairy shrimp (Branchinecta lynchi) and vernal pool tadpole shrimp... revision of the critical habitat for vernal pool fairy shrimp and vernal pool tadpole shrimp may be...

An introduction to mid-Atlantic seasonal pools

USGS Publications Warehouse

Brown, L.J.; Jung, R.E.

2005-01-01

Seasonal pools, also known as vernal ponds, provide important ecological services to the mid-Atlantic region. This publication serves as an introduction to seasonal pool ecology and management; it also provides tools for exploring seasonal pools, including a full-color field guide to wildlife. Seasonal pools are defined as having four distinctive features: surface water isolation, periodic drying, small size and shallow depth, and support of a characteristic biological community. Seasonal pools experience regular drying that excludes populations of predatory fish. Thus, pools in the mid-Atlantic region provide critical breeding habitat for amphibian and invertebrate species (e.g., spotted salamander (Ambystoma maculatum), wood frog (Rana sylvatica), and fairy shrimp (Order Anostraca)) that would be at increased risk of predation in more permanent waters. The distinctive features of seasonal pools also make them vulnerable to human disturbance. In the mid-Atlantic region, land-use changes pose the greatest challenges to seasonal pool conservation. Seasonal pools are threatened by direct loss (e.g., filling or draining of the pool) as well as by destruction and fragmentation of adjoining terrestrial habitat. Many of the species that depend on seasonal pools for breeding spend the majority of their lives in the surrounding lands that extend a radius of 1000 feet or more from the pools; these vital habitats are being transected by roads and converted to other land uses. Other threats to seasonal pools include biological introductions and removals, mosquito control practices, amphibian diseases, atmospheric deposition, and climate change. The authors recommend a three-pronged strategy for seasonal pool conservation and management in the mid-Atlantic region: education and research, inventory and monitoring of seasonal pools, and landscape-level planning and management.

Experimental study on the stability and failure of individual step-pool

NASA Astrophysics Data System (ADS)

Zhang, Chendi; Xu, Mengzhen; Hassan, Marwan A.; Chartrand, Shawn M.; Wang, Zhaoyin

2018-06-01

Step-pools are one of the most common bedforms in mountain streams, the stability and failure of which play a significant role for riverbed stability and fluvial processes. Given this importance, flume experiments were performed with a manually constructed step-pool model. The experiments were carried out with a constant flow rate to study features of step-pool stability as well as failure mechanisms. The results demonstrate that motion of the keystone grain (KS) caused 90% of the total failure events. The pool reached its maximum depth and either exhibited relative stability for a period before step failure, which was called the stable phase, or the pool collapsed before its full development. The critical scour depth for the pool increased linearly with discharge until the trend was interrupted by step failure. Variability of the stable phase duration ranged by one order of magnitude, whereas variability of pool scour depth was constrained within 50%. Step adjustment was detected in almost all of the runs with step-pool failure and was one or two orders smaller than the diameter of the step stones. Two discharge regimes for step-pool failure were revealed: one regime captures threshold conditions and frames possible step-pool failure, whereas the second regime captures step-pool failure conditions and is the discharge of an exceptional event. In the transitional stage between the two discharge regimes, pool and step adjustment magnitude displayed relatively large variabilities, which resulted in feedbacks that extended the duration of step-pool stability. Step adjustment, which was a type of structural deformation, increased significantly before step failure. As a result, we consider step deformation as the direct explanation to step-pool failure rather than pool scour, which displayed relative stability during step deformations in our experiments.

Dominant factors in controlling marine gas pools in South China

USGS Publications Warehouse

Xu, S.; Watney, W.L.

2007-01-01

In marine strata from Sinian to Middle Triassic in South China, there develop four sets of regional and six sets of local source rocks, and ten sets of reservoir rocks. The occurrence of four main formation periods in association with five main reconstruction periods, results in a secondary origin for the most marine gas pools in South China. To improve the understanding of marine gas pools in South China with severely deformed geological background, the dominant control factors are discussed in this paper. The fluid sources, including the gas cracked from crude oil, the gas dissolved in water, the gas of inorganic origin, hydrocarbons generated during the second phase, and the mixed pool fluid source, were the most significant control factors of the types and the development stage of pools. The period of the pool formation and the reconstruction controlled the pool evolution and the distribution on a regional scale. Owing to the multiple periods of the pool formation and the reconstruction, the distribution of marine gas pools was complex both in space and in time, and the gas in the pools is heterogeneous. Pool elements, such as preservation conditions, traps and migration paths, and reservoir rocks and facies, also served as important control factors to marine gas pools in South China. Especially, the preservation conditions played a key role in maintaining marine oil and gas accumulations on a regional or local scale. According to several dominant control factors of a pool, the pool-controlling model can be constructed. As an example, the pool-controlling model of Sinian gas pool in Weiyuan gas field in Sichuan basin was summed up. ?? Higher Education Press and Springer-Verlag 2007.

Timely topics on spent fuel storage

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

Selin, I.

1994-12-31

The history of spent fuel management in this country has taken several turns, with a final resolution still out of reach. Several repository programs started, stalled ans stopped. The latest effort at Yucca Mountain is progressing but, at best, is years from the early phases of licensing, much less the actual underground disposal of spent fuel. A monitored retrieval storage [MRS] facility was expected to start accepting commercial spent fuel beginning in 1998, but no such facility is clearly on the horizon. All of these recent developments changed the circumstances that we face in spent fuel management. The obvious conclusionmore » is that an increasing number of plants, both operating and permanently shut-down reactors, will have to provide for additional spent fuel storage on-site for a longer period than originally planned, and even after plant decommissioning, prudence requires that provision be made for continual, stand-alone, on-site storage. After pool capacity is reached, most utilities opt for some sort of dry storage. But the dry storage option has triggered an unprecedented amount of local opposition at many sites, further taxing NRC and industry resources.« less

Optimization of hydrolysis and volatile fatty acids production from sugarcane filter cake: Effects of urea supplementation and sodium hydroxide pretreatment.

PubMed

Janke, Leandro; Leite, Athaydes; Batista, Karla; Weinrich, Sören; Sträuber, Heike; Nikolausz, Marcell; Nelles, Michael; Stinner, Walter

2016-01-01

Different methods for optimization the anaerobic digestion (AD) of sugarcane filter cake (FC) with a special focus on volatile fatty acids (VFA) production were studied. Sodium hydroxide (NaOH) pretreatment at different concentrations was investigated in batch experiments and the cumulative methane yields fitted to a dual-pool two-step model to provide an initial assessment on AD. The effects of nitrogen supplementation in form of urea and NaOH pretreatment for improved VFA production were evaluated in a semi-continuously operated reactor as well. The results indicated that higher NaOH concentrations during pretreatment accelerated the AD process and increased methane production in batch experiments. Nitrogen supplementation resulted in a VFA loss due to methane formation by buffering the pH value at nearly neutral conditions (∼ 6.7). However, the alkaline pretreatment with 6g NaOH/100g FCFM improved both the COD solubilization and the VFA yield by 37%, mainly consisted by n-butyric and acetic acids. Copyright © 2015 Elsevier Ltd. All rights reserved.

U.S. Department of Energy physical protection upgrades at the Latvian Academy of Sciences Nuclear Research Center, Latvia

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

Haase, M.; Hine, C.; Robertson, C.

1996-12-31

Approximately five years ago, the Safe, Secure Dismantlement program was started between the US and countries of the Former Soviet Union (FSU). The purpose of the program is to accelerate progress toward reducing the risk of nuclear weapons proliferation, including such threats as theft, diversion, and unauthorized possession of nuclear materials. This would be accomplished by strengthening the material protection, control, and accounting systems within the FSU countries. Under the US Department of Energy`s program of providing cooperative assistance to the FSU countries in the areas of Material Protection, Control, and Accounting (MPC and A), the Latvian Academy of Sciencesmore » Nuclear Research Center (LNRC) near Riga, Latvia, was identified as a candidate site for a cooperative MPC and A project. The LNRC is the site of a 5-megawatt IRT-C pool-type research reactor. This paper describes: the process involved, from initial contracting to project completion, for the physical protection upgrades now in place at the LNRC; the intervening activities; and a brief overview of the technical aspects of the upgrades.« less

Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

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

Lindgren, Eric Richard; Durbin, Samuel G

2007-04-01

The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program providedmore » data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.« less

Experimental investigation of turbulent wall jet

NASA Astrophysics Data System (ADS)

Andre, Matthieu A.; Bardet, Philippe M.

2011-11-01

Water jet flowing on a flat plate surrounded by quiescent air constitutes a standard case for the study of the interaction between turbulence and the liquid-air interface. This is of particular interest in the understanding of heat and mass transfers across interfaces. The structure of the surface has a great influence on the rate of the transfers which is critical for chemical processes like separation or absorption; pool-type nuclear reactor; climate modeling etc. This study focuses on high Froude (8 to 12) and Weber (3300 to 7400) numbers at which the surface exhibits small wavelength and large amplitude deformations, such as ligaments, surface break up with air entrainment and droplets projection. The experiment features a high velocity (up to 7.5 m/s) water wall jet (19.05mm thick at the nozzle exit) flowing on a flat plate (Re =105 to 1 . 5 .105). High speed movies and PLIF visualization show the evolution of the surface from smooth to 2D structures, then 3D disturbances as the turbulence arising from the wall interacts with the surface.

Velocity field measurement in gas-liquid metal two-phase flow with use of PIV and neutron radiography techniques.

PubMed

Saito, Y; Mishima, K; Tobita, Y; Suzuki, T; Matsubayashi, M

2004-10-01

To establish reasonable safety concepts for the realization of commercial liquid-metal fast breeder reactors, it is indispensable to demonstrate that the release of excessive energy due to re-criticality of molten core could be prevented even if a severe core damage accident took place. Two-phase flow due to the boiling of fuel-steel mixture in the molten core pool has a larger liquid-to-gas density ratio and higher surface tension in comparison with those of ordinary two-phase flows such as air-water flow. In this study, to investigate the effect of the recirculation flow on the bubble behavior, visualization and measurement of nitrogen gas-molten lead bismuth in a rectangular tank was performed by using neutron radiography and particle image velocimetry techniques. Measured flow parameters include flow regime, two-dimensional void distribution, and liquid velocity field in the tank. The present technique is applicable to the measurement of velocity fields and void fraction, and the basic characteristics of gas-liquid metal two-phase mixture were clarified.

Magnetic levitation and MHD propulsion

NASA Astrophysics Data System (ADS)

Tixador, P.

1994-04-01

Magnetic levitation and MHD propulsion are now attracting attention in several countries. Different superconducting MagLev and MHD systems will be described concentrating on, above all, the electromagnetic aspect. Some programmes occurring throughout the world will be described. Magnetic levitated trains could be the new high speed transportation system for the 21st century. Intensive studies involving MagLev trains using superconductivity have been carried out in Japan since 1970. The construction of a 43 km long track is to be the next step. In 1991 a six year programme was launched in the United States to evaluate the performances of MagLev systems for transportation. The MHD (MagnetoHydroDynamic) offers some interesting advantages (efficiency, stealth characteristics, ...) for naval propulsion and increasing attention is being paid towards it nowadays. Japan is also up at the top with the tests of Yamato I, a 260 ton MHD propulsed ship. Depuis quelques années nous assistons à un redémarrage de programmes concernant la lévitation et la propulsion supraconductrices. Différents systèmes supraconducteurs de lévitation et de propulsion seront décrits en examinant plus particulièrement l'aspect électromagnétique. Quelques programmes à travers le monde seront abordés. Les trains à sustentation magnétique pourraient constituer un nouveau mode de transport terrestre à vitesse élevée (500 km/h) pour le 21^e siècle. Les japonais n'ont cessé de s'intéresser à ce système avec bobine supraconductrice. Ils envisagent un stade préindustriel avec la construction d'une ligne de 43 km. En 1991 un programme américain pour une durée de six ans a été lancé pour évaluer les performances des systèmes à lévitation pour le transport aux Etats Unis. La MHD (Magnéto- Hydro-Dynamique) présente des avantages intéressants pour la propulsion navale et un regain d'intérêt apparaît à l'heure actuelle. Le japon se situe là encore à la pointe des développements actuels avec en particulier les premiers essais en rade de Kobe de Yamato I, navire de 260 tonnes, entraîné par MHD.

13 CFR 120.1704 - Pool Loans eligible for Pooling.

Code of Federal Regulations, 2010 CFR

2010-01-01

..., construction or renovation of an aquarium, zoo, golf course, or swimming pool; or (iv) To a business covered by... zoos—712130 (“Zoos and Botanical Gardens”). (b) SBA review of a Pool Loan prior to pool formation. SBA...

13 CFR 120.1704 - Pool Loans eligible for Pooling.

Code of Federal Regulations, 2013 CFR

2013-01-01

..., construction or renovation of an aquarium, zoo, golf course, or swimming pool; or (iv) To a business covered by... zoos—712130 (“Zoos and Botanical Gardens”). (b) SBA review of a Pool Loan prior to pool formation. SBA...

13 CFR 120.1704 - Pool Loans eligible for Pooling.

Code of Federal Regulations, 2011 CFR

2011-01-01

..., construction or renovation of an aquarium, zoo, golf course, or swimming pool; or (iv) To a business covered by... zoos—712130 (“Zoos and Botanical Gardens”). (b) SBA review of a Pool Loan prior to pool formation. SBA...

13 CFR 120.1704 - Pool Loans eligible for Pooling.

Code of Federal Regulations, 2014 CFR

2014-01-01

..., construction or renovation of an aquarium, zoo, golf course, or swimming pool; or (iv) To a business covered by... zoos—712130 (“Zoos and Botanical Gardens”). (b) SBA review of a Pool Loan prior to pool formation. SBA...

13 CFR 120.1704 - Pool Loans eligible for Pooling.

Code of Federal Regulations, 2012 CFR

2012-01-01

..., construction or renovation of an aquarium, zoo, golf course, or swimming pool; or (iv) To a business covered by... zoos—712130 (“Zoos and Botanical Gardens”). (b) SBA review of a Pool Loan prior to pool formation. SBA...

A resistant-kernel model of connectivity for amphibians that breed in vernal pools

Treesearch

Bradley W. Compton; Kevin McGarigal; Samuel A. Cushman; Lloyd R. Gamble

2007-01-01

Pool-breeding amphibian populations operate at multiple scales, from the individual pool to surrounding upland habitat to clusters of pools. When metapopulation dynamics play a role in long-term viability, conservation efforts limited to the protection of individual pools or even pools with associated upland habitat may be ineffective over the long term if connectivity...

Life cycle environmental implications of residential swimming pools.

PubMed

Forrest, Nigel; Williams, Eric

2010-07-15

Ownership of private swimming pools in the U.S. grew 2 to 4% per annum from 1997 to 2007. The environmental implications of pool ownership are analyzed by hybrid life cycle assessment (LCA) for nine U.S. cities. An operational model is constructed estimating consumption of chemicals, water, and energy for a typical residential pool. The model incorporates geographical climatic variations and upstream water and energy use from electricity and water supply networks. Results vary considerably by city: a factor of 5-6 for both water and energy use. Water use is driven by aridness and length of the swimming season, while energy use is mainly driven by length of the swimming season. Water and energy impacts of pools are significant, particularly in arid climates. In Phoenix for example pools account for 22% and 13% of a household's electricity and water use, respectively. Measures to reduce water and energy use in pools such as optimizing the pump schedule and covering the pool in winter can realize greater savings than many common household efficiency improvements. Private versus community pools are also compared. Community pools in Phoenix use 60% less swimming pool water and energy per household than subdivisions without community pools.

Regional variation in the biogeochemical and physical characteristics of natural peatland pools.

PubMed

Turner, T Edward; Billett, Michael F; Baird, Andy J; Chapman, Pippa J; Dinsmore, Kerry J; Holden, Joseph

2016-03-01

Natural open-water pools are a common feature of northern peatlands and are known to be an important source of atmospheric methane (CH4). Pool environmental variables, particularly water chemistry, vegetation community and physical characteristics, have the potential to exert strong controls on carbon cycling in pools. A total of 66 peatland pools were studied across three regions of the UK (northern Scotland, south-west Scotland, and Northern Ireland). We found that within-region variability of pool water chemistry was low; however, for many pool variables measured there were significant differences between regions. PCA analysis showed that pools in SW Scotland were strongly associated with greater vegetative cover and shallower water depth which is likely to increase dissolved organic carbon (DOC) mineralisation rates, whereas pools in N Scotland were more open and deeper. Pool water DOC, particulate organic carbon and dissolved CH4 concentrations were significantly different between regions. Pools in Northern Ireland had the highest concentrations of DOC (mean=14.5 mg L(-1)) and CH4 (mean=20.6 μg C L(-1)). Chloride and sulphate concentrations were significantly higher in the pools in N Scotland (mean values 26.3 and 2.40 mg L(-1), respectively) than elsewhere, due to a stronger marine influence. The ratio of UV absorbance at 465 nm to absorbance at 665 nm for pools in Northern Ireland indicated that DOC was sourced from poorly humified peat, potentially increasing the bioavailability and mineralisation of organic carbon in pools compared to the pools elsewhere. This study, which specifically aims to address a lack of basic biogeochemical knowledge about pool water chemistry, clearly shows that peatland pools are highly regionally variable. This is likely to be a reflection of significant regional-scale differences in peatland C cycling. Copyright © 2015 Elsevier B.V. All rights reserved.

Cellular choline and glycine betaine pools impact osmoprotection and phospholipase C production in Pseudomonas aeruginosa.

PubMed

Fitzsimmons, Liam F; Hampel, Ken J; Wargo, Matthew J

2012-09-01

Choline is abundantly produced by eukaryotes and plays an important role as a precursor of the osmoprotectant glycine betaine. In Pseudomonas aeruginosa, glycine betaine has additional roles as a nutrient source and an inducer of the hemolytic phospholipase C, PlcH. The multiple functions for glycine betaine suggested that the cytoplasmic pool of glycine betaine is regulated in P. aeruginosa. We used (13)C nuclear magnetic resonance ((13)C-NMR) to demonstrate that P. aeruginosa maintains both choline and glycine betaine pools under a variety of conditions, in contrast to the transient glycine betaine pool reported for most bacteria. We were able to experimentally manipulate the choline and glycine betaine pools by overexpression of the cognate catabolic genes. Depletion of either the choline or glycine betaine pool reduced phospholipase production, a result unexpected for choline depletion. Depletion of the glycine betaine pool, but not the choline pool, inhibited growth under conditions of high salt with glucose as the primary carbon source. Depletion of the choline pool inhibited growth under high-salt conditions with choline as the sole carbon source, suggesting a role for the choline pool under these conditions. Here we have described the presence of a choline pool in P. aeruginosa and other pseudomonads that, with the glycine betaine pool, regulates osmoprotection and phospholipase production and impacts growth under high-salt conditions. These findings suggest that the levels of both pools are actively maintained and that perturbation of either pool impacts P. aeruginosa physiology.

Cellular Choline and Glycine Betaine Pools Impact Osmoprotection and Phospholipase C Production in Pseudomonas aeruginosa

PubMed Central

Fitzsimmons, Liam F.; Hampel, Ken J.

2012-01-01

Choline is abundantly produced by eukaryotes and plays an important role as a precursor of the osmoprotectant glycine betaine. In Pseudomonas aeruginosa, glycine betaine has additional roles as a nutrient source and an inducer of the hemolytic phospholipase C, PlcH. The multiple functions for glycine betaine suggested that the cytoplasmic pool of glycine betaine is regulated in P. aeruginosa. We used 13C nuclear magnetic resonance (13C-NMR) to demonstrate that P. aeruginosa maintains both choline and glycine betaine pools under a variety of conditions, in contrast to the transient glycine betaine pool reported for most bacteria. We were able to experimentally manipulate the choline and glycine betaine pools by overexpression of the cognate catabolic genes. Depletion of either the choline or glycine betaine pool reduced phospholipase production, a result unexpected for choline depletion. Depletion of the glycine betaine pool, but not the choline pool, inhibited growth under conditions of high salt with glucose as the primary carbon source. Depletion of the choline pool inhibited growth under high-salt conditions with choline as the sole carbon source, suggesting a role for the choline pool under these conditions. Here we have described the presence of a choline pool in P. aeruginosa and other pseudomonads that, with the glycine betaine pool, regulates osmoprotection and phospholipase production and impacts growth under high-salt conditions. These findings suggest that the levels of both pools are actively maintained and that perturbation of either pool impacts P. aeruginosa physiology. PMID:22753069

Comparison of fecal pooling strategies for detection of Mycobacterium avium ssp. paratuberculosis in cattle.

PubMed

McKenna, S L B; Ritter, C; Dohoo, I; Keefe, G P; Barkema, H W

2018-05-23

In herds with typical moderate to low within-herd prevalence, testing for Mycobacterium avium ssp. paratuberculosis (MAP), the infectious agent of Johne's disease, will be more cost-effective if individual fecal samples are cultured in composite pools. However, sensitivity to classify a pool containing 1 or more positive individual samples as positive may depend on pool size and number of individual positive samples within a pool. Fecal samples collected from 994 dairy cows sampled at slaughter were cultured to detect MAP. Culturing was done both individually and as composite pooled samples using the TREK ESP Culture System II broth medium (Thermo Fisher Scientific, Trek Diagnostic Systems Inc., Cleveland, OH). Composite samples consisted of pools containing feces from 3, 5, 8, 10, or 15 cows. The number of individual fecal culture-positive cows within each pool ranged from 0 to 4. Culture of individual fecal samples detected MAP in 36 (3.6%) of the 994 cows. Individual samples that were detected within the first 50 d by TREK ESP Culture System II were more likely to lead to a positive pool result. In total, 840 pooled fecal samples were examined for presence of MAP, and of those, 272 pools actually contained feces from fecal culture-positive cows. The crude sensitivity (proportion of pools that contained at least 1 fecal-positive cow that tested positive) for pools of 3, 5, 8, 10, and 15 was 47, 67, 44, 59, and 39%, respectively. Across pools, an increase of the number of fecal culture-positive samples from 1 to 2 enhanced overall crude sensitivity from 44 to 71%. However, sensitivity did not further increase for pools with 3 or 4 fecal culture-positive samples (63 and 60%, respectively). Additionally, a simulation analysis assessing probability of pooled fecal samples being positive in herds of 50 and 100 cows was conducted. The simulation assumed that 1, 2, or 5 cows per herd were MAP fecal culture-positive and that pools of 5 and 10 were used. This low-prevalence herd simulation indicated that weighted mean herd probabilities of detecting a positive herd ranged between 52 and 99.3%, with the lowest probability for pools of 10 with 1 positive cow in the herd and the highest probability for pools of 5 with 5 positive cows in the herd. However, overall, pools of 5 and 10 had similar diagnostic capabilities, enabling cost savings by utilizing pools of 10. Copyright © 2018 American Dairy Science Association. Published by Elsevier Inc. All rights reserved.

Effects of habitat isolation on the recovery of fish assemblages in experimentally defaunated stream pools in Arkansas

Treesearch

David George Lonzarich; Melvin L. Warren; Mary Ruth Elger Lonzrich

1998-01-01

The authors removed fish from pools in two Arkansas streams to determine recolonization rates and the effects of isolation (i.e., riffle length, riffle depth, distance to large source pools, and location), pool area, and assemblage size on recovery. To determine pool-specific recovery rates, the authors repeatedly snorkeled 12 pools over a 40-day recovery period....

Les cooperatives et l'electrification rurale du Quebec, 1945--1964

NASA Astrophysics Data System (ADS)

Dorion, Marie-Josee

Cette these est consacree a l'histoire de l'electrification rurale du Quebec, et, plus particulierement, a l'histoire des cooperatives d'electricite. Fondees par vagues successives a partir de 1945, les cooperatives rurales d'electricite ont ete actives dans plusieurs regions du Quebec et elles ont electrifie une partie significative des zones rurales. Afin de comprendre le contexte de la creation des cooperatives d'electricite, notre these debute (premiere partie) par une analyse du climat sociopolitique des annees precedant la naissance du systeme cooperatif d'electrification rurale. Nous y voyons de quelle facon l'electrification rurale devient progressivement, a partir de la fin des annees 1920, une question d'actualite a laquelle les divers gouvernements qui se succedent tentent de trouver une solution, sans engager---ou si peu---les fonds de l'Etat. En ce sens, la premiere etatisation et la mise sur pied d'Hydro-Quebec, en 1944, marquent une rupture quant au mode d'action privilegie jusque-la. La nouvelle societe d'Etat se voit cependant retirer son mandat d'electrifier le monde rural un an apres sa fondation, car le gouvernement Duplessis, de retour au pouvoir, prefere mettre en place son propre modele d'electrification rurale. Ce systeme repose sur des cooperatives d'electricite, soutenues par un organisme public, l'Office de l'electrification rurale (OER). L'OER suscite de grandes attentes de la part des ruraux et c'est par centaines qu'ils se manifestent. Cet engouement pour les cooperatives complique la tache de l'OER, qui doit superviser de nouvelles societes tout en assurant sa propre organisation. Malgre des hesitations et quelques delais introduits par un manque de connaissances techniques et de personnel qualifie, les commissaires de l'OER se revelent perspicaces et parviennent a mettre sur pied un systeme cooperatif d'electrification rurale qui produit des resultats rapides. Il leur faudra cependant compter sur l'aide des autres acteurs engages dans l'electrification, les organismes publics et les compagnies privees d'electricite. Cette periode de demarrage et d'organisation, traitee dans la deuxieme partie de la these, se termine en 1947-48, au moment ou l'OER et les cooperatives raffermissent leur maitrise du systeme cooperatif d'electrification rurale. Les annees 1948 a 1955 (troisieme partie de these) correspondent a une periode de croissance pour le mouvement cooperatif. Cette partie scrute ainsi le developpement des cooperatives, les vastes chantiers de construction et l'injection de millions de dollars dans l'electrification rurale. Cette troisieme partie prend egalement acte des premiers signes que quelque chose ne va pas si bien dans le monde cooperatif. Nous y verrons egalement les ruraux a l'oeuvre: comme membres, d'abord, mais aussi en tant que benevoles, puis a l'emploi des cooperatives. La quatrieme et derniere partie, les annees 1956 a 1964, aborde les changements majeurs qui ont cours dans l'univers cooperatif; il s'agit d'une ere nouvelle et difficile pour le mouvement cooperatif, dont les reseaux paraissent inadaptes aux changements de profil de la consommation d'electricite des usagers. L'OER sent alors le besoin de raffermir son controle des cooperatives, car il pressent les problemes et les defis auxquels elles auront a faire face. Notre etude se termine par l'acquisition des cooperatives par Hydro-Quebec, en 1963-64. Fondee sur des sources riches et variees, notre demarche propose un eclairage inedit sur une dimension importante de l'histoire de l'electricite au Quebec. Elle permet, ce faisant, de saisir les rouages et l'action de l'Etat sous un angle particulier, avant sa profonde transformation amorcee au cours des annees 1960. De meme, elle apporte quelques cles nouvelles pour une meilleure comprehension de la dynamique des milieux ruraux de cette periode.

Sediment transport through self-adjusting, bedrock-walled waterfall plunge pools

NASA Astrophysics Data System (ADS)

Scheingross, Joel S.; Lamb, Michael P.

2016-05-01

Many waterfalls have deep plunge pools that are often partially or fully filled with sediment. Sediment fill may control plunge-pool bedrock erosion rates, partially determine habitat availability for aquatic organisms, and affect sediment routing and debris flow initiation. Currently, there exists no mechanistic model to describe sediment transport through waterfall plunge pools. Here we develop an analytical model to predict steady-state plunge-pool depth and sediment-transport capacity by combining existing jet theory with sediment transport mechanics. Our model predicts plunge-pool sediment-transport capacity increases with increasing river discharge, flow velocity, and waterfall drop height and decreases with increasing plunge-pool depth, radius, and grain size. We tested the model using flume experiments under varying waterfall and plunge-pool geometries, flow hydraulics, and sediment size. The model and experiments show that through morphodynamic feedbacks, plunge pools aggrade to reach shallower equilibrium pool depths in response to increases in imposed sediment supply. Our theory for steady-state pool depth matches the experiments with an R2 value of 0.8, with discrepancies likely due to model simplifications of the hydraulics and sediment transport. Analysis of 75 waterfalls suggests that the water depths in natural plunge pools are strongly influenced by upstream sediment supply, and our model provides a mass-conserving framework to predict sediment and water storage in waterfall plunge pools for sediment routing, habitat assessment, and bedrock erosion modeling.

Automation of water supply and recirculation-filtration of water at a swimming pool using Zelio PLC

NASA Astrophysics Data System (ADS)

Diniş, C. M.; Popa, G. N.; Iagăr, A.

2018-01-01

The paper proposes the use of the Zelio PLC for the automation of the water supply and recirculation-filtration system of a swimming pool. To do this, the Zelio SR3B261BD - 24V DC with 10 digital inputs (24V DC) and 10 digital outputs (relay contacts) was used. The proposed application makes the control of the water supply pumps and the water recirculation-filtration from a swimming pool. The recirculation-filtration systems for pools and swimming pools are designed to ensure water cleaning and recirculation to achieve optimum quality and lasting service life. The water filtration process is one of the important steps in water treatment in polls and swimming pools. It consists in recirculation of the entire volume of water and begins by absorbing the water in the pool by means of a pump followed by the passing of water through the filter, disinfectant and pH dosing, and reintroducing the water back into the pool or swimming pool through the discharge holes. Filters must to work 24 hours a day to remove pollutants from pools or swimming pools users. Filtration removes suspension particles with different origins. All newly built pools and swimming pools must be fitted with water recirculation systems, and existing ones will be equipped with water recirculation and water treatment systems.

On-line fission products measurements during a PWR severe accident: the French DECA-PF project

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

Ducros, G.; Allinei, P.G.; Roure, C.

Following the Fukushima accident, a lot of recommendations was drawn by international organizations (IAEA, OECD, NUGENIA network...) in order to improve the safety in such accidental conditions and mitigate their consequences. One of these recommendations was to improve the robustness of the instrumentation, which was dramatically lacking at Fukushima, as well as to better determine the Source Term involved in nuclear accident. The DECA-PF project (Diagnosis of a degraded reactor core through Fission Product measurements) was elaborated in this context and selected as one of 21 collaborative R and D projects in the field of nuclear safety and radioprotection, fundedmore » in May 2013 by the French National Research Agency. Over the months following the Fukushima accident, a CEA crisis team was held in order to analyze on-line the situation taking into account the data delivered by TEPCO and other organizations. Despite the difficulties encountered concerning the reliability of these data, the work performed showed the high capacity of Fission Products (FP) measurements to get a diagnosis relative to the status of the reactors and the spent fuel pools (SFP). Based on these FP measurements, it was possible to conclude that the main origin of the releases was coming from the cores and not from the SFP, in particular for SFP-4 which was of high concern, and that the degradation level of the reactors was very large, including probably an extensive core melting. To improve the reliability of this kind of diagnosis, the necessity to get such measurements as soon as possible after the accident and as near as possible from the reactor was stressed. In this way the present DECA-PF project intends to develop a new and innovative instrumentation taking into account the design of the French nuclear power plants on which sand bed filters have been implemented for severe accident management. Three complementary techniques, devoted to measure the FP release on-line, are being studied: - Gamma spectrometry, with an industrial objective to build a prototype aimed at improving the capacity of the present radiation monitoring system, - Gas chromatography, for the quantification of the fission gases (Xe, Kr) as well as potential carbon oxides produced in case of Molten Corium Concrete Interaction, - Optical absorption spectroscopy, the objective of this most innovative technique being to quantify the tetra-oxide of ruthenium, which could be produced in case of lower head failure, and the gaseous forms of iodine (molecular and organic) released in the environment. A global description and the present status of this project is presented, focusing on the Source Term establishment at the outlet stack of the sand bed filters and on the perspectives of implementation of the on-line gamma spectrometry equipment. (authors)« less

75 FR 23262 - Notice of Attendance at the Arkansas Public Service Commission Technical Conference

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-03

... Power Pool, Inc. Docket No. ER10-696, Southwest Power Pool, Inc. Docket No. ER10-697, Southwest Power Pool, Inc. Docket No. ER10-698, Southwest Power Pool, Inc. Docket No. ER10-700, Southwest Power Pool...

HYDROLOGY AND LANDSCAPE CONNECTIVITY OF VERNAL POOLS

EPA Science Inventory

Vernal pools are shaped by hydrologic processes which influence many aspects of pool function. The hydrologic budget of a pool can be summarized by a water balance equation that relates changes in the amount of water in the pool to precipitation, ground- and surface-water flows, ...

The power of simplification: Operator interface with the AP1000{sup R} during design-basis and beyond design-basis events

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

Williams, M. G.; Mouser, M. R.; Simon, J. B.

2012-07-01

The AP1000{sup R} plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditionsmore » in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been designed to be reliable in these conditions. The primary goal of any such actions is to maintain or refill the passive inventory available to cool the core, containment and spent fuel pool in the safety-related and seismically qualified Passive Containment Cooling Water Storage Tank (PCCWST). The seismically-qualified, ground-mounted Passive Containment Cooling Ancillary Water Storage Tank (PCCAWST) is also available for this function as appropriate. The primary effect of these actions would be to increase the coping time for the AP1000 during design basis events, as well as events such as those described above, from 72 hours without operator intervention to 7 days with minimal operator actions. These Operator actions necessary to protect the health and safety of the public are addressed in the Post-72 Hour procedures, as well as some EOPs, AOPs, ARPs and the Severe Accident Management Guidelines (SAMGs). Should the event continue to become more severe and plant conditions degrade further with indications of inadequate core cooling, the SAMGs provide guidance for strategies to address these hypothetical severe accident conditions. The AP1000 SAMG diagnoses and actions are prioritized to first utilize the AP1000 features that are expected to retain a damaged core inside the reactor vessel. Only one strategy is undertaken at any time. This strategy will be followed and its effectiveness evaluated before other strategies are undertaken. This is a key feature of both the symptom-oriented AP1000 EOPs and the AP1000 SAMGs which maximizes the probability of retaining a damaged core inside the reactor vessel and containment while minimizing the chances for confusion and human errors during implementation. The AP1000 SAMGs are simple and straight-forward and have been developed with considerable input from human factors and plant operations experts. Most importantly, and different from severe accident management strategies for other plants, the AP1000 SAMGs do not require diagnosis of the location of the core (i.e., whether reactor vessel failure has occurred). This is a fundamental consequence of the AP1000 In-Vessel Retention approach, which allows severe accident management to be based on fundamental principles (e.g. provide coolant as close as possible to the core) that do not change during a specific event. This eliminates the need for one of the more difficult diagnostic requirements, since reactor vessel failure does not directly relate to any measurable plant parameter, and differs from other designs in that an engineered failure of the pressure vessel' (e.g. core catcher) is never required. (authors)« less

Eco-engineered rock pools: a concrete solution to biodiversity loss and urban sprawl in the marine environment

NASA Astrophysics Data System (ADS)

Firth, Louise B.; Browne, Keith A.; Knights, Antony M.; Hawkins, Stephen J.; Nash, Róisín

2016-09-01

In coastal habitats artificial structures typically support lower biodiversity and can support greater numbers of non-native and opportunistic species than natural rocky reefs. Eco-engineering experiments are typically trialed to succeed; but arguably as much is learnt from failure than from success. Our goal was to trial a generic, cost effective, eco-engineering technique that could be incorporated into rock armouring anywhere in the world. Artificial rock pools were created from manipulated concrete between boulders on the exposed and sheltered sides of a causeway. Experimental treatments were installed in locations where they were expected to fail and compared to controls installed in locations in which they were expected to succeed. Control pools were created lower on the structure where they were immersed on every tidal cycle; experimental pools were created above mean high water spring tide which were only immersed on spring tides. We hypothesised that lower and exposed pools would support significantly higher taxon and functional diversity than upper and sheltered pools. The concrete pools survived the severe winter storms of 2013/14. After 12 months, non-destructive sampling revealed significantly higher mean taxon and functional richness in lower pools than upper pools on the exposed side only. After 24 months the sheltered pools had become inundated with sediments, thus failing to function as rock pools as intended. Destructive sampling on the exposed side revealed significantly higher mean functional richness in lower than upper pools. However, a surprisingly high number of taxa colonised the upper pools leading to no significant difference in mean taxon richness among shore heights. A high number of rare taxa in the lower pools led to total taxon richness being almost twice that of upper pools. These findings highlight that even when expected to fail concrete pools supported diverse assemblages, thus representing an affordable, replicable means of enhancing biodiversity on a variety of artificial structures.

HYDROLOGY AND LANDSCAPE CONNECTIVITY OF VERNAL POOLS OF THE GLACIATED NORTHEAST

EPA Science Inventory

The hydrologic budget of a vernal pool influences many aspects of pool function and is the basis for pool life. Although alternating wet and dry periods occur in most wetlands, vernal pools can experience extreme moisture fluctuations. This variability results from intra- and i...

75 FR 15707 - Notice of FERC Staff Attendance at the Southwest Power Pool Board of Directors/Members Committee...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-30

..., Inc. Docket No. ER10-694, Southwest Power Pool, Inc. Docket No. ER10-696, Southwest Power Pool, Inc. Docket No. ER10-697, Southwest Power Pool, Inc. Docket No. ER10-698, Southwest Power Pool, Inc. Docket No...

Swimming Pools. Managing School Facilities, Guide 2.

ERIC Educational Resources Information Center

Department for Education and Employment, London (England). Architects and Building Branch.

This guide for schools with swimming pools offers advice concerning appropriate training for pool managers, the importance of water quality and testing, safety in the handling of chemicals, maintenance and cleaning requirements, pool security, and health concerns. The guide covers both indoor and outdoor pools, explains some technical terms,…

13 CFR 120.1705 - Pool formation requirements.

Code of Federal Regulations, 2010 CFR

2010-01-01

... requirements. SBA may adjust the Pool characteristics periodically based on program experience and market... a Pool involving a Pool Loan it does not own, it must purchase the Loan Interest it proposes to pool... purchase the Loan Interest and take it into inventory or settle the purchase of the Loan Interest through...

Vernal Pool Conservation in Connecticut: An Assessment and Recommendations

NASA Astrophysics Data System (ADS)

Preisser, Evan L.; Kefer, Jennifer Yelin; Lawrence, Jessica D.; Clark, Tim W.

2000-11-01

Vernal pools, a variety of ephemeral wetlands, are threatened in many areas of the United States. As habitat fragmentation and degradation increase, some vernal pool amphibian species are declining in numbers. Uneven implementation of state regulations further hampers effective conservation. To prevent further species decline and vernal pool loss, we evaluated alternatives for improving vernal pool conservation. We used transcripts from a recent vernal pool conference, interviews with members of relevant interest groups, and a literature review to determine opportunities for and constraints on improving vernal pool conservation policy. Participants from different interest groups had very diverse views about appropriate protection strategies. We have examined these different perspectives and alternatives and offer policy recommendations on both the state and local level. These recommendations can foster awareness of vernal pools as unique habitats, increase protection of these areas, and expand citizen participation in the vernal pool regulatory process.

A revised velocity-reversal and sediment-sorting model for a high-gradient, pool-riffle stream

USGS Publications Warehouse

Thompson, D.M.; Wohl, E.E.; Jarrett, R.D.

1996-01-01

Sediment-sorting processes related to varying channel-bed morphology were investigated from April to November 1993 along a 1-km pool-riffle and step-pool reach of North Saint Vrain Creek, a small mountain stream in the Rocky Mountains of northern Colorado. Measured cross-sectional areas of flow were used to suggest higher velocities in pools than in riffles at high flow. Three hundred and sixteen tracer particles, ranging in size from 16 mm to 256 mm, were placed in two separate pool-riffle-pool sequences and used to assess sediment-sorting patterns and sediment-transport competence variations. Tracer-particle depositional evidence indicated higher sediment-transport competence in pools than in riffles at high flow. Pool-riffle sediment sorting may be created by velocity reversals, and more localized sorting results from gravitational forces along the upstream sloping portion of the channel bed located at the downstream end of pools.

From antidunes to step-pools

NASA Astrophysics Data System (ADS)

Recking, Alain; Leduc, Pauline

2014-05-01

Step-pools are bed morphologies that are typical in high-gradient streams , recognizable by a staircase-like longitudinal profile resulting from accumulation of cobbles and boulders that are transverse to the channel and alternating with pools containing finer sediments. Within the last two decades step-pools have been the subject of increased efforts to characterize their nature; however their origin is still in debate. Researchers have very soon suspected step-pools to be the residual form of antidunes produced during flooding, but this hypothesis was continuously contested. Other theories has been proposed, considering, that step-pool profile develops a maximum flow resistance, or that pools geometry is controlled by the energy of a falling jet, or that steps form by boulders accumulation in a channel-spanning manner. All these theories gave very satisfying results when compared with experimental data, but does it mean that the antidune theory should we abandoned? We performed new flume experiments on steep slopes to investigate the antidune origin for step-pools. Our experiments showed that step-pools can have several origins, depending on the flow conditions and sediment mixture used. In some circumstances antidunes were well observed but did not produce stable step-pools morphology. In many occasions, step-pools obtained in the flume were isolated step-pools, with no real apparent periodicity. Only a few flow and sediment conditions allowed us to reproduce trains of antidunes which stabilized at the flow recession to produce stable periodical step-pools. These conditions are presented and discussed.

An Experimental Investigation On The Antidune Origin of Step-pools

NASA Astrophysics Data System (ADS)

Recking, A.; Leduc, P.

2013-12-01

Step-pools are bed morphologies that are typical in high-gradient streams , recognizable by a staircase-like longitudinal profile resulting from accumulation of cobbles and boulders that are transverse to the channel and alternating with pools containing finer sediments. Within the last two decades step-pools have been the subject of increased efforts to characterize their nature; however their origin is still in debate. Researchers have very soon suspected step-pools to be the residual form of antidunes produced during flooding, but this hypothesis was continuously contested. Other theories has been proposed, considering, that step-pool profile develops a maximum flow resistance, or that pools geometry is controlled by the energy of a falling jet, or that steps form by boulders accumulation in a channel-spanning manner. All these theories gave very satisfying results when compared with experimental data, but does it mean that the antidune theory should we abandoned? We performed new flume experiments on steep slopes to investigate the antidune origin for step-pools. Our experiments showed that step-pools can have several origins, depending on the flow conditions and sediment mixture used. In some circumstances antidunes were well observed but did not produce stable step-pools morphology. In many occasions, step-pools obtained in the flume were isolated step-pools, with no real apparent periodicity. Only a few flow and sediment conditions allowed us to reproduce trains of antidunes which stabilized at the flow recession to produce stable periodical step-pools. These conditions are presented and discussed.

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

Peatland Open-water Pool Biogeochemistry: The Influence of Hydrology and Vegetation

NASA Astrophysics Data System (ADS)

Arsenault, J.; Talbot, J.; Moore, T. R.

2017-12-01

Peatland open-water pools are net sources of carbon to the atmosphere. However, their interaction with the surrounding peat remains poorly known. In a previous study, we showed that shallow pools are richer in nutrients than deep pools. While depth was the main driver of biogeochemistry variations across time and space, analyses also showed that pool's adjacent vegetation may have an influence on water chemistry. Our goal is to understand the relationship between the biogeochemistry of open-water pools and their surroundings in a subboreal ombrotrophic peatland of southern Quebec (Canada). To assess the influence of vegetation on pool water chemistry, we compare two areas covered with different types of vegetation: a forested zone dominated by spruce trees and an open area mostly covered by Sphagnum spp. To evaluate the direction of water (in or out of the pools), we installed capacitance water level probes in transects linking pools in the two zones. Wells were also installed next to each probe to collect peat pore water samples. Samples were taken every month during summer 2017 and analyzed for dissolved organic carbon, nitrogen and phosphorus, pH and specific UV absorbance. Preliminary results show differences in peat water chemistry depending on the dominant vegetation. In both zones, water levels fluctuations are disconnected between peat and the pools, suggesting poor horizontal water movement. Pool water chemistry may be mostly influenced by the immediate surrounding vegetation than by the local vegetation pattern. Climate and land-use change may affect the vegetation structure of peatlands, thus affecting pool biogeochemistry. Considering the impact of pools on the overall peatland capacity to accumulate carbon, our results show that more focus must be placed on pools to better understand peatland stability over time.

Monitoring Physicochemical and Nutrient Dynamics Along a Development Gradient in Maine Ephemeral Wetlands

NASA Astrophysics Data System (ADS)

Podzikowski, L. Y.; Capps, K. A.; Calhoun, A.

2014-12-01

Vernal pools are ephemeral wetlands in forested landscapes that fill with snowmelt, precipitation, and/or groundwater in the spring, and characteristically dry down through the summer months. Typically, vernal pool research has focused on the population and community ecology of pool-breeding organisms (amphibians and macroinvertebrates) conducted during their relatively short breeding season. Yet, little is known about the temporal variability of biogeochemical processes within and among vernal pools in urbanizing landscapes. In this study, we monitored physicochemical characteristics and nutrient dynamics in 22 vernal pools in central Maine post thaw in 2014. Four pristine pools were sampled weekly in five locations within the pool for ambient nutrient concentrations (SRP, NH4, NOx) and at three locations for physicochemical characteristics (DO, pH, temperature, conductivity). In the remaining 18 pools, we sampled one location for nutrients and three locations for physicochemical characteristics at least monthly to estimate the influence of increasing urbanization on the physical and chemical environment. Our data suggest most pools found in urbanizing areas have higher conductivity (developed sites ranging 18.52 - 1238 μS cm-1 compared to pristine between 14.08 - 58.4 μS cm-1). Previous work suggests forested pools exhibit dystrophic conditions with high coloration from DOC limiting primary production due to increased light attenuation in pools. However, both pristine and urban pools experienced spikes in DO (>100% saturation) throughout the day, suggesting that high productivity is not a reliable indicator of the effects of urbanization on vernal pools. We argue that continued monitoring of vernal pools along a gradient of urbanization could give insight into the role of ephemeral wetlands as potential biogeochemical hotspots and may also indicate how human development may alter biogeochemical cycling in ephemeral wetlands.

A computational proposal for designing structured RNA pools for in vitro selection of RNAs.

PubMed

Kim, Namhee; Gan, Hin Hark; Schlick, Tamar

2007-04-01

Although in vitro selection technology is a versatile experimental tool for discovering novel synthetic RNA molecules, finding complex RNA molecules is difficult because most RNAs identified from random sequence pools are simple motifs, consistent with recent computational analysis of such sequence pools. Thus, enriching in vitro selection pools with complex structures could increase the probability of discovering novel RNAs. Here we develop an approach for engineering sequence pools that links RNA sequence space regions with corresponding structural distributions via a "mixing matrix" approach combined with a graph theory analysis. We define five classes of mixing matrices motivated by covariance mutations in RNA; these constructs define nucleotide transition rates and are applied to chosen starting sequences to yield specific nonrandom pools. We examine the coverage of sequence space as a function of the mixing matrix and starting sequence via clustering analysis. We show that, in contrast to random sequences, which are associated only with a local region of sequence space, our designed pools, including a structured pool for GTP aptamers, can target specific motifs. It follows that experimental synthesis of designed pools can benefit from using optimized starting sequences, mixing matrices, and pool fractions associated with each of our constructed pools as a guide. Automation of our approach could provide practical tools for pool design applications for in vitro selection of RNAs and related problems.

Study on Dynamic Development of Three-dimensional Weld Pool Surface in Stationary GTAW

NASA Astrophysics Data System (ADS)

Huang, Jiankang; He, Jing; He, Xiaoying; Shi, Yu; Fan, Ding

2018-04-01

The weld pool contains abundant information about the welding process. In particular, the type of the weld pool surface shape, i. e., convex or concave, is determined by the weld penetration. To detect it, an innovative laser-vision-based sensing method is employed to observe the weld pool surface of the gas tungsten arc welding (GTAW). A low-power laser dots pattern is projected onto the entire weld pool surface. Its reflection is intercepted by a screen and captured by a camera. Then the dynamic development process of the weld pool surface can be detected. By observing and analyzing, the change of the reflected laser dots reflection pattern, for shape of the weld pool surface shape, was found to closely correlate to the penetration of weld pool in the welding process. A mathematical model was proposed to correlate the incident ray, reflected ray, screen and surface of weld pool based on structured laser specular reflection. The dynamic variation of the weld pool surface and its corresponding dots laser pattern were simulated and analyzed. By combining the experimental data and the mathematical analysis, the results show that the pattern of the reflected laser dots pattern is closely correlated to the development of weld pool, such as the weld penetration. The concavity of the pool surface was found to increase rapidly after the surface shape was changed from convex to concave during the stationary GTAW process.

Variability of chlorination by-product occurrence in water of indoor and outdoor swimming pools.

PubMed

Simard, Sabrina; Tardif, Robert; Rodriguez, Manuel J

2013-04-01

Swimming is one of the most popular aquatic activities. Just like natural water, public pool water may contain microbiological and chemical contaminants. The purpose of this study was to study the presence of chemical contaminants in swimming pools, in particular the presence of disinfection by-products (DBPs) such as trihalomethanes (THMs), haloacetic acids (HAAs) and inorganic chloramines (CAMi). Fifty-four outdoor and indoor swimming pools were investigated over a period of one year (monthly or bi-weekly sampling, according to the type of pool) for the occurrence of DBPs. The results showed that DBP levels in swimming pools were greater than DBP levels found in drinking water, especially for HAAs. Measured concentrations of THMs (97.9 vs 63.7 μg/L in average) and HAAs (807.6 vs 412.9 μg/L in average) were higher in outdoor pools, whereas measured concentrations of CAMi (0.1 vs 0.8 mg/L in average) were higher in indoor pools. Moreover, outdoor pools with heated water contained more DBPs than unheated pools. Finally, there was significant variability in tTHM, HAA9 and CAMi levels in pools supplied by the same municipal drinking water network, suggesting that individual pool characteristics (number of swimmers) and management strategies play a major role in DBP formation. Copyright © 2012 Elsevier Ltd. All rights reserved.

Vernal Pool Lessons and Activities.

ERIC Educational Resources Information Center

Childs, Nancy; Colburn, Betsy

This curriculum guide accompanies Certified: A Citizen's Step-by-Step Guide to Protecting Vernal Pools which is designed to train volunteers in the process of identifying vernal pool habitat so that as many of these pools as possible can be certified by the Massachusetts Natural Heritage and Endangered Species Program. Vernal pools are a kind of…

Mercury bioaccumulation in wood frogs developing in seasonal pools

USGS Publications Warehouse

Loftin, Cynthia S.; Calhoun, Aram J.K.; Nelson, Sarah J.; Elskus, Adria; Simon, Kevin S.

2012-01-01

Seasonal woodland pools contribute significant biomass to terrestrial ecosystems through production of pool-breeding amphibians. The movement of amphibian metamorphs potentially transports toxins bioaccumulated during larval development in the natal pool into the surrounding terrestrial environment. We documented total mercury (THg) in seasonal woodland pool water, sediment, litter, and Lithobates sylvaticus LeConte (Wood Frog) in Acadia National Park, ME. THg concentrations in pool water varied over the study season, increasing during April—June and remaining high in 2 of 4 pools upon October refill. Water in pools surrounded by softwoods had lower pH, greater dissolved organic carbon, and greater THg concentrations than pools surrounded by hardwoods, with seasonal patterns in sediment THg but not litter THg. THg increased rapidly from near or below detection in 1–2 week old embryos (<0.2 ng; 0–0.49 ppb wet weight) to 17.1–54.2 ppb in tadpoles within 6 weeks; 7.2–42.0% of THg was methyl Hg in tadpoles near metamorphosis. Metamorphs emigrating from seasonal pools may transfer mercury into terrestrial food webs.

Impact of the 1993 flood on the distribution of organic contaminants in bed sediments of the Upper Mississippi River

USGS Publications Warehouse

Barber, L.B.; Writer, J.H.

1998-01-01

The 1500 km Upper Mississippi River (UMR) consists of 29 navigation pools and can be divided into the upper reach (pools 1-4), the middle reach (pools 5-13), and the lower reach (pools 14-26). Comparison of composite bed sediment samples collected from the downstream third of 24 pools before and after the 1993 UMR flood provides fieldscale data on the effect of the flood on sediment organic compound distributions. The sediments were analyzed for organic carbon, coprostanol, polynuclear aromatic hydrocarbons including pyrene, linear alkylbenzene-sulfonates, polychlorinated biphenyls (PCBs), and organochlorine pesticides. Most of the target compounds were detected in all of the sediment samples, although concentrations were generally <1 mg/kg. The highest concentrations typically occurred in the upper reach, an urbanized area on a relatively small river. Pool 4 (Lake Pepin) is an efficient sediment trap, and concentrations of the compounds below pool 4 were substantially lower than those in pools 2-4. Differences in concentrations before and after the 1993 flood also were greatest in the upper reach. In pools 1-4, concentrations of pyrene and PCBs decreased after the flood whereas coprostanol increased. These results suggest that bed sediments stored in the pools were diluted or buried by sediments with different organic compound compositions washed in from urban and agricultural portions of the watershed.The 1500 km Upper Mississippi River (UMR) consists of 29 navigation pools and can be divided into the upper reach (pools 1-4), the middle reach (pools 5-13), and the lower reach (pools 14-26). Comparison of composite bed sediment samples collected from the downstream third of 24 pools before and after the 1993 UMR flood provides field-scale data on the effect of the flood on sediment organic compound distributions. The sediments were analyzed for organic carbon, coprostanol, polynuclear aromatic hydrocarbons including pyrene, linear alkylbenzene-sulfonates, polychlorinated biphenyls (PCBs), and organochlorine pesticides. Most of the target compounds were detected in all of the sediment samples, although concentrations were generally <1 mg/kg. The highest concentrations typically occurred in the upper reach, an urbanized area on a relatively small river. Pool 4 (Lake Pepin) is an efficient sediment trap, and concentrations of the compounds below pool 4 were substantially lower than those in pools 2-4. Differences in concentrations before and after the 1993 flood also were greatest in the upper reach. In pools 1-4, concentrations of pyrene and PCBs decreased after the flood whereas coprostanol increased. These results suggest that bed sediments stored in the pools were diluted or buried by sediments with different organic compound compositions washed in from urban and agricultural portions of the watershed.

Longevity of Wood-Forced Pools in an Old-Growth Forest

NASA Astrophysics Data System (ADS)

Buffington, J. M.; Woodsmith, R. D.; Johnson, A. C.

2009-12-01

Wood debris plays an important role in scouring pools in forest channels and providing resultant habitat for aquatic organisms. We investigated the longevity of such pools in a gravel-bed river flowing through old-growth forest in southeastern Alaska by aging trees and “bear’s bread” fungi (Ganoderma applanatum, Fomitopsis pinicola) growing on pool-forming wood debris. Ages were determined by counting annual growth rings from cores and cross sections of trees and fungi growing on the wood debris. These ages are minimum values because they do not account for lag time between debris recruitment and seedling/spore establishment on the debris, nor do they account for flood scour that may periodically reset tree and fungi growth on the debris. The study stream has a gradient of about 1%, bankfull width and depth of 13.3 and 0.78 m, respectively, median grain size of 18 mm, a high wood loading (0.8 pieces/m), and a correspondingly low pool spacing (0.3 bankfull widths/pool), with 81% of the pools forced by wood debris. The size of wood debris in the study stream is large relative to the channel (average log length of 7.6 m and diameter of 0.35 m), rendering most debris immobile. Eighty-one pool-forming pieces of wood were dated over 1.2 km of stream length, with 28% of these pieces causing scour of more than one pool. In all, 122 wood-forced pools were dated, accounting for 38% of all pools at the site and 47% of the wood-forced pools. Fifty-three percent of the wood-forced pools lacked datable wood because these pieces either: were newly recruited; had been scoured by floods; or were contained below the active channel elevation, prohibiting vegetation establishment on the wood debris (the most common cause). The debris age distribution declined exponentially from 2 to 113 yrs., with a median value of 18 yrs. Similar exponential residence time distributions have been reported in other studies, but our analysis focused specifically on the ages of pool-forming wood as opposed to all in-channel wood. Most pool scour was relatively recent (60% ≤ 25 yrs. old), but 16% of the pools were old features (50-100+ yrs.), indicating long-term availability of pool habitats at the study site. Future studies will use these results to develop a wood budget model that accounts for pool scour and availability of pool habitats. For such modeling, our data suggest that stand-replacing disturbances (e.g. wildfire, riparian clear cutting) will cause a sharp drop in the numbers of wood-forced pools, as most of those are ≤ 25 yrs. old.

A method of measuring a molten metal liquid pool volume

DOEpatents

Garcia, G.V.; Carlson, N.M., Donaldson, A.D.

1990-12-12

A method of measuring a molten metal liquid pool volume and in particular molten titanium liquid pools, including the steps of (a) generating an ultrasonic wave at the surface of the molten metal liquid pool, (b) shining a light on the surface of a molten metal liquid pool, (c) detecting a change in the frequency of light, (d) detecting an ultrasonic wave echo at the surface of the molten metal liquid pool, and (e) computing the volume of the molten metal liquid. 3 figs.

Method of measuring a liquid pool volume

DOEpatents

Garcia, G.V.; Carlson, N.M.; Donaldson, A.D.

1991-03-19

A method of measuring a molten metal liquid pool volume and in particular molten titanium liquid pools is disclosed, including the steps of (a) generating an ultrasonic wave at the surface of the molten metal liquid pool, (b) shining a light on the surface of a molten metal liquid pool, (c) detecting a change in the frequency of light, (d) detecting an ultrasonic wave echo at the surface of the molten metal liquid pool, and (e) computing the volume of the molten metal liquid. 3 figures.

Method of measuring a liquid pool volume

DOEpatents

Garcia, Gabe V.; Carlson, Nancy M.; Donaldson, Alan D.

1991-01-01

A method of measuring a molten metal liquid pool volume and in particular molten titanium liquid pools, including the steps of (a) generating an ultrasonic wave at the surface of the molten metal liquid pool, (b) shining a light on the surface of a molten metal liquid pool, (c) detecting a change in the frequency of light, (d) detecting an ultrasonic wave echo at the surface of the molten metal liquid pool, and (e) computing the volume of the molten metal liquid.

Biofilm reactors for industrial bioconversion processes: employing potential of enhanced reaction rates

PubMed Central

Qureshi, Nasib; Annous, Bassam A; Ezeji, Thaddeus C; Karcher, Patrick; Maddox, Ian S

2005-01-01

This article describes the use of biofilm reactors for the production of various chemicals by fermentation and wastewater treatment. Biofilm formation is a natural process where microbial cells attach to the support (adsorbent) or form flocs/aggregates (also called granules) without use of chemicals and form thick layers of cells known as "biofilms." As a result of biofilm formation, cell densities in the reactor increase and cell concentrations as high as 74 gL-1 can be achieved. The reactor configurations can be as simple as a batch reactor, continuous stirred tank reactor (CSTR), packed bed reactor (PBR), fluidized bed reactor (FBR), airlift reactor (ALR), upflow anaerobic sludge blanket (UASB) reactor, or any other suitable configuration. In UASB granular biofilm particles are used. This article demonstrates that reactor productivities in these reactors have been superior to any other reactor types. This article describes production of ethanol, butanol, lactic acid, acetic acid/vinegar, succinic acid, and fumaric acid in addition to wastewater treatment in the biofilm reactors. As the title suggests, biofilm reactors have high potential to be employed in biotechnology/bioconversion industry for viable economic reasons. In this article, various reactor types have been compared for the above bioconversion processes. PMID:16122390

Comparison of individual and pooled stool samples for the assessment of intensity of Schistosoma mansoni and soil-transmitted helminth infections using the Kato-Katz technique.

PubMed

Kure, Ashenafi; Mekonnen, Zeleke; Dana, Daniel; Bajiro, Mitiku; Ayana, Mio; Vercruysse, Jozef; Levecke, Bruno

2015-09-24

Our group has recently provided a proof-of-principle for the examination of pooled stool samples using McMaster technique as a strategy for the rapid assessment of intensity of soil-transmitted helminth infections (STH, Ascaris lumbricoides, Trichuris trichiura and hookworm). In the present study we evaluated this pooling strategy for the assessment of intensity of both STH and Schistosoma mansoni infections using the Kato-Katz technique. A cross-sectional survey was conducted in 360 children aged 5-18 years from six schools in Jimma Zone (southwest Ethiopia). We performed faecal egg counts (FECs) in both individual and pooled samples (pools sizes of 5, 10 and 20) to estimate the number of eggs per gram of stool (EPG) using the Kato-Katz technique. We also assessed the time to screen both individual and pooled samples. Except for hookworms, there was a significant correlation (correlation coefficient = 0.53-0.95) between the mean of individual FECs and the FECs of pooled samples for A. lumbricoides, T. trichiura and S. mansoni, regardless of the pool size. Mean FEC were 2,596 EPG, 125 EPG, 47 EPG, and 41 EPG for A. lumbricoides, T. trichiura, S. mansoni and hookworm, respectively. There was no significant difference in FECs between the examination of individual and pooled stool samples, except for hookworms. For this STH, pools of 10 resulted in a significant underestimation of infection intensity. The total time to obtain individual FECs was 65 h 5 min. For pooled FECs, this was 19 h 12 min for pools of 5, 14 h 39 min for pools of 10 and 12 h 42 min for pools of 20. The results indicate that pooling of stool sample holds also promise as a rapid assessment of infections intensity for STH and S. mansoni using the Kato-Katz technique. In this setting, the time in the laboratory was reduced by 70 % when pools of 5 instead of individual stool samples were screened.

What’s in the Pool? A Comprehensive Identification of Disinfection By-Products and Assessment of Mutagenicity of Chlorinated and Brominated Swimming Pool Water

EPA Science Inventory

Swimming pool disinfectants and disinfection by-products (DBPs) have been linked to human health effects, including asthma and bladder cancer, but no studies have provided a comprehensive identification of DBPs in pool water and related those DBPs to the mutagenicity of pool wate...

Measure Guideline. Replacing Single-Speed Pool Pumps with Variable Speed Pumps for Energy Savings

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

Hunt, A.; Easley, S.

2012-05-01

This measure guideline evaluates potential energy savings by replacing traditional single-speed pool pumps with variable speed pool pumps, and provides a basic cost comparison between continued uses of traditional pumps verses new pumps. A simple step-by-step process for inspecting the pool area and installing a new pool pump follows.

Measure Guideline: Replacing Single-Speed Pool Pumps with Variable Speed Pumps for Energy Savings

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

Hunt, A.; Easley, S.

2012-05-01

The report evaluates potential energy savings by replacing traditional single-speed pool pumps with variable speed pool pumps, and provide a basic cost comparison between continued uses of traditional pumps verses new pumps. A simple step-by-step process for inspecting the pool area and installing a new pool pump follows.

7 CFR 1425.17 - Eligible commodity and pooling.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Eligible commodity and pooling. (a) A CMA may establish separate loan pools as needed for quantities of a... loan pool as provided in paragraph (e) of this section and the beneficial interest provisions of parts 1421 and 1427 of this chapter. (c) A pool shall be eligible for loans and LDP's if: (1) All of the...

Numerical simulation of ground-water flow in La Crosse County, Wisconsin, and into nearby pools of the Mississippi River

USGS Publications Warehouse

Hunt, Randall J.; Saad, David A.; Chapel, Dawn M.

2003-01-01

The models provide estimates of the locations and amount of ground-water flow into Pool 8 and the southern portion of Pool 7 of the Mississippi River. Ground-water discharges into all areas of the pools, except along the eastern shore in the vicinity of the city of La Crosse and immediately downgradient from lock and dam 7 and 8. Ground-water flow into the pools is generally greatest around the perimeter with decreasing amounts away from the perimeter. An area of relatively high ground-water discharge extends out towards the center of Pool 7 from the upper reaches of the pool and may

Certified: A Citizen's Step-by-Step Guide to Protecting Vernal Pools. 6th Edition.

ERIC Educational Resources Information Center

Colburn, Elizabeth A., Ed.

The objective of this manual is to train volunteers in the process of identifying vernal pool habitat so that as many of these pools as possible can be certified by the Massachusetts Natural Heritage and Endangered Species Program. Vernal pools are a kind of temporary pond in which salamanders and other animals breed. The pools are unique…

The impact of urbanization on water and sediment chemistry of ephemeral forest pools

Treesearch

Robert T. Brooks; Suzanne D. Miller; John Newsted

2002-01-01

We compared the water and sediment composition of two ephemeral pools located in forested settings in a developed suburban area with two similar pools located in extensive forest in Massachusetts. We also compared the macroinvertebrate communities. The sediments of the forest pools were 100% organic material, while those of the urban pools were predominantly silt. The...

Livestock Grazing as a Driver of Vernal Pool Ecohydrology

NASA Astrophysics Data System (ADS)

Michaels, J.; McCarten, N. F.

2017-12-01

Vernal pools are seasonal wetlands that host rare plant communities of high conservation priority. Plant community composition is largely driven by pool hydroperiod. A previous study found that vernal pools grazed by livestock had longer hydroperiods compared with pools excluded from grazing for 10 years, and suggests that livestock grazing can be used to protect plant diversity. It is important to assess whether observed differences are due to the grazing or due to water balance variables including upland discharge into or out of the pools since no a priori measurements were made of the hydrology prior to grazing. To address this question, in 2016 we compared 15 pools that have been grazed continuously and 15 pools that have been fenced off for over 40 years at a site in Sacramento County. We paired pools based on abiotic characteristics (size, shape, slope, soil type) to minimize natural variation. We sampled vegetation and water depth using Solinst level loggers. We found that plant diversity and average hydroperiod was significantly higher in the grazed pools. We are currently measuring groundwater connectivity and upland inputs in order to compare the relative strength of livestock grazing as a driver of hydroperiod to these other drivers.

Morphology of drying blood pools

NASA Astrophysics Data System (ADS)

Laan, Nick; Smith, Fiona; Nicloux, Celine; Brutin, David; D-Blood project Collaboration

2016-11-01

Often blood pools are found on crime scenes providing information concerning the events and sequence of events that took place on the scene. However, there is a lack of knowledge concerning the drying dynamics of blood pools. This study focuses on the drying process of blood pools to determine what relevant information can be obtained for the forensic application. We recorded the drying process of blood pools with a camera and measured the weight. We found that the drying process can be separated into five different: coagulation, gelation, rim desiccation, centre desiccation, and final desiccation. Moreover, we found that the weight of the blood pool diminishes similarly and in a reproducible way for blood pools created in various conditions. In addition, we verify that the size of the blood pools is directly related to its volume and the wettability of the surface. Our study clearly shows that blood pools dry in a reproducible fashion. This preliminary work highlights the difficult task that represents blood pool analysis in forensic investigations, and how internal and external parameters influence its dynamics. We conclude that understanding the drying process dynamics would be advancement in timeline reconstitution of events. ANR funded project: D-Blood Project.

Assessing ecological risk at a hazardous waste site containing vernal pools

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

DeLong, T.; Millard, J.; Timmer, E.

1995-12-31

An ecological risk assessment was conducted for a Superfund site in central California. As part of this assessment an evaluation of vernal pools was conducted. Vernal pools are amphibious ecosystems that support unique biotic communities. Many of the endemic species associated with vernal pools in central California are currently listed as state or Federally endangered, threatened, or rare species and include: Contra Costa goldfields (Lasthenia conjugens), vernal pool fairy shrimp (Branchinecta lynchl), vernal pool tadpole shrimp (Lepidurus packardi) and the California tiger salamander (Ambystoma tigrinum califomiense). The protection of these habitats is essential for the preservation of the special statusmore » species dependent on them for survival. As part of the risk assessment, vernal pools in the study area were identified and surveyed for special status flora and fauna for two consecutive years. Information regarding the relative quality of each pool was also collected. In order to assess potential impacts from chemical exposures to communities inhabiting these vernal pools, a weight-of-evidence approach was employed that included: evaluation of vernal pool biological composition; assessment of physical and chemical conditions; invertebrate sediment toxicity evaluations, and Frog Embryo Teratogenesis Analysis -- Xenopus (FETAX) testing.« less

Vernal Pools Detection Using High-Resolution LiDAR Data and Aerial Imagery in Hubbardston, Massachusetts

NASA Astrophysics Data System (ADS)

Jiang, Jiaxin

Vernal pool refers to temporary or semi-permanent pools that occur in surface depressions without permanent inlets or outlets. Because they periodically dry out, vernal pools are free of fish and essential to amphibians, some reptiles, birds, and mammals for breeding habitats. In Massachusetts, vernal pool habitats are found in woodland depressions, swales or kettle holes where water is contained for at least two months in most years. However, vernal pools are delicate ecosystems. These systems are fragile to human activities such as urbanization. Understanding the current situation of vernal pools helps city planners make wiser decisions. This study focuses on identifying vernal pools in the state of Massachusetts with high-resolution light detection and ranging (LiDAR) data and aerial imagery. By using high-resolution light detection and ranging data, aerial imagery, land use data, the MassDEP Hydrography layer and the Soil Survey Geographic Database, the approach located over 1800 potential vernal pools in a 108 km 2 study area in Massachusetts. The assessment of the study result shows the commission rate was 5.6% and omission rate was 7.1%. This method provides an efficient way of locating vernal pools over large areas.

Information content of incubation experiments for inverse estimation of pools in the Rothamsted carbon model: a Bayesian perspective

NASA Astrophysics Data System (ADS)

Scharnagl, B.; Vrugt, J. A.; Vereecken, H.; Herbst, M.

2010-02-01

A major drawback of current soil organic carbon (SOC) models is that their conceptually defined pools do not necessarily correspond to measurable SOC fractions in real practice. This not only impairs our ability to rigorously evaluate SOC models but also makes it difficult to derive accurate initial states of the individual carbon pools. In this study, we tested the feasibility of inverse modelling for estimating pools in the Rothamsted carbon model (ROTHC) using mineralization rates observed during incubation experiments. This inverse approach may provide an alternative to existing SOC fractionation methods. To illustrate our approach, we used a time series of synthetically generated mineralization rates using the ROTHC model. We adopted a Bayesian approach using the recently developed DiffeRential Evolution Adaptive Metropolis (DREAM) algorithm to infer probability density functions of the various carbon pools at the start of incubation. The Kullback-Leibler divergence was used to quantify the information content of the mineralization rate data. Our results indicate that measured mineralization rates generally provided sufficient information to reliably estimate all carbon pools in the ROTHC model. The incubation time necessary to appropriately constrain all pools was about 900 days. The use of prior information on microbial biomass carbon significantly reduced the uncertainty of the initial carbon pools, decreasing the required incubation time to about 600 days. Simultaneous estimation of initial carbon pools and decomposition rate constants significantly increased the uncertainty of the carbon pools. This effect was most pronounced for the intermediate and slow pools. Altogether, our results demonstrate that it is particularly difficult to derive reasonable estimates of the humified organic matter pool and the inert organic matter pool from inverse modelling of mineralization rates observed during incubation experiments.

The impact on accuracy and cost of ligase chain reaction testing by pooling urine specimens for the diagnosis of Chlamydia trachomatis infections.

PubMed

Krepel, J; Patel, J; Sproston, A; Hopkins, F; Jang, D; Mahony, J; Chernesky, M

1999-10-01

Nucleic acid amplification testing is the most accurate approach to diagnosing Chlamydia trachomatis infections. Our objective was to compare the accuracy and cost savings of pooling urines as opposed to individual testing. Strategies of pooling urine specimens into groups of four (4x pool) or eight (8x pool) followed by testing the positive pools individually were compared to individual specimen testing to determine if significant cost savingS could be realized without compromising the sensitivity and specificity of the LCx C. trachomatis Assay (Abbott Laboratories, Abbott Park, Chicago, IL) performed in a busy private medical laboratory. A total of 1,220 patient urine samples, 1,187 male (97%) and 33 female (3%), were tested using the normal LCx specimen to cutoff ratio (S/CO) of 1.0 and a decreased S/CO value of 0.2. Individual testing identified 98.2% (109/111) of positive urines. The 4x pooling maneuver identified 92.8% (103/111) of positive patients with the regular cutoff and 96.4% (107/111) when the cutoff was decreased. These values were 95.9% (47/49) and 97.9% (48/49), respectively, when eight urines were pooled. Both pooling and individual testing strategies identified all the negative samples accurately. Cost savings of pooling were calculated to be 44.5% for pools of four and 37.5% for pools of eight, applying the lowered cutoff. Pooling urine specimens for testing with the C. trachomatis LCx system is a simple, accurate, and cost-saving approach that can significantly reduce the cost of amplified nucleic acid testing with minimal sacrifice of testing accuracy.

Deep Trouble.

ERIC Educational Resources Information Center

Fawcett, Paul

1997-01-01

Discusses how accident prevention can be built into the swimming-pool design phase by paying attention to swimming-pool regulations, materials for basin and deck construction, pool-fixture placement, and signs and markings. A pool planning checklist is provided. (GR)

Evidence for two distinct intracellular pools of inorganic sulfate in Penicillium notatum

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

Hunter, D.R.; Segel, I.H.

1985-06-01

A strain of Penicillium notatum unable to metabolize inorganic sulfate can accumulate sulfate internally to an apparent equilibrium concentration 10/sup 5/ times greater than that remaining in the medium. The apparent K/sub eq/ is near constant at all initial external sulfate concentrations below that which would eventually exceed the internal capacity of the cells. Under equilibrium conditions of zero net flux, external /sup 35/SO/sub 4//sup 2 -/ exchanges with internal, unlabeled SO/sub 4//sup 2 -/ at a rate consistent with the kinetic constants with the sulfate transport system. Efflux experiments demonstrated that sulfate occupies two distinct intracellular pools. Pool 1more » is characterized by the rapid release of /sup 35/SO/sub 4//sup 2 -/ when the suspension of preloaded cells is adjusted to 10 mM azide at pH 8.4 (t/sub 1/2/, 0.38 min). /sup 35/SO/sub 4//sup 2 -/ in pool 1 also rapidly exchanges with unlabeled medium sulfate. Pool 2 is characterized by the slow release of /sup 35/SO/sub 4//sup 2 -/ induced by azide at pH 8.4 or unlabeled sulfate (t/sub 1/2/, 32 to 49 min). Early in the /sup 35/SO/sub 4//sup 2 -/ accumulation process, up to 78% of the total transported substrate is found in pool 1. At equilibrium, pool 1 accounts for only about 2% of the total accumulated /sup 35/SO/sub 4//sup 2 -/. Monensin (33 ..mu..m) accelerates the transfer of /sup 35/SO/sub 4//sup 2 -/ from pool 1 to pool 2. Valinomycin (0.2 ..mu..M) and tetraphynylboron/sup -/ (1 mM) retard the transfer of /sup 35/SO/sub 4//sup 2 -/ from pool 1 to pool 2. Pool 2 may reside in a vacuole or other intracellular organelle. A model for the transfer of sulfate from pool 1 to pool 2 is presented.« less

Comparison of Individual and Pooled Stool Samples for the Assessment of Soil-Transmitted Helminth Infection Intensity and Drug Efficacy

PubMed Central

Mekonnen, Zeleke; Meka, Selima; Ayana, Mio; Bogers, Johannes; Vercruysse, Jozef; Levecke, Bruno

2013-01-01

Background In veterinary parasitology samples are often pooled for a rapid assessment of infection intensity and drug efficacy. Currently, studies evaluating this strategy in large-scale drug administration programs to control human soil-transmitted helminths (STHs; Ascaris lumbricoides, Trichuris trichiura, and hookworm), are absent. Therefore, we developed and evaluated a pooling strategy to assess intensity of STH infections and drug efficacy. Methods/Principal Findings Stool samples from 840 children attending 14 primary schools in Jimma, Ethiopia were pooled (pool sizes of 10, 20, and 60) to evaluate the infection intensity of STHs. In addition, the efficacy of a single dose of mebendazole (500 mg) in terms of fecal egg count reduction (FECR; synonym of egg reduction rate) was evaluated in 600 children from two of these schools. Individual and pooled samples were examined with the McMaster egg counting method. For each of the three STHs, we found a significant positive correlation between mean fecal egg counts (FECs) of individual stool samples and FEC of pooled stool samples, ranging from 0.62 to 0.98. Only for A. lumbricoides was any significant difference in mean FEC of the individual and pooled samples found. For this STH species, pools of 60 samples resulted in significantly higher FECs. FECR for the different number of samples pooled was comparable in all pool sizes, except for hookworm. For this parasite, pools of 10 and 60 samples provided significantly higher FECR results. Conclusion/Significance This study highlights that pooling stool samples holds promise as a strategy for rapidly assessing infection intensity and efficacy of administered drugs in programs to control human STHs. However, further research is required to determine when and how pooling of stool samples can be cost-effectively applied along a control program, and to verify whether this approach is also applicable to other NTDs. PMID:23696905

Non-stationary Drainage Flows and Cold Pools in Gentle Terrain

NASA Astrophysics Data System (ADS)

Mahrt, L.

2015-12-01

Previous studies have concentrated on organized topography with well-defined slopes or valleys in an effort to understand the flow dynamics. However, most of the Earth's land surface consists of gentle terrain that is quasi three dimensional. Different scenarios are briefly classified. A network of measurements are analyzed to examine shallow cold pools and drainage flow down the valley which develop for weak ambient wind and relatively clear skies. However, transient modes constantly modulate or intermittently eliminate the cold pool, which makes extraction and analysis of the horizontal structure of the cold pool difficult with traditional analysis methods. Singular value decomposition successfully isolates the effects of large-scale flow from local down-valley cold air drainage within the cold pool in spite of the intermittent nature of this local flow. The traditional concept of a cold pool must be generalized to include cold pool intermittency, complex variation of temperature related to some three-dimensionality and a diffuse cold pool top. Different types of cold pools are classified in terms of the stratification and gradient of potential temperature along the slope. The strength of the cold pool is related to a forcing temperature scale proportional to the net radiative cooling divided by the wind speed above the valley. The scatter is large partly due to nonstationarity of the marginal cold pool in this shallow valley

Science communication and vernal pool conservation: a study of local decision maker attitudes in a knowledge-action system.

PubMed

McGreavy, Bridie; Webler, Thomas; Calhoun, Aram J K

2012-03-01

In this study, we describe local decision maker attitudes towards vernal pools to inform science communication and enhance vernal pool conservation efforts. We conducted interviews with town planning board and conservation commission members (n = 9) from two towns in the State of Maine in the northeastern United States. We then mailed a questionnaire to a stratified random sample of planning board members in August and September 2007 with a response rate of 48.4% (n = 320). The majority of survey respondents favored the protection and conservation of vernal pools in their towns. Decision makers were familiar with the term "vernal pool" and demonstrated positive attitudes to vernal pools in general. General appreciation and willingness to conserve vernal pools predicted support for the 2006 revisions to the Natural Resource Protection Act regulating Significant Vernal Pools. However, 48% of respondents were unaware of this law and neither prior knowledge of the law nor workshop attendance predicted support for the vernal pool law. Further, concerns about private property rights and development restrictions predicted disagreement with the vernal pool law. We conclude that science communication must rely on specific frames of reference, be sensitive to cultural values, and occur in an iterative system to link knowledge and action in support of vernal pool conservation. Copyright © 2011 Elsevier Ltd. All rights reserved.

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.

Design of inventory pools in spare part support operation systems

NASA Astrophysics Data System (ADS)

Mo, Daniel Y.; Tseng, Mitchell M.; Cheung, Raymond K.

2014-06-01

The objective of a spare part support operation is to fulfill the part request order with different service contracts in the agreed response time. With this objective to achieve different service targets for multiple service contracts and the considerations of inventory investment, it is not only important to determine the inventory policy but also to design the structure of inventory pools and the order fulfilment strategies. In this research, we focused on two types of inventory pools: multiple inventory pool (MIP) and consolidated inventory pool (CIP). The idea of MIP is to maintain separated inventory pools based on the types of service contract, while CIP solely maintains a single inventory pool regardless of service contract. Our research aims to design the inventory pool analytically and propose reserve strategies to manage the order fulfilment risks in CIP. Mathematical models and simulation experiments would be applied for analysis and evaluation.

Hospital hydrotherapy pools treated with ultra violet light: bad bacteriological quality and presence of thermophilic Naegleria.

PubMed Central

De Jonckheere, J. F.

1982-01-01

The microbiological quality of eight halogenated and two u.v.-treated hydrotherapy pools in hospitals was investigated. The microbiological quality of halogenated hydrotherapy pools was comparable to halogenated public swimming pools, although in some Pseudomonas aeruginosa and faecal pollution indicators were more frequent due to bad management. On the other hand u.v.-treated hydrotherapy pools had very bad microbiological quality. Apart from faecal pollution indicators, P. aeruginosa was present in very high numbers. Halogenated hydrotherapy pools were not highly contaminated with amoebae, and Naegleria spp. were never detected. On the other hand u.v.-treated pools contained very high numbers of thermophilic Naegleria. The Naegleria isolated were identified as N. lovaniensis, a species commonly found in association with N. fowleri. Isoenzyme analysis showed a different type of N. lovaniensis was present in each of two u.v.-treated pools. Images Plate 1 PMID:7061835

Characteristics of pools used by adult summer steelhead oversummering in the New River, California

Treesearch

Rodney J. Nakamoto

1994-01-01

Abstract - I assessed characteristics of pools used by oversummering adults of summer steelhead Oncorhynchus mykiss between July and October 1991 in the New River, northwestern California. Most fish occupied channel confluence pools and other pools of moderate size (200-1,200 m 2); these pools had less than 35% substrate embeddedness and mean water depths of about 1.0...

Biogenic silicon pools in terrestrial biogeosystems and their significance for silicon cycling

NASA Astrophysics Data System (ADS)

Puppe, Daniel; Höhn, Axel; Kaczorek, Danuta; Ehrmann, Otto; Wanner, Manfred; Sommer, Michael

2017-04-01

On a global scale the biogeochemical cycles of silicon (Si) and carbon are connected by weathering processes and fluxes of dissolved Si from terrestrial to aquatic ecosystems. Various pro- and eukaryotic organisms are evolutionarily adapted to synthesize amorphous siliceous structures (biosilicification). In soils these siliceous structures can accumulate and form biogenic Si (BSi) pools, whereat it can be differentiated between phytogenic (BSi synthesized by plants), zoogenic (BSi synthesized by sponges), microbial (BSi synthesized by bacteria and fungi) and protistic (BSi synthesized by unicellular organisms) pools. Accumulation and recycling of BSi in terrestrial biogeosystems influence fluxes of dissolved Si from the continents to the oceans, thus act as a filter in the global Si cycle. As research has primarily been focused on the role of phytogenic Si pools until now there is only little information available on the other BSi pools in soils. In order to fill this knowledge gap we examined different BSi pools in soils of initial and forested terrestrial biogeosystems using modern microscopical methods (laser scanning and scanning electron microscopy). In forested biogeosystems we further analyzed abiotic (e.g. soil pH) and biotic (earthworm biomasses) influencing factors on BSi pool size, while samples of initial biogeosystems were used to analyze spatiotemporal BSi pool dynamics. We found that especially biotic interactions are important factors for protistic BSi pools (represented by testate amoebae) and that phytogenic Si pools are about several 100-times bigger than protistic (testate amoebae) Si pools (0.2-4.7 kg Si ha-1). However, annual biosilicification rates of testate amoebae (up to 80 kg Si ha-1) are comparable to or even can exceed annual silicon uptake by trees. Our studies of initial biogeosystems revealed that BSi pool sizes increased markedly within a relatively short time span (<10 years) of ecosystem development. Differences in quantities, dynamics and resistibility against dissolution of various BSi pools indicated their possibility to influence biogeochemical Si cycling relatively rapid (protistic Si pools) or slow (zoogenic Si pools). In conclusion, our results are crucial for a detailed understanding and a more precise modeling of Si fluxes from terrestrial to aquatic ecosystems.

Pool Formation in Boulder-Bed Streams: Implications From 1-D and 2-D Numerical Modeling

NASA Astrophysics Data System (ADS)

Harrison, L. R.; Keller, E. A.

2003-12-01

In mountain rivers of Southern California, boulder-large roughness elements strongly influence flow hydraulics and pool formation and maintenance. In these systems, boulders appear to control the stream morphology by converging flow and producing deep pools during channel forming discharges. Our research goal is to develop quantitative relationships between boulder roughness elements, temporal patterns of scour and fill, and geomorphic processes that are important in producing pool habitat. The longitudinal distribution of shear stress, unit stream power and velocity were estimated along a 48 m reach on Rattlesnake Creek, using the HEC-RAS v 3.0 and River 2-D numerical models. The reach has an average slope of 0.02 and consists of a pool-riffle sequence with a large boulder constriction directly above the pool. Model runs were performed for a range of stream discharges to test if scour and fill thresholds for pool and riffle environments could be identified. Results from the HEC-RAS simulations identified that thresholds in shear stress, unit stream power and mean velocity occur above a discharge of 5.0 cms. Results from the one-dimensional analysis suggest that the reversal in competency is likely due to changes in cross-sectional width at varying flows. River 2-D predictions indicated that strong transverse velocity gradients were present through the pool at higher modeled discharges. At a flow of 0.5 cms (roughly 1/10th bankfull discharge), velocities are estimated at 0.6 m/s and 1.3 m/s for the pool and riffle, respectively. During discharges of 5.15 cms (approximate bankfull discharge), the maximum velocity in the pool center increased to nearly 3.0 m/s, while the maximum velocity over the riffle is estimated at approximately 2.5 cms. These results are consistent with those predicted by HEC-RAS, though the reversal appears to be limited to a narrow jet that occurs through the pool head and pool center. Model predictions suggest that the velocity reversal is produced by a boulder-bedrock constriction that rapidly decreases the channel width above the pool by roughly 25 percent. The width constriction creates highly turbulent flow capable of scouring bed material through the pool. The high velocity core that is produced through the pool center appears to be enhanced by the formation of a large eddy directly below the boulder. Values of unit stream power and shear stress indicate that the pool exit is an area of deposition of bed material due to a decrease in tractive force. The presence of a strong transverse velocity gradient suggests that only a portion of the flow is responsible for scouring bed material. After we eliminate the dead water zone, the lowest five percent of the velocity range, patterns of effective width between pools and riffles begin to emerge. The ratio of flow width between adjacent pools and riffles is one measure of flow convergence. At a discharge of 0.5 cms, the ratio of effective width between pools and riffles is roughly 1:1, implying that there is uniform flow with little flow convergence. At a discharge of 5.15 cms the width ratio between the pool and riffle is about 1:3, demonstrating the strong convergent flow patterns at the pool head. The observed effective width relationship suggests that when considering restoration designs, boulders should be placed in areas that replicate natural convergence and divergence patterns in order to maximize pool area and depth.

Refuge habitats for fishes during seasonal drying in an intermittent stream: movement, survival and abundance of three minnow species

USGS Publications Warehouse

Hodges, S.W.; Magoulick, Daniel D.

2011-01-01

Drought and summer drying can be important disturbance events in many small streams leading to intermittent or isolated habitats. We examined what habitats act as refuges for fishes during summer drying, hypothesizing that pools would act as refuge habitats. We predicted that during drying fish would show directional movement into pools from riffle habitats, survival rates would be greater in pools than in riffles, and fish abundance would increase in pool habitats. We examined movement, survival and abundance of three minnow species, bigeye shiner (Notropis boops), highland stoneroller (Campostoma spadiceum) and creek chub (Semotilus atromaculatus), during seasonal stream drying in an Ozark stream using a closed robust multi-strata mark-recapture sampling. Population parameters were estimated using plausible models within program MARK, where a priori models are ranked using Akaike's Information Criterion. Creek chub showed directional movement into pools and increased survival and abundance in pools during drying. Highland stonerollers showed strong directional movement into pools and abundance increased in pools during drying, but survival rates were not significantly greater in pools than riffles. Bigeye shiners showed high movement rates during drying, but the movement was non-directional, and survival rates were greater in riffles than pools. Therefore, creek chub supported our hypothesis and pools appear to act as refuge habitats for this species, whereas highland stonerollers partly supported the hypothesis and bigeye shiners did not support the pool refuge hypothesis. Refuge habitats during drying are species dependent. An urgent need exists to further understand refuge habitats in streams given projected changes in climate and continued alteration of hydrological regimes.

Microbiological Analysis in Three Diverse Natural Geothermal Bathing Pools in Iceland

PubMed Central

Thorolfsdottir, Berglind Osk Th.; Marteinsson, Viggo Thor

2013-01-01

Natural thermal bathing pools contain geothermal water that is very popular to bathe in but the water is not sterilized, irradiated or treated in any way. Increasing tourism in Iceland will lead to increasing numbers of bath guests, which can in turn affect the microbial flora in the pools and therefore user safety. Today, there is no legislation that applies to natural geothermal pools in Iceland, as the water is not used for consumption and the pools are not defined as public swimming pools. In this study, we conducted a microbiological analysis on three popular but different natural pools in Iceland, located at Lýsuhóll, Hveravellir and Landmannalaugar. Total bacterial counts were performed by flow cytometry, and with plate count at 22 °C, 37 °C and 50 °C. The presence of viable coliforms, Enterococcus spp. and pseudomonads were investigated by growth experiments on selective media. All samples were screened for noroviruses by real time PCR. The results indicate higher fecal contamination in the geothermal pools where the geothermal water flow was low and bathing guest count was high during the day. The number of cultivated Pseudomonas spp. was high (13,000–40,000 cfu/100 mL) in the natural pools, and several strains were isolated and classified as opportunistic pathogens. Norovirus was not detected in the three pools. DNA was extracted from one-liter samples in each pool and analyzed by partial 16S rRNA gene sequencing. Microbial diversity analysis revealed different microbial communities between the pools and they were primarily composed of alpha-, beta- and gammaproteobacteria. PMID:23493033

First United States Outbreak of Mycobacterium abscessus Hand and Foot Disease Among Children Associated With a Wading Pool.

PubMed

Carter, Kris K; Lundgren, Ingrid; Correll, Sarah; Schmalz, Tom; McCarter, Tammie; Stroud, Joshua; Bruesch, Amanda; Hahn, Christine G

2018-05-29

Mycobacterium abscessus, an emerging pathogen in healthcare settings, has rarely been associated with community outbreaks. During February-May 2013, Idaho public health officials and pediatric infectious disease physicians investigated an outbreak of M abscessus skin infections in children whose only common exposure was an indoor wading pool. Healthcare providers and parents reported possible M abscessus cases. We used a standardized questionnaire to interview parents of affected children. Clinical specimens were submitted for mycobacterial examination. We conducted an environmental investigation of the pool. Microbial isolates from clinical and environmental samples were identified by sequencing polymerase chain reaction amplicons and underwent pulsed-field gel electrophoresis. Twelve cases were identified. Specimens from 4 of 7 children grew M abscessus or Mycobacterium abscessus/Mycobacterium chelonae . Ten (83%) of 12 children were female; median age was 3 years (range, 2 to 6 years); and all were immunocompetent. Pool maintenance did not fully comply with Idaho state rules governing pool operation. Mycobacterium abscessus/chelonae was isolated from pool equipment. Pulsed-field gel electrophoresis composite patterns were 87% similar between isolates from the pool ladder and 1 patient, and they were 90% similar between isolates from 2 patients. Environmental remediation included hyperchlorination, scrubbing and disinfection of pool surfaces, draining the pool, and replacement of worn pool materials. Immunocompetent children acquired M abscessus cutaneous infection involving hands and feet after exposure to a wading pool. Environmental remediation and proper pool maintenance likely halted transmission. Medical and public health professionals' collaboration effectively detected and controlled an outbreak caused by an emerging recreational waterborne pathogen.

Between-Habitat Variation of Benthic Cover, Reef Fish Assemblage and Feeding Pressure on the Benthos at the Only Atoll in South Atlantic: Rocas Atoll, NE Brazil

PubMed Central

Longo, G. O.; Morais, R. A.; Martins, C. D. L.; Mendes, T. C.; Aued, A. W.; Cândido, D. V.; de Oliveira, J. C.; Nunes, L. T.; Fontoura, L.; Sissini, M. N.; Teschima, M. M.; Silva, M. B.; Ramlov, F.; Gouvea, L. P.; Ferreira, C. E. L.; Segal, B.; Horta, P. A.; Floeter, S. R.

2015-01-01

The Southwestern Atlantic harbors unique and relatively understudied reef systems, including the only atoll in South Atlantic: Rocas atoll. Located 230 km off the NE Brazilian coast, Rocas is formed by coralline red algae and vermetid mollusks, and is potentially one of the most “pristine” areas in Southwestern Atlantic. We provide the first comprehensive and integrative description of the fish and benthic communities inhabiting different shallow reef habitats of Rocas. We studied two contrasting tide pool habitats: open pools, which communicate with the open ocean even during low tides, thus more exposed to wave action; and closed pools, which remain isolated during low tide and are comparatively less exposed. Reef fish assemblages, benthic cover, algal turfs and fish feeding pressure on the benthos remarkably varied between open and closed pools. The planktivore Thalassoma noronhanum was the most abundant fish species in both habitats. In terms of biomass, the lemon shark Negaprion brevirostris and the omnivore Melichtys niger were dominant in open pools, while herbivorous fishes (mainly Acanthurus spp.) prevailed in closed pools. Overall benthic cover was dominated by algal turfs, composed of articulated calcareous algae in open pools and non-calcified algae in closed pools. Feeding pressure was dominated by acanthurids and was 10-fold lower in open pools than in closed pools. Besides different wave exposure conditions, such pattern could also be related to the presence of sharks in open pools, prompting herbivorous fish to feed more in closed pools. This might indirectly affect the structure of reef fish assemblages and benthic communities. The macroalgae Digenea simplex, which is uncommon in closed pools and abundant in the reef flat, was highly preferred in herbivory assays, indicating that herbivory by fishes might be shaping this distribution pattern. The variations in benthic and reef fish communities, and feeding pressure on the benthos between open and closed pools suggest that the dynamics in open pools is mostly driven by physical factors and the tolerance of organisms to harsh conditions, while in closed pools direct and indirect effects of species interactions also play an important role. Understanding the mechanisms shaping biological communities and how they scale-up to ecosystem functioning is particularly important on isolated near-pristine systems where natural processes can still be studied under limited human impact. PMID:26061735

Between-Habitat Variation of Benthic Cover, Reef Fish Assemblage and Feeding Pressure on the Benthos at the Only Atoll in South Atlantic: Rocas Atoll, NE Brazil.

PubMed

Longo, G O; Morais, R A; Martins, C D L; Mendes, T C; Aued, A W; Cândido, D V; de Oliveira, J C; Nunes, L T; Fontoura, L; Sissini, M N; Teschima, M M; Silva, M B; Ramlov, F; Gouvea, L P; Ferreira, C E L; Segal, B; Horta, P A; Floeter, S R

2015-01-01

The Southwestern Atlantic harbors unique and relatively understudied reef systems, including the only atoll in South Atlantic: Rocas atoll. Located 230 km off the NE Brazilian coast, Rocas is formed by coralline red algae and vermetid mollusks, and is potentially one of the most "pristine" areas in Southwestern Atlantic. We provide the first comprehensive and integrative description of the fish and benthic communities inhabiting different shallow reef habitats of Rocas. We studied two contrasting tide pool habitats: open pools, which communicate with the open ocean even during low tides, thus more exposed to wave action; and closed pools, which remain isolated during low tide and are comparatively less exposed. Reef fish assemblages, benthic cover, algal turfs and fish feeding pressure on the benthos remarkably varied between open and closed pools. The planktivore Thalassoma noronhanum was the most abundant fish species in both habitats. In terms of biomass, the lemon shark Negaprion brevirostris and the omnivore Melichtys niger were dominant in open pools, while herbivorous fishes (mainly Acanthurus spp.) prevailed in closed pools. Overall benthic cover was dominated by algal turfs, composed of articulated calcareous algae in open pools and non-calcified algae in closed pools. Feeding pressure was dominated by acanthurids and was 10-fold lower in open pools than in closed pools. Besides different wave exposure conditions, such pattern could also be related to the presence of sharks in open pools, prompting herbivorous fish to feed more in closed pools. This might indirectly affect the structure of reef fish assemblages and benthic communities. The macroalgae Digenea simplex, which is uncommon in closed pools and abundant in the reef flat, was highly preferred in herbivory assays, indicating that herbivory by fishes might be shaping this distribution pattern. The variations in benthic and reef fish communities, and feeding pressure on the benthos between open and closed pools suggest that the dynamics in open pools is mostly driven by physical factors and the tolerance of organisms to harsh conditions, while in closed pools direct and indirect effects of species interactions also play an important role. Understanding the mechanisms shaping biological communities and how they scale-up to ecosystem functioning is particularly important on isolated near-pristine systems where natural processes can still be studied under limited human impact.

Hydrology of the Floral City Pool of Tsala Apopka Lake, west-central Florida

USGS Publications Warehouse

Bradner, L.A.

1988-01-01

Tsala Apopka Lake, in west-central Florida, has an area of about 19,000 acres and is divided into three water-management pools, with the Floral City Pool, the most upgradient. The Floral City Pool, which has a surface area of approximately 4,750 acres, contains an extensive combination of lakes, wetlands, and connecting canals. The Pool receives inflow from the Withlacoochee River through two canals. Outflow is through one manmade canal and one natural slough. Canal flow is partially controlled by manmade structures. A cumulative deficit of 19.4 inches of rainfall from August 1984 through May 1985 reduced surface-water inflow to the Floral City Pool to about 0.5 cu ft/sec by May 1985. During May 1985, pool levels declined approximately 0.04 ft/day. By the end of May, there was no observable outflow. From June 1985 through September 1985, 39.8 inches of rainfall caused above-average inflow to the Floral City Pool and a pool-level increase of 6.2 ft. The inflow of 340 CFS nearly equaled the outflow of 338 CFS by the end of September. (USGS)

Alcoholism Detection by Data Augmentation and Convolutional Neural Network with Stochastic Pooling.

PubMed

Wang, Shui-Hua; Lv, Yi-Ding; Sui, Yuxiu; Liu, Shuai; Wang, Su-Jing; Zhang, Yu-Dong

2017-11-17

Alcohol use disorder (AUD) is an important brain disease. It alters the brain structure. Recently, scholars tend to use computer vision based techniques to detect AUD. We collected 235 subjects, 114 alcoholic and 121 non-alcoholic. Among the 235 image, 100 images were used as training set, and data augmentation method was used. The rest 135 images were used as test set. Further, we chose the latest powerful technique-convolutional neural network (CNN) based on convolutional layer, rectified linear unit layer, pooling layer, fully connected layer, and softmax layer. We also compared three different pooling techniques: max pooling, average pooling, and stochastic pooling. The results showed that our method achieved a sensitivity of 96.88%, a specificity of 97.18%, and an accuracy of 97.04%. Our method was better than three state-of-the-art approaches. Besides, stochastic pooling performed better than other max pooling and average pooling. We validated CNN with five convolution layers and two fully connected layers performed the best. The GPU yielded a 149× acceleration in training and a 166× acceleration in test, compared to CPU.

Towards dropout training for convolutional neural networks.

PubMed

Wu, Haibing; Gu, Xiaodong

2015-11-01

Recently, dropout has seen increasing use in deep learning. For deep convolutional neural networks, dropout is known to work well in fully-connected layers. However, its effect in convolutional and pooling layers is still not clear. This paper demonstrates that max-pooling dropout is equivalent to randomly picking activation based on a multinomial distribution at training time. In light of this insight, we advocate employing our proposed probabilistic weighted pooling, instead of commonly used max-pooling, to act as model averaging at test time. Empirical evidence validates the superiority of probabilistic weighted pooling. We also empirically show that the effect of convolutional dropout is not trivial, despite the dramatically reduced possibility of over-fitting due to the convolutional architecture. Elaborately designing dropout training simultaneously in max-pooling and fully-connected layers, we achieve state-of-the-art performance on MNIST, and very competitive results on CIFAR-10 and CIFAR-100, relative to other approaches without data augmentation. Finally, we compare max-pooling dropout and stochastic pooling, both of which introduce stochasticity based on multinomial distributions at pooling stage. Copyright © 2015 Elsevier Ltd. All rights reserved.

Reactor engineering support of operations at the Davis-Besse nuclear power station

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

Kelley, D.B.

1995-12-31

Reactor engineering functions differ greatly from unit to unit; however, direct support of the reactor operators during reactor startups and operational transients is common to all units. This paper summarizes the support the reactor engineers provide the reactor operators during reactor startups and power changes through the use of automated computer programs at the Davis-Besse nuclear power station.

Cool Water Formation and Trout Habitat Use in a Deep Pool in the Sierra Nevada, California

Treesearch

KATHLEEN R. MATTHEWS; NEIL H. BERG; AZUMA DAVID L.

1994-01-01

We documented temperature stratification in a deep bedrock pool in the North Fork of the American River, described the diel movement of rainbow trout Oncorhynchus mykiss and brown trout Salmo trutta. and determined whether these trout used cooler portions of the pool.From July 30 to October 10, 1992, the main study pool and an adjacent pool were stratified(temperature...

10 CFR Appendix P to Subpart B of... - Uniform Test Method for Measuring the Energy Consumption of Pool Heaters

Code of Federal Regulations, 2012 CFR

2012-01-01

...-1994. The expression of fuel consumption for oil-fired pool heaters shall be in Btu. 4.2Average annual fossil fuel energy for pool heaters. The average annual fuel energy for pool heater, EF, is defined as... of pool operating hours=4464 h QIN=rated fuel energy input as defined according to 2.9.1 or 2.9.2 of...

Experimental and numerical studies on three dimensional GTA weld pool convection: Non-axisymmetric effects

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

Joshi, Y.; Dutta, P.; Schupp, P.E.

1995-12-31

Observations of surface flow patterns of steel and aluminum GTAW pools have been made using a pulsed laser visualization system. The weld pool convection is found to be three dimensional, with the azimuthal circulation depending on the location of the clamp with respect to the torch. Oscillation of steel pools and undulating motion in aluminum weld pools are also observed even with steady process parameters. Current axisymmetric numerical models are unable to explain such phenomena. A three dimensional computational study is carried out in this study to explain the rotational flow in aluminum weld pools.

Tactical reproductive parasitism via larval cannibalism in Peruvian poison frogs

PubMed Central

Brown, Jason L.; Morales, Victor; Summers, Kyle

2008-01-01

We report an unusual example of reproductive parasitism in amphibians. Dendrobates variabilis, an Amazonian poison frog, oviposits at the surface of the water in small pools in plants and deposits tadpoles within the pools. Tadpoles are highly cannibalistic and consume young tadpoles if they are accessible. Deposition of embryos and tadpoles in the same pool is common. Genetic analyses indicate that tadpoles are frequently unrelated to embryos in the same pool. A pool choice experiment in the field demonstrated that males carrying tadpoles prefer to place them in pools with embryos, facilitating reproductive parasitism via cannibalism. PMID:19042178

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2011 CFR

2011-01-01

... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director...

Nuclear Reactors. Revised.

ERIC Educational Resources Information Center

Hogerton, John F.

This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2012 CFR

2012-01-01

... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director...

Strong influence of regional species pools on continent-wide structuring of local communities.

PubMed

Lessard, Jean-Philippe; Borregaard, Michael K; Fordyce, James A; Rahbek, Carsten; Weiser, Michael D; Dunn, Robert R; Sanders, Nathan J

2012-01-22

There is a long tradition in ecology of evaluating the relative contribution of the regional species pool and local interactions on the structure of local communities. Similarly, a growing number of studies assess the phylogenetic structure of communities, relative to that in the regional species pool, to examine the interplay between broad-scale evolutionary and fine-scale ecological processes. Finally, a renewed interest in the influence of species source pools on communities has shown that the definition of the source pool influences interpretations of patterns of community structure. We use a continent-wide dataset of local ant communities and implement ecologically explicit source pool definitions to examine the relative importance of regional species pools and local interactions for shaping community structure. Then we assess which factors underlie systematic variation in the structure of communities along climatic gradients. We find that the average phylogenetic relatedness of species in ant communities decreases from tropical to temperate regions, but the strength of this relationship depends on the level of ecological realism in the definition of source pools. We conclude that the evolution of climatic niches influences the phylogenetic structure of regional source pools and that the influence of regional source pools on local community structure is strong.

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

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

D. Kokkinos

2005-04-28

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophymore » on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.« less

Heat transfer analysis of cylindrical anaerobic reactors with different sizes: a heat transfer model.

PubMed

Liu, Jiawei; Zhou, Xingqiu; Wu, Jiangdong; Gao, Wen; Qian, Xu

2017-10-01

The temperature is the essential factor that influences the efficiency of anaerobic reactors. During the operation of the anaerobic reactor, the fluctuations of ambient temperature can cause a change in the internal temperature of the reactor. Therefore, insulation and heating measures are often used to maintain anaerobic reactor's internal temperature. In this paper, a simplified heat transfer model was developed to study heat transfer between cylindrical anaerobic reactors and their surroundings. Three cylindrical reactors of different sizes were studied, and the internal relations between ambient temperature, thickness of insulation, and temperature fluctuations of the reactors were obtained at different reactor sizes. The model was calibrated by a sensitivity analysis, and the calibrated model was well able to predict reactor temperature. The Nash-Sutcliffe model efficiency coefficient was used to assess the predictive power of heat transfer models. The Nash coefficients of the three reactors were 0.76, 0.60, and 0.45, respectively. The model can provide reference for the thermal insulation design of cylindrical anaerobic reactors.

Seasonal warming of the Middle Atlantic Bight Cold Pool

NASA Astrophysics Data System (ADS)

Lentz, S. J.

2017-02-01

The Cold Pool is a 20-60 m thick band of cold, near-bottom water that persists from spring to fall over the midshelf and outer shelf of the Middle Atlantic Bight (MAB) and Southern Flank of Georges Bank. The Cold Pool is remnant winter water bounded above by the seasonal thermocline and offshore by warmer slope water. Historical temperature profiles are used to characterize the average annual evolution and spatial structure of the Cold Pool. The Cold Pool gradually warms from spring to summer at a rate of order 1°C month-1. The warming rate is faster in shallower water where the Cold Pool is thinner, consistent with a vertical turbulent heat flux from the thermocline to the Cold Pool. The Cold Pool warming rate also varies along the shelf; it is larger over Georges Bank and smaller in the southern MAB. The mean turbulent diffusivities at the top of the Cold Pool, estimated from the spring to summer mean heat balance, are an order of magnitude larger over Georges Bank than in the southern MAB, consistent with much stronger tidal mixing over Georges Bank than in the southern MAB. The stronger tidal mixing causes the Cold Pool to warm more rapidly over Georges Bank and the eastern New England shelf than in the New York Bight or southern MAB. Consequently, the coldest Cold Pool water is located in the New York Bight from late spring to summer.

Spatial analysis of paediatric swimming pool submersions by housing type.

PubMed

Shenoi, Rohit P; Levine, Ned; Jones, Jennifer L; Frost, Mary H; Koerner, Christine E; Fraser, John J

2015-08-01

Drowning is a major cause of unintentional childhood death. The relationship between childhood swimming pool submersions, neighbourhood sociodemographics, housing type and swimming pool location was examined in Harris County, Texas. Childhood pool submersion incidents were examined for spatial clustering using the Nearest Neighbor Hierarchical Cluster (Nnh) algorithm. To relate submersions to predictive factors, an Markov Chain Monte Carlo (MCMC) Poisson-Lognormal-Conditional Autoregressive (CAR) spatial regression model was tested at the census tract level. There were 260 submersions; 49 were fatal. Forty-two per cent occurred at single-family residences and 36% at multifamily residential buildings. The risk of a submersion was 2.7 times higher for a child at a multifamily than a single-family residence and 28 times more likely in a multifamily swimming pool than a single family pool. However, multifamily submersions were clustered because of the concentration of such buildings with pools. Spatial clustering did not occur in single-family residences. At the tract level, submersions in single-family and multifamily residences were best predicted by the number of pools by housing type and the number of children aged 0-17 by housing type. Paediatric swimming pool submersions in multifamily buildings are spatially clustered. The likelihood of submersions is higher for children who live in multifamily buildings with pools than those who live in single-family homes with pools. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://group.bmj.com/group/rights-licensing/permissions.

Saltmarsh pool and tidal creek morphodynamics: Dynamic equilibrium of northern latitude saltmarshes?

NASA Astrophysics Data System (ADS)

Wilson, Carol A.; Hughes, Zoe J.; FitzGerald, Duncan M.; Hopkinson, Charles S.; Valentine, Vinton; Kolker, Alexander S.

2014-05-01

Many saltmarsh platforms in New England and other northern climates (e.g. Canada, northern Europe) exhibit poor drainage, creating waterlogged regions where short-form Spartina alterniflora dominates and stagnant pools that experience tidal exchange only during spring tides and storm-induced flooding events. The processes related to pool formation and tidal creek incision (via headward erosion) that may eventually drain these features are poorly understood, however it has been suggested that an increase in pool occurrence in recent decades is due to waterlogging stress from sea-level rise. We present evidence here that saltmarshes in Plum Island Estuary of Massachusetts are keeping pace with sea-level rise, and that the recent increase in saltmarsh pool area coincides with changes in drainage density from a legacy of anthropogenic ditching (reversion to natural drainage conditions). Gradients, in addition to elevation and hydroperiod, are critical for saltmarsh pool formation. Additionally, elevation and vegetative changes associated with pool formation, creek incision, subsequent drainage of pools, and recolonization by S. alterniflora are quantified. Pool and creek dynamics were found to be cyclic in nature, and represent platform elevation in dynamic equilibrium with sea level whereby saltmarsh elevation may be lowered (due to degradation of organic matter and formation of a pool), however may be regained on short timescales (101-2 yr) with creek incision into pools and restoration of tidal exchange. Rapid vertical accretion is associated with sedimentation and S. alterniflora plant recolonization.

Laboratory investigation and simulation of breakthrough curves in karst conduits with pools

NASA Astrophysics Data System (ADS)

Zhao, Xiaoer; Chang, Yong; Wu, Jichun; Peng, Fu

2017-12-01

A series of laboratory experiments are performed under various hydrological conditions to analyze the effect of pools in pipes on breakthrough curves (BTCs). The BTCs are generated after instantaneous injections of NaCl tracer solution. In order to test the feasibility of reproducing the BTCs and obtain transport parameters, three modeling approaches have been applied: the equilibrium model, the linear graphical method and the two-region nonequilibrium model. The investigation results show that pools induce tailing of the BTCs, and the shapes of BTCs depend on pool geometries and hydrological conditions. The simulations reveal that the two-region nonequilibrium model yields the best fits to experimental BTCs because the model can describe the transient storage in pools by the partition coefficient and the mass transfer coefficient. The model parameters indicate that pools produce high dispersion. The increased tailing occurs mainly because the partition coefficient decreases, as the number of pools increases. When comparing the tracer BTCs obtained using the two types of pools with the same size, the more appreciable BTC tails that occur for symmetrical pools likely result mainly from the less intense exchange between the water in the pools and the water in the pipe, because the partition coefficients for the two types of pools are virtually identical. Dispersivity values decrease as flow rates increase; however, the trend in dispersion is not clear. The reduced tailing is attributed to a decrease in immobile water with increasing flow rate. It provides evidence for hydrodynamically controlled tailing effects.

Prevention of needless deaths from drowning.

PubMed

Modell, Jerome H

2010-07-01

To determine whether faulty pool maintenance and substandard lifeguard performance critically delayed retrieval and resuscitation of a significant number of pool drowning victims. One hundred and eighty drowning incidents that resulted in litigation from 1998 to 2008 were studied to determine whether faulty pool maintenance and/or substandard lifeguard performance delayed retrieval and thereby contributed to the death of these persons. A total of 180 swimming pools-commercial and private-were included. Ninety-seven of these pools were manned by lifeguards. Subjects who underwent the drowning process and suffered severe brain injury or death were reviewed to determine the rescue and resuscitation attempts by lifeguards or bystanders at the pool. One hundred and seventy-seven of the 180 persons who underwent the drowning process died. Cases were analyzed as to whether faulty pool maintenance and/or substandard lifeguard performance contributed to their demise. At fault were cloudy or dirty water; drain pipes that created underwater suction to trap victims; inadequate fencing around pools through which small children gained access; permitting small children to be at the pool without adult supervision; permitting dangerous exercises such as hyperventilation while underwater swimming, resulting in shallow water blackout; lifeguards not being attentive, being distracted by other persons, performing nonrelated chores, leaving their positions without proper relief, while failure to enter the water when told persons were submerged. Faculty pool maintenance and substandard lifeguard performance critically delayed retrieval and resuscitation of a significant number of pool drowning victims.

Etude spectroscopique des collisions moleculaires (hydrogene-azote et hydrogene-oxygene) a des energies de quelques MeV

NASA Astrophysics Data System (ADS)

Plante, Jacinthe

1998-09-01

Les resultats presentes ici proviennent d'une etude systematique portant sur les collisions a vitesse constante, entre les projectiles d'hydrogene (H+, H2+ et H3+ a 1 MeV/nucleon) et deux cibles gazeuses (N2 et O2), soumises a differentes pressions. Les collisions sont analysees a l'aide des spectres d'emission (de 400 A a 6650 A) et des graphiques intensite/pression. Les spectres ont revele la presence des raies d'azote atomique, d'azote moleculaire, d'oxygene atomique, d'oxygene moleculaire et d'hydrogene atomique. Les raies d'hydrogene sont observees seulement avec les projectiles H2+ et H3+. Donc les processus responsables de la formation de ces raies sont des mecanismes de fragmentation des projectiles. Pour conclure, il existe une difference notable entre les projectiles et les differentes pressions. Les raies d'azote et d'oxygene augmentent selon la pression et les raies d'hydrogene atomique presentent une relation non lineaire avec la pression.

Diffraction des neutrons : principe, dispositifs expérimentaux et applications

NASA Astrophysics Data System (ADS)

Muller, C.

2003-02-01

La diffraction de neutrons, sur monocristal ou sur échantillon polycristallin (ou poudre), est une technique très largement utilisée, en science des matériaux comme en biologie, lorsque l'on souhaite déterminer la structure cristalline d'un composé ou d'une molécule. Toutefois, le degré de précision de la détermination structurale est très corrélé au choix de l'instrument utilisé. Il s'en suit que la question “comment choisir l'instrument le mieux adapté au composé et à la problématique ?" apparaît comme fondamentale. L'objectif de ce cours est de tenter de répondre à cette question en décrivant brièvement les caractéristiques instrumentales de différents diffractomètres, en exposant les avantages spécifiques des expériences de diffraction de neutrons et en donnant quelques exemples d'application.

Andrei Sakharov

ScienceCinema

None

2017-12-09

Ce discours donné par Mons.Jonauch qui est né en Tchécoslovaquie et a fait ses études à Leningrad, Moscou et Prague, est organisé par le comité Youri Orlov. Le conférencier parle de Andrei Sakharov, ce physicien et homme soviétique qui fit ses études à Moscou, effectua des recherches sur les armes thermonucléaires et entra à l'Académie des Sciences d'URSS en 1953. Il participa à la mise au point de la bombe à hydrogène, mais s'opposa quelques années plus tard à la poursuite des expériences nucléaires. Il créa en 1970 le comité pour la défense des droits de l'homme ce que lui valut le prix Nobel de la paix en 1975.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, Robert M.

1983-01-01

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, R.M.

1983-11-08

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

47 CFR 13.215 - Question pools.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 47 Telecommunication 1 2010-10-01 2010-10-01 false Question pools. 13.215 Section 13.215 Telecommunication FEDERAL COMMUNICATIONS COMMISSION GENERAL COMMERCIAL RADIO OPERATORS Examination System § 13.215 Question pools. The question pool for each written examination element will be composed of questions...

Development of the Chacon Dakota associated pool

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

Walsh, E.N.

1979-01-01

Discovery and development of the Chacon Dakota associated pool is a very important Dakota formation development outside of the basic Dakota gas pool in the San Juan Basin area. Other Dakota formation developments are not of the magnitude as the Chacon Dakota associated pool.

48 CFR 28.304 - Risk-pooling arrangements.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Risk-pooling arrangements... CONTRACTING REQUIREMENTS BONDS AND INSURANCE Insurance 28.304 Risk-pooling arrangements. Agencies may establish risk-pooling arrangements. These arrangements are designed to use the services of the insurance...

17 CFR 4.22 - Reporting to pool participants.

Code of Federal Regulations, 2010 CFR

2010-04-01

..., the financial statements are not required to include consolidated information for all series. (7) For... event that the International Financial Reporting Standards require consolidated financial statements for... reporting pool's consolidated financial statements. (ii) The commodity pool operator of a pool that meets...

Reserve growth in oil pools of Alberta: Model and forecast

USGS Publications Warehouse

Verma, M.; Cook, T.

2010-01-01

Reserve growth is recognized as a major component of additions to reserves in most oil provinces around the world, particularly in mature provinces. It takes place as a result of the discovery of new pools/reservoirs and extensions of known pools within existing fields, improved knowledge of reservoirs over time leading to a change in estimates of original oil-in-place, and improvement in recovery factor through the application of new technology, such as enhanced oil recovery methods, horizontal/multilateral drilling, and 4D seismic. A reserve growth study was conducted on oil pools in Alberta, Canada, with the following objectives: 1) evaluate historical oil reserve data in order to assess the potential for future reserve growth; 2) develop reserve growth models/ functions to help forecast hydrocarbon volumes; 3) study reserve growth sensitivity to various parameters (for example, pool size, porosity, and oil gravity); and 4) compare reserve growth in oil pools and fields in Alberta with those from other large petroleum provinces around the world. The reported known recoverable oil exclusive of Athabasca oil sands in Alberta increased from 4.5 billion barrels of oil (BBO) in 1960 to 17 BBO in 2005. Some of the pools that were included in the existing database were excluded from the present study for lack of adequate data. Therefore, the known recoverable oil increased from 4.2 to 13.9 BBO over the period from 1960 through 2005, with new discoveries contributing 3.7 BBO and reserve growth adding 6 BBO. This reserve growth took place mostly in pools with more than 125,000 barrels of known recoverable oil. Pools with light oil accounted for most of the total known oil volume, therefore reflecting the overall pool growth. Smaller pools, in contrast, shrank in their total recoverable volumes over the years. Pools with heavy oil (gravity less than 20o API) make up only a small share (3.8 percent) of the total recoverable oil; they showed a 23-fold growth compared to about 3.5-fold growth in pools with medium oil and 2.2-fold growth in pools with light oil over a fifty-year period. The analysis indicates that pools with high porosity reservoirs (greater than 30 percent porosity) grew more than pools with lower porosity reservoirs which could possibly be attributed to permeability differences between the two types. Reserve growth models for Alberta, Canada, show the growth at field level is almost twice as much as at pool level, possibly because the analysis has evaluated fields with two or more pools with different discovery years. Based on the models, the growth in oil volumes in Alberta pools over the next five-year period (2006-2010) is expected to be about 454 million barrels of oil. Over a twenty-five year period, the cumulative reserve growth in Alberta oil pools has been only 2-fold compared to a 4- to- 5-fold increase in other petroleum producing areas such as Saskatchewan, Volga-Ural, U.S. onshore fields, and U.S. Gulf of Mexico. However, the growth at the field level compares well with that of U.S. onshore fields. In other petroleum provinces, the reserves are reported at field levels rather than at pool levels, the latter basically being the equivalent of individual reservoirs. ?? 2010 by the Canadian Society of Petroleum Geologists.

Nuclear reactor neutron shielding

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

Speaker, Daniel P; Neeley, Gary W; Inman, James B

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactormore » cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.« less

Reactor pressure vessel head vents and methods of using the same

DOEpatents

Gels, John L; Keck, David J; Deaver, Gerald A

2014-10-28

Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

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

Refuge habitats for fishes during seasonal drying in an intermittent stream: Movement, survival and abundance of three minnow species

USGS Publications Warehouse

Hodges, S.W.; Magoulick, D.D.

2011-01-01

Drought and summer drying can be important disturbance events in many small streams leading to intermittent or isolated habitats. We examined what habitats act as refuges for fishes during summer drying, hypothesizing that pools would act as refuge habitats. We predicted that during drying fish would show directional movement into pools from riffle habitats, survival rates would be greater in pools than in riffles, and fish abundance would increase in pool habitats. We examined movement, survival and abundance of three minnow species, bigeye shiner (Notropis boops), highland stoneroller (Campostoma spadiceum) and creek chub (Semotilus atromaculatus), during seasonal stream drying in an Ozark stream using a closed robust multi-strata mark-recapture sampling. Population parameters were estimated using plausible models within program MARK, where a priori models are ranked using Akaike's Information Criterion. Creek chub showed directional movement into pools and increased survival and abundance in pools during drying. Highland stonerollers showed strong directional movement into pools and abundance increased in pools during drying, but survival rates were not significantly greater in pools than riffles. Bigeye shiners showed high movement rates during drying, but the movement was non-directional, and survival rates were greater in riffles than pools. Therefore, creek chub supported our hypothesis and pools appear to act as refuge habitats for this species, whereas highland stonerollers partly supported the hypothesis and bigeye shiners did not support the pool refuge hypothesis. Refuge habitats during drying are species dependent. An urgent need exists to further understand refuge habitats in streams given projected changes in climate and continued alteration of hydrological regimes. ?? 2011 Springer Basel AG (outside the USA).

Environmental versus genetic influences on growth rates of the corals Pocillopora eydouxi and Porites lobata (Anthozoa: Scleractinia)

USGS Publications Warehouse

Smith, L.W.; Wirshing, H.H.; Baker, A.C.; Birkeland, C.

2008-01-01

Reciprocal transplant experiments of the corals Pocillopora eydouxi Milne Edwards & Haime and Porites lobata Dana were carried out for an 18-month period from September 2004 to March 2006 between two back reef pools on Ofu Island, American Samoa, to test environmental versus genetic effects on skeletal growth rates. Skeletal growth of P. eydouxi showed environmental but not genetic effects, resulting in doubling of growth in Pool 300 compared with Pool 400. There were no environmental or genetic effects on skeletal growth of P. lobata. Pool 300 had more frequent and longer durations of elevated seawater temperatures than Pool 400, characteristics likely to decrease rather than increase skeletal growth. Pool 300 also had higher nutrient levels and flow velocities than Pool 400, characteristics that may increase skeletal growth. However, higher nutrient levels would be expected to increase skeletal growth in both species, but there was no difference between the pools in P. lobata growth. P. eydouxi is much more common in high-energy environments than P. lobata; thus the higher flow velocities in Pool 300 than in Pool 400 may have positively affected skeletal growth of P. eydouxi while not having a detectable effect on P. lobata. The greater skeletal growth of P. eydouxi in Pool 300 occurred despite the presence of clade D zooxanthellae in several source colonies in Pool 300, a genotype known to result in greater heat resistance but slower skeletal growth. Increased skeletal growth rates in higher water motion may provide P. eydouxi a competitive advantage in shallow, high-energy enviromnents where competition for space is intense. ?? 2008 by University of Hawai'i Press. All rights reserved.

10 CFR 52.167 - Issuance of manufacturing license.

Code of Federal Regulations, 2010 CFR

2010-01-01

... proposed reactor(s) can be incorporated into a nuclear power plant and operated at sites having... design and manufacture the proposed nuclear power reactor(s); (5) The proposed inspections, tests... the construction of a nuclear power facility using the manufactured reactor(s). (2) A holder of a...

Cool pool development. Quarterly technical report No. 2, June-December 1979

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

Crowther, K.

1980-01-05

The Cool Pool is a variation of the evaporating roof pond idea. The pool is isolated from the living space and the cooled pond water thermosiphons into the water columns located within the building. A computer model of the Cool Pool and the various heat and mass transfer mechanisms involved in the system are discussed. Theory will be compared to experimental data collected from a Cool Pool test building.

47 CFR 97.523 - Question pools.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 47 Telecommunication 5 2010-10-01 2010-10-01 false Question pools. 97.523 Section 97.523 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES AMATEUR RADIO... question pool for each written examination element. Each question pool must contain at least 10 times the...

16 CFR 1633.5 - Prototype pooling and confirmation testing requirements.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Prototype pooling and confirmation testing... Prototype pooling and confirmation testing requirements. (a) Prototype pooling. One or more manufacturers may rely on a qualified prototype produced by another manufacturer or prototype developer provided...

Rift Valley Fever Virus Epidemic in Kenya, 2006/2007: The Entomologic Investigations

PubMed Central

Sang, Rosemary; Kioko, Elizabeth; Lutomiah, Joel; Warigia, Marion; Ochieng, Caroline; O'Guinn, Monica; Lee, John S.; Koka, Hellen; Godsey, Marvin; Hoel, David; Hanafi, Hanafi; Miller, Barry; Schnabel, David; Breiman, Robert F.; Richardson, Jason

2010-01-01

In December 2006, Rift Valley fever (RVF) was diagnosed in humans in Garissa Hospital, Kenya and an outbreak reported affecting 11 districts. Entomologic surveillance was performed in four districts to determine the epidemic/epizootic vectors of RVF virus (RVFV). Approximately 297,000 mosquitoes were collected, 164,626 identified to species, 72,058 sorted into 3,003 pools and tested for RVFV by reverse transcription-polymerase chain reaction. Seventy-seven pools representing 10 species tested positive for RVFV, including Aedes mcintoshi/circumluteolus (26 pools), Aedes ochraceus (23 pools), Mansonia uniformis (15 pools); Culex poicilipes, Culex bitaeniorhynchus (3 pools each); Anopheles squamosus, Mansonia africana (2 pools each); Culex quinquefasciatus, Culex univittatus, Aedes pembaensis (1 pool each). Positive Ae. pembaensis, Cx. univittatus, and Cx. bitaeniorhynchus was a first time observation. Species composition, densities, and infection varied among districts supporting hypothesis that different mosquito species serve as epizootic/epidemic vectors of RVFV in diverse ecologies, creating a complex epidemiologic pattern in East Africa. PMID:20682903

Diving into Cold Pools

NASA Astrophysics Data System (ADS)

van den Heever, S. C.; Grant, L. D.; Drager, A. J.

2017-12-01

Cold pools play a significant role in convective storm initiation, organization and longevity. Given their role in convective life cycles, recent efforts have been focused on improving the representation of cold pool processes within weather forecast models, as well as on developing cold pool parameterizations in order to better represent their impacts within global climate models. Understanding the physical processes governing cold pool formation, intensity and dissipation is therefore critical to these efforts. Cold pool characteristics are influenced by numerous factors, including those associated with precipitation formation and evaporation, variations in the environmental moisture and shear, and land surface interactions. The focus of this talk will be on the manner in which the surface characteristics and associated processes impact cold pool genesis and dissipation. In particular, the results from high-resolution modeling studies focusing on the role of sensible and latent heat fluxes, soil moisture and SST will be presented. The results from a recent field campaign examining cold pools over northern Colorado will also be discussed.

In vivo insertion pool sequencing identifies virulence factors in a complex fungal–host interaction

PubMed Central

Uhse, Simon; Pflug, Florian G.; Stirnberg, Alexandra; Ehrlinger, Klaus; von Haeseler, Arndt

2018-01-01

Large-scale insertional mutagenesis screens can be powerful genome-wide tools if they are streamlined with efficient downstream analysis, which is a serious bottleneck in complex biological systems. A major impediment to the success of next-generation sequencing (NGS)-based screens for virulence factors is that the genetic material of pathogens is often underrepresented within the eukaryotic host, making detection extremely challenging. We therefore established insertion Pool-Sequencing (iPool-Seq) on maize infected with the biotrophic fungus U. maydis. iPool-Seq features tagmentation, unique molecular barcodes, and affinity purification of pathogen insertion mutant DNA from in vivo-infected tissues. In a proof of concept using iPool-Seq, we identified 28 virulence factors, including 23 that were previously uncharacterized, from an initial pool of 195 candidate effector mutants. Because of its sensitivity and quantitative nature, iPool-Seq can be applied to any insertional mutagenesis library and is especially suitable for genetically complex setups like pooled infections of eukaryotic hosts. PMID:29684023

Pooling Cervical Swabs and Testing by Ligase Chain Reaction Are Accurate and Cost-Saving Strategies for Diagnosis of Chlamydia trachomatis

PubMed Central

Kapala, J.; Copes, D.; Sproston, A.; Patel, J.; Jang, D.; Petrich, A.; Mahony, J.; Biers, K.; Chernesky, M.

2000-01-01

Specimen pooling to achieve efficiency when testing urine specimens for Chlamydia trachomatis nucleic acids has been suggested. We pooled endocervical swabs from 1,288 women and also tested individual swabs by ligase chain reaction (LCR). Out of 53 positive specimens, pools of 4 or 8 specimens missed two positives, providing 96.2% accuracy compared to individual test results. Dilution and positive-control spiking experiments showed that negative specimens with inhibitors of LCR in the pool reduced the signal. Conversely, two extra positives, detected only through pooling, were negative by individual testing but became positive after storage, suggesting that fresh positive specimens with labile inhibitors may be positive in a pool because of dilution of inhibitors. For this population of women with a 4% prevalence of C. trachomatis infection, substantial savings in cost of reagents (55 to 63%) and technologist time (50 to 63%) made pooling strategies a desirable alternative to individual testing. PMID:10878029

Fire control method and analytical model for large liquid hydrocarbon pool fires

NASA Technical Reports Server (NTRS)

Fenton, D. L.

1986-01-01

The dominate parameter governing the behavior of a liquid hydrocarbon (JP-5) pool fire is wind speed. The most effective method of controlling wind speed in the vicinity of a large circular (10 m dia.) pool fire is a set of concentric screens located outside the perimeter. Because detailed behavior of the pool fire structure within one pool fire diameter is unknown, an analytical model supported by careful experiments is under development. As a first step toward this development, a regional pool fire model was constructed for the no-wind condition consisting of three zones -- liquid fuel, combustion, and plume -- where the predicted variables are mass burning rate and characteristic temperatures of the combustion and plume zones. This zone pool fire model can be modified to incorporate plume bending by wind, radiation absorption by soot particles, and a different ambient air flow entrainment rate. Results from the zone model are given for a pool diameter of 1.3 m and are found to reproduce values in the literature.

Rift Valley fever virus epidemic in Kenya, 2006/2007: the entomologic investigations.

PubMed

Sang, Rosemary; Kioko, Elizabeth; Lutomiah, Joel; Warigia, Marion; Ochieng, Caroline; O'Guinn, Monica; Lee, John S; Koka, Hellen; Godsey, Marvin; Hoel, David; Hanafi, Hanafi; Miller, Barry; Schnabel, David; Breiman, Robert F; Richardson, Jason

2010-08-01

In December 2006, Rift Valley fever (RVF) was diagnosed in humans in Garissa Hospital, Kenya and an outbreak reported affecting 11 districts. Entomologic surveillance was performed in four districts to determine the epidemic/epizootic vectors of RVF virus (RVFV). Approximately 297,000 mosquitoes were collected, 164,626 identified to species, 72,058 sorted into 3,003 pools and tested for RVFV by reverse transcription-polymerase chain reaction. Seventy-seven pools representing 10 species tested positive for RVFV, including Aedes mcintoshi/circumluteolus (26 pools), Aedes ochraceus (23 pools), Mansonia uniformis (15 pools); Culex poicilipes, Culex bitaeniorhynchus (3 pools each); Anopheles squamosus, Mansonia africana (2 pools each); Culex quinquefasciatus, Culex univittatus, Aedes pembaensis (1 pool each). Positive Ae. pembaensis, Cx. univittatus, and Cx. bitaeniorhynchus was a first time observation. Species composition, densities, and infection varied among districts supporting hypothesis that different mosquito species serve as epizootic/epidemic vectors of RVFV in diverse ecologies, creating a complex epidemiologic pattern in East Africa.

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

A novel plant protection system designed for use in the TREAT Upgrade (TU) reactor is described. The TU reactor is designed for controlled transient operation in the testing of reactor fuel behavior under simulated reactor accident conditions. Safe operation of the reactor is of paramount importance and the Plant Protection System (PPS) had to be designed to exacting requirements. Researchers believe that the strategy developed for the TU has potential application to the multimegawatt space reactors and represents the state of the art in terrestrial transient reactor protection systems.

Process and apparatus for adding and removing particles from pressurized reactors

DOEpatents

Milligan, John D.

1983-01-01

A method for adding and removing fine particles from a pressurized reactor is provided, which comprises connecting the reactor to a container, sealing the container from the reactor, filling the container with particles and a liquid material compatible with the reactants, pressurizing the container to substantially the reactor pressure, removing the seal between the reactor and the container, permitting particles to fall into or out of the reactor, and resealing the container from the reactor. An apparatus for adding and removing particles is also disclosed.

Effects of imperfect mixing on low-density polyethylene reactor dynamics

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

Villa, C.M.; Dihora, J.O.; Ray, W.H.

1998-07-01

Earlier work considered the effect of feed conditions and controller configuration on the runaway behavior of LDPE autoclave reactors assuming a perfectly mixed reactor. This study provides additional insight on the dynamics of such reactors by using an imperfectly mixed reactor model and bifurcation analysis to show the changes in the stability region when there is imperfect macroscale mixing. The presence of imperfect mixing substantially increases the range of stable operation of the reactor and makes the process much easier to control than for a perfectly mixed reactor. The results of model analysis and simulations are used to identify somemore » of the conditions that lead to unstable reactor behavior and to suggest ways to avoid reactor runaway or reactor extinction during grade transitions and other process operation disturbances.« less

Co-evolving Physical and Biological Organization in Step-pool Channels: Experiments from a Restoration Reach on Wildcat Creek, California

NASA Astrophysics Data System (ADS)

Chin, A.; O'Dowd, A. P.; Mendez, P. K.; Velasco, K. Z.; Leventhal, R. D.; Storesund, R.; Laurencio, L. R.

2014-12-01

Step-pools are important features in fluvial systems. Through energy dissipation, step-pools provide stability in high-energy environments that otherwise may erode and degrade. Although research has focused on geomorphological aspects of step-pool channels, the ecological significance of step-pool streams is increasingly recognized. Step-pool streams often contain higher density and diversity of benthic macroinvertebrates and are critical habitats for organisms such as salmonids and tailed frogs. Step-pools are therefore increasingly used to restore eroding channels and improve ecological conditions. This paper addresses a restoration reach of Wildcat Creek in Berkeley, California that featured an installation of step-pools in 2012. The design framework recognized step-pool formation as a self-organizing process that produces a rhythmic morphology. After placing step particles at locations where step-pools are expected to form according to hydraulic theory, the self-organizing approach allowed fluvial processes to refine the rocks into adjusted sequences over time. In addition, a 30-meter "experimental" reach was created to explore the co-evolution of geomorphological and ecological characteristics. After constructing a plane bed channel, boulders and cobbles piled at the upstream end allowed natural flows to mobilize and sort them into step-pool sequences. Ground surveys and LiDAR recorded the development of step-pool sequences over several seasons. Concurrent sampling of benthic macroinvertebrates documented the formation of biological communities in conjunction with habitat. Biological sampling in an upstream reference reach provided a comparison with the restored reach over time. Results to date show an emergent step-pool channel with steps that segment the plane bed into initial step and pool habitats. Biological communities are beginning to form, showing more distinction among habitat types during some seasons, although they do not yet approach reference values at this stage of development. Research over longer timeframes is needed to reveal how biological and physical characteristics may co-organize toward an equilibrium landscape. Such integrated understanding will assist development of innovative restoration designs.

Analysis of Phenix end-of-life natural convection test with the MARS-LMR code

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

Jeong, H. Y.; Ha, K. S.; Lee, K. L.

The end-of-life test of Phenix reactor performed by the CEA provided an opportunity to have reliable and valuable test data for the validation and verification of a SFR system analysis code. KAERI joined this international program for the analysis of Phenix end-of-life natural circulation test coordinated by the IAEA from 2008. The main objectives of this study were to evaluate the capability of existing SFR system analysis code MARS-LMR and to identify any limitation of the code. The analysis was performed in three stages: pre-test analysis, blind posttest analysis, and final post-test analysis. In the pre-test analysis, the design conditionsmore » provided by the CEA were used to obtain a prediction of the test. The blind post-test analysis was based on the test conditions measured during the tests but the test results were not provided from the CEA. The final post-test analysis was performed to predict the test results as accurate as possible by improving the previous modeling of the test. Based on the pre-test analysis and blind test analysis, the modeling for heat structures in the hot pool and cold pool, steel structures in the core, heat loss from roof and vessel, and the flow path at core outlet were reinforced in the final analysis. The results of the final post-test analysis could be characterized into three different phases. In the early phase, the MARS-LMR simulated the heat-up process correctly due to the enhanced heat structure modeling. In the mid phase before the opening of SG casing, the code reproduced the decrease of core outlet temperature successfully. Finally, in the later phase the increase of heat removal by the opening of the SG opening was well predicted with the MARS-LMR code. (authors)« less

State High-Risk Pools: An Update on the Minnesota Comprehensive Health Association

PubMed Central

Spencer, Donna; Burke, Courtney E.

2011-01-01

State health insurance high-risk pools are a key component of the US health care system's safety net, because they provide health insurance to the “uninsurable.” In 2007, 34 states had individual high-risk pools, which covered more than 200 000 people at a total cost of $1.8 billion. We examine the experience of the largest and oldest pool in the nation, the Minnesota Comprehensive Health Association, to document key issues facing state high-risk pools in enrollment and financing. We also considered the role and future of high-risk pools in light of national health care finance reform. PMID:21228286

A Hox regulatory network establishes motor neuron pool identity and target-muscle connectivity.

PubMed

Dasen, Jeremy S; Tice, Bonnie C; Brenner-Morton, Susan; Jessell, Thomas M

2005-11-04

Spinal motor neurons acquire specialized "pool" identities that determine their ability to form selective connections with target muscles in the limb, but the molecular basis of this striking example of neuronal specificity has remained unclear. We show here that a Hox transcriptional regulatory network specifies motor neuron pool identity and connectivity. Two interdependent sets of Hox regulatory interactions operate within motor neurons, one assigning rostrocaudal motor pool position and a second directing motor pool diversity at a single segmental level. This Hox regulatory network directs the downstream transcriptional identity of motor neuron pools and defines the pattern of target-muscle connectivity.

Inter-progenitor pool wiring: An evolutionarily conserved strategy that expands neural circuit diversity.

PubMed

Suzuki, Takumi; Sato, Makoto

2017-11-15

Diversification of neuronal types is key to establishing functional variations in neural circuits. The first critical step to generate neuronal diversity is to organize the compartmental domains of developing brains into spatially distinct neural progenitor pools. Neural progenitors in each pool then generate a unique set of diverse neurons through specific spatiotemporal specification processes. In this review article, we focus on an additional mechanism, 'inter-progenitor pool wiring', that further expands the diversity of neural circuits. After diverse types of neurons are generated in one progenitor pool, a fraction of these neurons start migrating toward a remote brain region containing neurons that originate from another progenitor pool. Finally, neurons of different origins are intermingled and eventually form complex but precise neural circuits. The developing cerebral cortex of mammalian brains is one of the best examples of inter-progenitor pool wiring. However, Drosophila visual system development has revealed similar mechanisms in invertebrate brains, suggesting that inter-progenitor pool wiring is an evolutionarily conserved strategy that expands neural circuit diversity. Here, we will discuss how inter-progenitor pool wiring is accomplished in mammalian and fly brain systems. Copyright © 2017 Elsevier Inc. All rights reserved.