COOLING TOWER PUMP HOUSE, TRA606. THREE OF SIX SECTIONS OF ...

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

COOLING TOWER PUMP HOUSE, TRA-606. THREE OF SIX SECTIONS OF COOLING TOWER ARE VISIBLE ABOVE RAILING. PUMP HOUSE IN FOREGROUND IS ON SOUTH SIDE OF COOLING TOWER. NOTE THREE PIPES TAKING WATER FROM PUMP HOUSE TO HOT DECK OF COOLING TOWER. EMERGENCY WATER SUPPLY TOWER IS ALSO IN VIEW. INL NEGATIVE NO. 6197. Unknown Photographer, 6/27/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Locating hot and cold-legs in a nuclear powered steam generation system

DOEpatents

Ekeroth, D.E.; Corletti, M.M.

1993-11-16

A nuclear reactor steam generator includes a reactor vessel for heating water and a steam generator with a pump casing at the lowest point on the steam generator. A cold-leg pipe extends horizontally between the steam generator and the reactor vessel to return water from the steam generator to the reactor vessel. The bottom of the cold-leg pipe is at a first height above the bottom of the reactor vessel. A hot-leg pipe with one end connected to the steam generator and a second end connected to the reactor vessel has a first pipe region extending downwardly from the steam generator to a location between the steam generator and the reactor vessel at which a bottom of the hot-leg pipe is at a second height above the bottom of the reactor vessel. A second region extends from that location in a horizontal direction at the second height to the point at which the hot-leg pipe connects to the reactor vessel. A pump is attached to the casing at a location below the first and second heights and returns water from the steam generator to the reactor vessel over the cold-leg. The first height is greater than the second height and the bottom of the steam generator is at a height above the bottom of the reactor vessel that is greater than the first and second heights. A residual heat recovery pump is below the hot-leg and has an inlet line from the hot-leg that slopes down continuously to the pump inlet. 2 figures.

Locating hot and cold-legs in a nuclear powered steam generation system

DOEpatents

Ekeroth, Douglas E.; Corletti, Michael M.

1993-01-01

A nuclear reactor steam generator includes a reactor vessel for heating water and a steam generator with a pump casing at the lowest point on the steam generator. A cold-leg pipe extends horizontally between the steam generator and the reactor vessel to return water from the steam generator to the reactor vessel. The bottom of the cold-leg pipe is at a first height above the bottom of the reactor vessel. A hot-leg pipe with one end connected to the steam generator and a second end connected to the reactor vessel has a first pipe region extending downwardly from the steam generator to a location between the steam generator and the reactor vessel at which a bottom of the hot-leg pipe is at a second height above the bottom of the reactor vessel. A second region extends from that location in a horizontal direction at the second height to the point at which the hot-leg pipe connects to the reactor vessel. A pump is attached to the casing at a location below the first and second heights and returns water from the steam generator to the reactor vessel over the cold-leg. The first height is greater than the second height and the bottom of the steam generator is at a height above the bottom of the reactor vessel that is greater than the first and second heights. A residual heat recovery pump is below the hot-leg and has an inlet line from the hot-leg that slopes down continuously to the pump inlet.

APPARATUS FOR CONTROL OF A BOILING REACTOR RESPONSIVE TO STEAM DEMAND

DOEpatents

Treshow, M.

1963-07-23

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

Analysis of boron dilution in a four-loop PWR

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

Sun, J.G.; Sha, W.T.

1995-12-31

Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plantmore » is at hot standby and the reactor coolant system has been heated up to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated.« less

ETR HEAT EXCHANGER BUILDING, TRA644. A PRIMARY COOLANT PUMP AND ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. A PRIMARY COOLANT PUMP AND 24-INCH CHECK VALVE ARE MOUNTED IN A SHIELDED CUBICLE. NOTE CONNECTION AT RIGHT THROUGH SHIELD WALL TO PUMP MOTOR ON OTHER SIDE. INL NEGATIVE NO. 56-4177. Jack L. Anderson, Photographer, 12/21/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

FALCON nuclear-reactor-pumped laser program and wireless power transmission

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

Lipinski, R.J.; Pickard, P.S.

1992-12-31

FALCON is a high-power, reactor-pumped laser concept. The major strengths of a reactor-pumped laser are (1) simple, modular construction, (2) long-duration, closed-cycle capability, (3) self-contained power, (4) compact size, and (5) a variety of wavelengths (from visible to infrared). Reactor-pumped lasing has been demonstrated experimentally in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Powers up to 300 W for 2 ms have been demonstrated. Projected beam quality for FALCON is good enough that frequency doubling at reasonablemore » efficiencies could be expected to yield wavelengths at 353, 363, 636, 867, 896, 1016, 1315, 1325, and 1685 nm. Appropriate missions for FALCON are described and include power beaming to satellites, the moon, and unmanned surveillance planes; lunar mapping; space debris removal; and laser propulsion.« less

FALCON nuclear-reactor-pumped laser program and wireless power transmission

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

Lipinski, R.J.; Pickard, P.S.

1992-01-01

FALCON is a high-power, reactor-pumped laser concept. The major strengths of a reactor-pumped laser are (1) simple, modular construction, (2) long-duration, closed-cycle capability, (3) self-contained power, (4) compact size, and (5) a variety of wavelengths (from visible to infrared). Reactor-pumped lasing has been demonstrated experimentally in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Powers up to 300 W for 2 ms have been demonstrated. Projected beam quality for FALCON is good enough that frequency doubling at reasonablemore » efficiencies could be expected to yield wavelengths at 353, 363, 636, 867, 896, 1016, 1315, 1325, and 1685 nm. Appropriate missions for FALCON are described and include power beaming to satellites, the moon, and unmanned surveillance planes; lunar mapping; space debris removal; and laser propulsion.« less

Fusion reactor pumped laser

DOEpatents

Jassby, D.L.

1987-09-04

A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

Fusion reactor pumped laser

DOEpatents

Jassby, Daniel L.

1988-01-01

A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

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

Two-phase flow in the cooling circuit of a cryogenic rocket engine

NASA Astrophysics Data System (ADS)

Preclik, D.

1992-07-01

Transient two-phase flow was investigated for the hydrogen cooling circuit of the HM7 rocket engine. The nuclear reactor code ATHLET/THESEUS was adapted to cryogenics and applied to both principal and prototype experiments for validation and simulation purposes. The cooling circuit two-phase flow simulation focused on the hydrogen prechilling and pump transient phase prior to ignition. Both a single- and a multichannel model were designed and employed for a valve leakage flow, a nominal prechilling flow, and a prechilling with a subsequent pump-transient flow. The latter case was performed in order to evaluate the difference between a nominal and a delayed turbo-pump start-up. It was found that an extension of the nominal prechilling sequence in the order of 1 second is sufficient to finally provide for liquid injection conditions of hydrogen which, as commonly known, is undesirable for smooth ignition and engine starting transients.

Analysis of boron dilution in a four-loop PWR

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

Sun, J.G.; Sha, W.T.

1995-03-01

Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four-loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plant ismore » at hot standby and the reactor coolant system has been heated to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated and a sensitivity study is performed to assess the accuracy of the numerical modeling of the geometry of the reactor coolant system.« less

77 FR 41814 - Entergy Operations, Inc.; Grand Gulf Nuclear Station, Unit 1

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-16

... Unit 1 result primarily from periodic testing of diesel generators and fire water pump diesel engines... rural. GGNS Unit 1 is a General Electric Mark 3 boiling-water reactor. Identification of the Proposed... following: replacing the reactor feed pump turbine rotors; replacing the main generator current transformers...

Design of a Mechanical NaK Pump for Fission Space Power Systems

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Bradley, David; Godfroy, Thomas

2010-01-01

Alkali liquid metal cooled fission reactor concepts are under development for mid-range spaceflight power requirements. One such concept utilizes a sodium-potassium eutectic (NaK) as the primary loop working fluid. Traditionally, linear induction pumps have been used to provide the required flow and head conditions for liquid metal systems but can be limited in performance. This paper details the design, build, and check-out test of a mechanical NaK pump. The pump was designed to meet reactor cooling requirements using commercially available components modified for high temperature NaK service.

Design of a Mechanical NaK Pump for Fission Space Power

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Bradley, David E.; Godfroy, Thomas

2011-01-01

Alkali liquid metal cooled fission reactor concepts are under development for spaceflight power requirements. One such concept utilizes a sodium-potassium eutectic (NaK) as the primary loop working fluid, which has specific pumping requirements. Traditionally, electromagnetic linear induction pumps have been used to provide the required flow and pressure head conditions for NaK systems but they can be limited in performance, efficiency, and number of available vendors. The objective of the project was to develop a mechanical NaK centrifugal pump that takes advantages of technology advances not available in previous liquid metal mechanical pump designs. This paper details the design, build, and performance test of a mechanical NaK pump developed at NASA Marshall Space Flight Center. The pump was designed to meet reactor cooling requirements using commercially available components modified for high temperature NaK service.

Space rocket engine on the base of the reactor-pumped laser for the interplanetary flights and earth orbital applications

NASA Astrophysics Data System (ADS)

Gulevich, Andrey V.; Dyachenko, Peter P.; Kukharchuk, Oleg F.; Zrodnikov, Anatoly V.

2000-01-01

In this report the concept of vehicle-based reactor-laser engine for long time interplanetary and interorbital (LEO to GEO) flights is proposed. Reactor-pumped lasers offer the perspective way to create on the base of modern nuclear and lasers technologies the low mass and high energy density, repetitively pulsed vehicle-based laser of average power 100 kW. Nowadays the efficiency of nuclear-to-optical energy conversion reached the value of 2-3%. The demo model of reactor-pumped laser facility is under construction in Institute for Physics and Power Engineering (Obninsk, Russia). It enable us to hope that using high power laser on board of the vehicle could make the effective space laser engine possible. Such engine may provide the high specific impulse ~1000-2000 s with the thrust up to 10-100 n. Some calculation results of the characteristics of vehicle-based reactor-laser thermal engine concept are also presented. .

PUMP HOUSE FOR MTR WELL NO. 1, TRA601. FLOOR PLAN, ...

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

PUMP HOUSE FOR MTR WELL NO. 1, TRA-601. FLOOR PLAN, ELEVATIONS, SECTION SHOWING WELL CASING, ROOF FRAMING PLAN. AS BUILT. WELL HOUSE FOR WELL NO. 2, TRA-602, WAS IDENTICAL IN ALL PARTICULARS EXCEPT FLOOR DIMENSIONS AND ARRANGEMENT OF PUMP AND ELECTRICAL EQUIPMENT INSIDE. IDAHO OPERATIONS OFFICE MTR-601-IDO-1, 12/1954. INL INDEX NO. 531-0601-00-396-110463, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Control Optimization for a Dual-Mode Single-State Nuclear Shuttle,

DTIC Science & Technology

1980-01-01

Variables at a •.2 as Functions of the Pump ! Power# ..... ............ ......... 36 ’’i I ’I [ I I OIAPTER I INTRODUCTION Since the end of the Apollo...If this is not the case, the assIumption is slightly optimistic. 4. The effective pump power and the reactor-exit stagnation tempar- ature are...independent of the reactor-exit stagnation pressure. I ("Effective puImp power" is the power required to pump the propellants, assumed to be incompressible

Section 7 reactor incident file general information from 1945

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

Not Available

1969-01-10

At 0308 on January 10, 1966, both B and C Reactors ``scrammed`` due to an electrical fault on Line C2-L8 caused by a raccoon coming in contact with the 13-8 KV line on top of transformer No. 2 at 182-B Building. Line C2-L8 relayed out at the 151-B Building. Details of the occurrence at 151-B are covered in the attachment. C-Reactor scrammed due to reduced voltage on the pressure monitor system. The reduction in voltage caused the auxiliary relays of the pressure monitor ground detector to open, de-energizing the end result relays PSR and PSRA. The safety circuit trip identificationmore » system displayed ``Pressure Monitor`` and ``Ground Detector.`` B-Reactor scrammed by a power failure signal from 190-B Building. The power failure relays for pump numbers 1 and 3 opened due to these pumps contributing power to the fault. The power failure relays at 190-B remained open long enough for the end result relays PF and PFA to open. Since these relays are timed delayed, 0.26 seconds, the power failure relays must have remained open at least that long. At the 190-B Building the steam turbines started due to the power failure relays for pump numbers 1 and 3 opening. The main process pumps remained stable and continued to supply normal flow to the reactor. Pumps were tripped from the line at 182-B and 183-B Buildings. The surge suppressors cycled normally and the turbine export pumps started as a result of low export line pressure. No power equipment was affected in C Area.« less

Fluid sampling system for a nuclear reactor

DOEpatents

Lau, Louis K.; Alper, Naum I.

1994-01-01

A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump.

Fluid sampling system for a nuclear reactor

DOEpatents

Lau, L.K.; Alper, N.I.

1994-11-22

A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump. 1 fig.

Operational Test Report (OTR): On-Site Degradation of Oily Sludge in a Tenthousand Gallon Sequencing Batch Reactor at Navsta Pearl Harbor, HI

DTIC Science & Technology

2003-11-01

treated anaerobically . To accommodate the longer residence times needed to treat waste anaerobically , the capacity is often much larger than a...the receiving tank (T1), where it is diluted and run through a trash pump (P1) to produce a homogenous slurry. 3 Figure 1. Sequencing...blower provides air to the reactor and receiving tank. The trash pump is also used to transfer sludge to the reactor and to recirculate sludge in

Thermodynamic Analysis of the Use a Chemical Heat Pump to Link a Supercritical Water-Cooled Nuclear Reactor and a Thermochemical Water-Splitting Cycle for Hydrogen Production

NASA Astrophysics Data System (ADS)

Granovskii, Mikhail; Dincer, Ibrahim; Rosen, Marc A.; Pioro, Igor

Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved “steam” parameters (outlet temperatures up to 625°C and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600°C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the “nuclear” heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of the combined system comprising a SCW nuclear power generation plant and a chemical heat pump, which provides high-temperature heat to a thermochemical water splitting cycle for hydrogen production. It is concluded that the proposed chemical heat pump permits the utilization efficiency of nuclear energy to be improved by at least 2% without jeopardizing nuclear reactor safety. Based on this analysis, further research appears to be merited on the proposed advanced design of a nuclear power generation plant combined with a chemical heat pump, and implementation in appropriate applications seems worthwhile.

Heat-Powered Pump for Liquid Metals

NASA Technical Reports Server (NTRS)

Campana, R. J.

1986-01-01

Proposed thermoelectromagnetic pump for liquid metal powered by waste heat; needs no battery, generator, or other external energy source. Pump turns part of heat in liquid metal into pumping energy. In combination with primary pump or on its own, thermoelectric pump circulates coolant between reactor and radiator. As long as there is decay heat to be removed, unit performs function.

Reactor coolant pump flywheel

DOEpatents

Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

2013-11-26

A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

Attrition reactor system

DOEpatents

Scott, Charles D.; Davison, Brian H.

1993-01-01

A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

Bottom head to shell junction assembly for a boiling water nuclear reactor

DOEpatents

Fife, Alex Blair; Ballas, Gary J.

1998-01-01

A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening.

Bottom head to shell junction assembly for a boiling water nuclear reactor

DOEpatents

Fife, A.B.; Ballas, G.J.

1998-02-24

A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening. 5 figs.

LMFBR with booster pump in pumping loop

DOEpatents

Rubinstein, H.J.

1975-10-14

A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation.

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

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

Kirkland, Karen Vierow; Ross, Kyle; Beeny, Bradley

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that themore » system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.« less

76 FR 16842 - Request for a License To Export Reactor Components

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-25

... NUCLEAR REGULATORY COMMISSION Request for a License To Export Reactor Components Pursuant to 10.... Mechanical Corporation. coolant pump 1000 (design) maintenance, and systems, related reactors. operation of AP- equipment, and 1000 (design) spare parts. nuclear reactors. February 10, 2011 February 23, 2011...

Attrition reactor system

DOEpatents

Scott, C.D.; Davison, B.H.

1993-09-28

A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

Method for producing components with internal architectures, such as micro-channel reactors, via diffusion bonding sheets

DOEpatents

Alman, David E [Corvallis, OR; Wilson, Rick D [Corvallis, OR; Davis, Daniel L [Albany, OR

2011-03-08

This invention relates to a method for producing components with internal architectures, and more particularly, this invention relates to a method for producing structures with microchannels via the use of diffusion bonding of stacked laminates. Specifically, the method involves weakly bonding a stack of laminates forming internal voids and channels with a first generally low uniaxial pressure and first temperature such that bonding at least between the asperites of opposing laminates occurs and pores are isolated in interfacial contact areas, followed by a second generally higher isostatic pressure and second temperature for final bonding. The method thereby allows fabrication of micro-channel devices such as heat exchangers, recuperators, heat-pumps, chemical separators, chemical reactors, fuel processing units, and combustors without limitation on the fin aspect ratio.

IN-PILE CORROSION TEST LOOPS FOR AQUEOUS HOMOGENEOUS REACTOR SOLUTIONS

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

Savage, H.C.; Jenks, G.H.; Bohlmann, E.G.

1960-12-21

An in-pile corrosion test loop is described which is used to study the effect of reactor radiation on the corrosion of materials of construction and the chemical stability of fuel solutions of interest to the Aqueous Homogeneous Reactor Program at ORNL. Aqueous solutions of uranyl sulfate are circulated in the loop by means of a 5-gpm canned-rotor pump, and the pump loop is designed for operation at temperatures to 300 ts C and pressures to 2000 psia while exposed to reactor radiation in beam-hole facilities of the LITR and ORR. Operation of the first loop in-pile was begun in Octobermore » 1954, and since that time 17 other in-pile loop experiments were completed. Design criteria of the pump loop and its associated auxiliary equipment and instrumentation are described. In-pile operating procedures, safety features, and operating experience are presented. A cost summary of the design, fabrication, and installation of the loop and experimental facillties is also included. (auth)« less

On the possibility of connecting a non-operating main circulation pump with three pumps in operation without preliminary coast-down of power-generating unit No. 5 in the Novovoronezh nuclear power plant

NASA Astrophysics Data System (ADS)

Vitkovskii, I. L.; Nikonov, S. P.; Ryasnyi, S. I.

2014-02-01

The subject of this paper is a transient caused by connection of a standby loop to three operating circulation pumps at the initial reactor heat rate equal to 70% of the rated value without preliminarily reducing it to 30% of the rated level as required by the safe operation regulations. Failure of the following normal operation systems is supposed: the first- and the second-type warning protection systems, all quick-acting reducing devices releasing steam into the auxiliary manifold, the electric heaters of the pressurizer, the pressurizer injection system, the primary cooling circuit fluid makeup/blow-through systems, and the blocking systems to shut down the main circulation pump after the level in the steam generator is exceeded. In addition, it is supposed that, under transient conditions, the valves of the turbine regulation system will be in the position in which they were at the moment of the initial event until generation of the signal for positive closing of the turbine stop valves. The first signal to actuate the reactor emergency protection system (EPS) is skipped. The failure of all quick-acting reducing devices releasing steam into the atmosphere is assumed. In addition to equipment failure, at the moment when the main circulation pump is connected, the operator erroneously puts in a new setting to maintain the power allowable for four pumps in operation-in the calculations it was taken equal to 104% of the rated level at most considering the accuracy of evaluating and maintaining the reactor heat rate-and the working group of the reactor protection and control system (P&CS) starts moving upward. On reaching the set power level, the automatic reactor power regulator stops operating and the P&CS elements remain in the position in which they are at the moment. Compliance with the design safety criteria for the adopted scenario of the transient is demonstrated.

Space nuclear system thermoelectric NaK pump development

NASA Technical Reports Server (NTRS)

Johnson, J. L.

1973-01-01

The engineering, design, fabrication, and test history of the dual-throat thermoelectric NaK development pump is summarized, along with the engineering and design status of a similar prototype pump intended for use on the 5-kwe reactor thermoelectric system. The history of dc pump development and testing on previous programs is also summarized.

A Semi-Batch Reactor Experiment for the Undergraduate Laboratory

ERIC Educational Resources Information Center

Derevjanik, Mario; Badri, Solmaz; Barat, Robert

2011-01-01

This experiment and analysis offer an economic yet challenging semi-batch reactor experience. Household bleach is pumped at a controlled rate into a batch reactor containing pharmaceutical hydrogen peroxide solution. Batch temperature, product molecular oxygen, and the overall change in solution conductivity are metered. The reactor simulation…

A liquid-metal filling system for pumped primary loop space reactors

NASA Astrophysics Data System (ADS)

Crandall, D. L.; Reed, W. C.

Some concepts for the SP-100 space nuclear power reactor use liquid metal as the primary coolant in a pumped loop. Prior to filling ground engineering test articles or reactor systems, the liquid metal must be purified and circulated through the reactor primary system to remove contaminants. If not removed, these contaminants enhance corrosion and reduce reliability. A facility was designed and built to support Department of Energy Liquid Metal Fast Breeder Reactor tests conducted at the Idaho National Engineering Laboratory. This test program used liquid sodium to cool nuclear fuel in in-pile experiments; thus, a system was needed to store and purify sodium inventories and fill the experiment assemblies. This same system, with modifications and potential changeover to lithium or sodium-potassium (NaK), can be used in the Space Nuclear Power Reactor Program. This paper addresses the requirements, description, modifications, operation, and appropriateness of using this liquid-metal system to support the SP-100 space reactor program.

Fission fragment assisted reactor concept for space propulsion: Foil reactor

NASA Technical Reports Server (NTRS)

Wright, Steven A.

1991-01-01

The concept is to fabricate a reactor using thin films or foils of uranium, uranium oxide and then to coat them on substrates. These coatings would be made so thin as to allow the escaping fission fragments to directly heat a hydrogen propellant. The idea was studied of direct gas heating and direct gas pumping in a nuclear pumped laser program. Fission fragments were used to pump lasers. In this concept two substrates are placed opposite each other. The internal faces are coated with thin foil of uranium oxide. A few of the advantages of this technology are listed. In general, however, it is felt that if one look at all solid core nuclear thermal rockets or nuclear thermal propulsion methods, one is going to find that they all pretty much look the same. It is felt that this reactor has higher potential reliability. It has low structural operating temperatures, very short burn times, with graceful failure modes, and it has reduced potential for energetic accidents. Going to a design like this would take the NTP community part way to some of the very advanced engine designs, such as the gas core reactor, but with reduced risk because of the much lower temperatures.

Nuclear-Pumped Lasers. [efficient conversion of energy liberated in nuclear reactions to coherent radiation

NASA Technical Reports Server (NTRS)

1979-01-01

The state of the art in nuclear pumped lasers is reviewed. Nuclear pumped laser modeling, nuclear volume and foil excitation of laser plasmas, proton beam simulations, nuclear flashlamp excitation, and reactor laser systems studies are covered.

Flow rate analysis of wastewater inside reactor tanks on tofu wastewater treatment plant

NASA Astrophysics Data System (ADS)

Mamat; Sintawardani, N.; Astuti, J. T.; Nilawati, D.; Wulan, D. R.; Muchlis; Sriwuryandari, L.; Sembiring, T.; Jern, N. W.

2017-03-01

The research aimed to analyse the flow rate of the wastewater inside reactor tanks which were placed a number of bamboo cutting. The resistance of wastewater flow inside reactor tanks might not be occurred and produce biogas fuel optimally. Wastewater from eleven tofu factories was treated by multi-stages anaerobic process to reduce its organic pollutant and produce biogas. Biogas plant has six reactor tanks of which its capacity for waste water and gas dome was 18 m3 and 4.5 m3, respectively. Wastewater was pumped from collecting ponds to reactors by either serial or parallel way. Maximum pump capacity, head, and electrical motor power was 5m3/h, 50m, and 0.75HP, consecutively. Maximum pressure of biogas inside the reactor tanks was 55 mbar higher than atmosphere pressure. A number of 1,400 pieces of cutting bamboo at 50-60 mm diameter and 100 mm length were used as bacteria growth media inside each reactor tank, covering around 14,287 m2 bamboo area, and cross section area of inner reactor was 4,9 m2. In each reactor, a 6 inches PVC pipe was installed vertically as channel. When channels inside reactor were opened, flow rate of wastewater was 6x10-1 L.sec-1. Contrary, when channels were closed on the upper part, wastewater flow inside the first reactor affected and increased gas dome. Initially, wastewater flowed into each reactor by a gravity mode with head difference between the second and third reactor was 15x10-2m. However, head loss at the second reactor was equal to the third reactor by 8,422 x 10-4m. As result, wastewater flow at the second and third reactors were stagnant. To overcome the problem pump in each reactor should be installed in serial mode. In order to reach the output from the first reactor and the others would be equal, and biogas space was not filled by wastewater, therefore biogas production will be optimum.

Self-pumping impurity control

DOEpatents

Brooks, J.N.; Mattas, R.F.

1983-12-21

It is an object of the present invention to provide an apparatus for removing impurities from the plasma in a fusion reactor without an external vacuum pumping system. It is also an object of the present invention to provide an apparatus for removing the helium ash from a fusion reactor. It is another object of the present invention to provide an apparatus which removes helium ash and minimizes tritium recycling and inventory.

Analysis of failed nuclear plant components

NASA Astrophysics Data System (ADS)

Diercks, D. R.

1993-12-01

Argonne National Laboratory has conducted analyses of failed components from nuclear power- gener-ating stations since 1974. The considerations involved in working with and analyzing radioactive compo-nents are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in serv-ice. The failures discussed are (1) intergranular stress- corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor.

Self-actuated nuclear reactor shutdown system using induction pump to facilitate sensing of core coolant temperature

DOEpatents

Sievers, Robert K.; Cooper, Martin H.; Tupper, Robert B.

1987-01-01

A self-actuated shutdown system incorporated into a reactivity control assembly in a nuclear reactor includes pumping means for creating an auxiliary downward flow of a portion of the heated coolant exiting from the fuel assemblies disposed adjacent to the control assembly. The shutdown system includes a hollow tubular member which extends through the outlet of the control assembly top nozzle so as to define an outer annular flow channel through the top nozzle outlet separate from an inner flow channel for primary coolant flow through the control assembly. Also, a latching mechanism is disposed in an inner duct of the control assembly and is operable for holding absorber bundles in a raised position in the control assembly and for releasing them to drop them into the core of the reactor for shutdown purposes. The latching mechanism has an inner flow passage extending between and in flow communication with the absorber bundles and the inner flow channel of the top nozzle for accommodating primary coolant flow upwardly through the control assembly. Also, an outer flow passage separate from the inner flow passage extends through the latching mechanism between and in flow communication with the inner duct and the outer flow channel of the top nozzle for accommodating inflow of a portion of the heated coolant from the adjacent fuel assemblies. The latching mechanism contains a magnetic material sensitive to temperature and operable to cause mating or latching together of the components of the latching mechanism when the temperature sensed is below a known temperature and unmating or unlatching thereof when the temperature sensed is above a given temperature. The temperature sensitive magnetic material is positioned in communication with the heated coolant flow through the outer flow passage for directly sensing the temperature thereof. Finally, the pumping means includes a jet induction pump nozzle and diffuser disposed adjacent the bottom nozzle of the control assembly and in flow communication with the inlet thereof. The pump nozzle is operable to create an upward driving flow of primary coolant through the pump diffuser and then to the absorber bundles. The upward driving flow of primary coolant, in turn, creates a suction head within the outer flow channel of the top nozzle and thereby an auxiliary downward flow of the heated coolant portion exiting from the upper end of the adjacent fuel assemblies through the outer flow channel to the pump nozzle via the outer flow passage of the latching mechanism and an annular space between the outer and inner spaced ducts of the control assembly housing. The temperature of the heated coolant exiting from the adjacent fuel assemblies can thereby be sensed directly by the temperature sensitive magnetic material in the latching mechanism.

Cracked shaft detection on large vertical nuclear reactor coolant pump

NASA Technical Reports Server (NTRS)

Jenkins, L. S.

1985-01-01

Due to difficulty and radiation exposure associated with examination of the internals of large commercial nuclear reactor coolant pumps, it is necessary to be able to diagnose the cause of an excessive vibration problem quickly without resorting to extensive trial and error efforts. Consequently, it is necessary to make maximum use of all available data to develop a consistent theory which locates the problem area in the machine. This type of approach was taken at Three Mile Island, Unit #1, in February 1984 to identify and locate the cause of a continuously climbing vibration level of the pump shaft. The data gathered necessitated some in-depth knowledge of the pump internals to provide proper interpretation and avoid misleading conclusions. Therefore, the raw data included more than just the vibration characteristics. Pertinent details of the data gathered is shown and is necessary and sufficient to show that the cause of the observed vibration problem could logically only be a cracked pump shaft in the shaft overhang below the pump bearing.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 1: Objectives, summary results and introduction

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

The objective was to determine which reactor, conversion, and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. Specifically, the requirement was 10 megawatts for 5 years of full power operation and 10 years systems life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study. The concepts are: a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heat pipe and pumped tube-fin heat rejection; a lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator; a lithium cooled reactor with potassium Rankine turbine-alternator and heat pipe radiator; and a lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the lithium cooled incore thermionic reactor with heat pipe radiator.

ETR COOLING TOWER. PUMP HOUSE (TRA645) IN SHADOW OF TOWER ...

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

ETR COOLING TOWER. PUMP HOUSE (TRA-645) IN SHADOW OF TOWER ON LEFT. AT LEFT OF VIEW, HIGH-BAY BUILDING IS ETR. ONE STORY ATTACHMENT IS ETR ELECTRICAL BUILDING. STACK AT RIGHT IS ETR STACK; MTR STACK IS TOWARD LEFT. CAMERA FACING NORTHEAST. INL NEGATIVE NO. 56-3799. Jack L. Anderson, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

SP-100 system thaw and start-up strategies

NASA Astrophysics Data System (ADS)

Kirpich, A.; Choe, H.

The authors review several strategies that have been considered during calendar year 1990 for SP-100 system start-up and thaw, and provide results of screening analyses. Screening studies have identified three concepts which are capable of thawing the SP-100 system: (1) a heat pipe approach in which reactor-heated heat pipes are routed to equipment enclosed within a thaw cavity; (2) a bleed-tube approach in which perforated tubes routed through ducting and components achieve thaw progressively as the result of successively thawing bleed-holes; and (3) a NaK traceline approach in which reactor-heated NaK is routed along ducting and components and achieves lithium thaw by both radiative and conductive coupling methods. Key issues have been identified in connection with pump start-up, reactor and power converter transient behavior, and hydraulic circuit stability. Pump start-up stability is accomplished by a linked-pump hydraulic arrangement.

Appendix to HDC 2118 design criteria 100-X reactor water plant, general description - section II

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

Not Available

1952-03-29

The factors responsible for the advances of 100-X compared with the older areas are: Simplification of the process, such as elimination of separate process water clearwells, by having the filtered water reservoirs perform that function. Combination of separate buildings into one building, such as combining filter pump house and process pump house. Use of electric standby. Use of higher capacity pumps and filter basins, and so fewer number of units. Centralization of control and operation. More compact arrangement of plant components. Use of waste heat for space heating, recovered from reactor effluent, backed up by steam plant.

Optimization of post-column reactor radius in capillary high performance liquid chromatography Effect of chromatographic column diameter and particle diameter

PubMed Central

Xu, Hongjuan; Weber, Stephen G.

2006-01-01

A post-column reactor consisting of a simple open tube (Capillary Taylor Reactor) affects the performance of a capillary LC in two ways: stealing pressure from the column and adding band spreading. The former is a problem for very small radius reactors, while the latter shows itself for large reactor diameters. We derived an equation that defines the observed number of theoretical plates (Nobs) taking into account the two effects stated above. Making some assumptions and asserting certain conditions led to a final equation with a limited number of variables, namely chromatographic column radius, reactor radius and chromatographic particle diameter. The assumptions and conditions are that the van Deemter equation applies, the mass transfer limitation is for intraparticle diffusion in spherical particles, the velocity is at the optimum, the analyte’s retention factor, k′, is zero, the post-column reactor is only long enough to allow complete mixing of reagents and analytes and the maximum operating pressure of the pumping system is used. Optimal ranges of the reactor radius (ar) are obtained by comparing the number of observed theoretical plates (and theoretical plates per time) with and without a reactor. Results show that the acceptable reactor radii depend on column diameter, particle diameter, and maximum available pressure. Optimal ranges of ar become narrower as column diameter increases, particle diameter decreases or the maximum pressure is decreased. When the available pressure is 4000 psi, a Capillary Taylor Reactor with 12 μm radius is suitable for all columns smaller than 150 μm (radius) packed with 2–5 μm particles. For 1 μm packing particles, only columns smaller than 42.5 μm (radius) can be used and the reactor radius needs to be 5 μm. PMID:16494886

CRBR pump water test experience

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

Cook, M.E.; Huber, K.A.

1983-01-01

The hydraulic design features and water testing of the hydraulic scale model and prototype pump of the sodium pumps used in the primary and intermediate sodium loops of the Clinch River Breeder Reactor Plant (CRBRP) are described. The Hydraulic Scale Model tests are performed and the results of these tests are discussed. The Prototype Pump tests are performed and the results of these tests are discussed.

Nuclear reactor with makeup water assist from residual heat removal system

DOEpatents

Corletti, Michael M.; Schulz, Terry L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

Nuclear reactor with makeup water assist from residual heat removal system

DOEpatents

Corletti, M.M.; Schulz, T.L.

1993-12-07

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

In-situ Condition Monitoring of Components in Small Modular Reactors Using Process and Electrical Signature Analysis. Final report, volume 1. Development of experimental flow control loop, data analysis and plant monitoring

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

Upadhyaya, Belle; Hines, J. Wesley; Damiano, Brian

The research and development under this project was focused on the following three major objectives: Objective 1: Identification of critical in-vessel SMR components for remote monitoring and development of their low-order dynamic models, along with a simulation model of an integral pressurized water reactor (iPWR). Objective 2: Development of an experimental flow control loop with motor-driven valves and pumps, incorporating data acquisition and on-line monitoring interface. Objective 3: Development of stationary and transient signal processing methods for electrical signatures, machinery vibration, and for characterizing process variables for equipment monitoring. This objective includes the development of a data analysis toolbox. Themore » following is a summary of the technical accomplishments under this project: - A detailed literature review of various SMR types and electrical signature analysis of motor-driven systems was completed. A bibliography of literature is provided at the end of this report. Assistance was provided by ORNL in identifying some key references. - A review of literature on pump-motor modeling and digital signal processing methods was performed. - An existing flow control loop was upgraded with new instrumentation, data acquisition hardware and software. The upgrading of the experimental loop included the installation of a new submersible pump driven by a three-phase induction motor. All the sensors were calibrated before full-scale experimental runs were performed. - MATLAB-Simulink model of a three-phase induction motor and pump system was completed. The model was used to simulate normal operation and fault conditions in the motor-pump system, and to identify changes in the electrical signatures. - A simulation model of an integral PWR (iPWR) was updated and the MATLAB-Simulink model was validated for known transients. The pump-motor model was interfaced with the iPWR model for testing the impact of primary flow perturbations (upsets) on plant parameters and the pump electrical signatures. Additionally, the reactor simulation is being used to generate normal operation data and data with instrumentation faults and process anomalies. A frequency controller was interfaced with the motor power supply in order to vary the electrical supply frequency. The experimental flow control loop was used to generate operational data under varying motor performance characteristics. Coolant leakage events were simulated by varying the bypass loop flow rate. The accuracy of motor power calculation was improved by incorporating the power factor, computed from motor current and voltage in each phase of the induction motor.- A variety of experimental runs were made for steady-state and transient pump operating conditions. Process, vibration, and electrical signatures were measured using a submersible pump with variable supply frequency. High correlation was seen between motor current and pump discharge pressure signal; similar high correlation was exhibited between pump motor power and flow rate. Wide-band analysis indicated high coherence (in the frequency domain) between motor current and vibration signals. - Wide-band operational data from a PWR were acquired from AMS Corporation and used to develop time-series models, and to estimate signal spectrum and sensor time constant. All the data were from different pressure transmitters in the system, including primary and secondary loops. These signals were pre-processed using the wavelet transform for filtering both low-frequency and high-frequency bands. This technique of signal pre-processing provides minimum distortion of the data, and results in a more optimal estimation of time constants of plant sensors using time-series modeling techniques.« less

WATER PUMP HOUSE, TRA619, AND TWO WATER STORAGE RESERVOIRS. INDUSTRIAL ...

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

WATER PUMP HOUSE, TRA-619, AND TWO WATER STORAGE RESERVOIRS. INDUSTRIAL WINDOWS AND COPING STRIPS AT TOP OF WALLS AND ENTRY VESTIBULE. BOLLARDS PROTECT UNDERGROUND FACILITIES. SWITCHYARD AT RIGHT EDGE OF VIEW. CARD IN LOWER RIGHT WAS INSERTED BY INL PHOTOGRAPHER TO COVER AN OBSOLETE SECURITY RESTRICTION PRINTED ON ORIGINAL NEGATIVE. INL NEGATIVE NO. 3816. Unknown Photographer, 11/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Ichthyoplankton entrainment study at the SRS Savannah River water intakes for Westinghouse Savannah River Company

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

Paller, M.

1992-03-26

Cooling water for L and K Reactors and makeup water for Par Pond is pumped from the Savannah River at the 1G, 3G, and 5G pump houses. Ichthyoplankton (drifting fish larvae and eggs) from the river are entrained into the reactor cooling systems with the river water and passed through the reactor's heat exchangers where temperatures may reach 70[degrees]C during full power operation. Ichthyoplankton mortality under such conditions is assumed to be 100 percent. The number of ichthyoplankton entrained into the cooling system depends on a variety of variables, including time of year, density and distribution of ichthyoplankton in themore » river, discharge levels in the river, and the volume of water withdrawn by the pumps. Entrainment at the 1 G pump house, which is immediately downstream from the confluence of Upper Three Runs Creek and the Savannah River, is also influenced by discharge rates and ichthyoplankton densities in Upper Three Runs Creek. Because of the anticipated restart of several SRS reactors and the growing concern surrounding striped bass and American shad stocks in the Savannah River, the Department of Energy requested that the Environmental Sciences Section (ESS) of the Savannah River Laboratory sample ichthyoplankton at the SRS Savannah River intakes. Dams Moore, Inc., under a contract with Westinghouse Savannah River Company performed the sampling and data analysis for the ESS.« less

Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

NASA Astrophysics Data System (ADS)

Fic, Adam; Składzień, Jan; Gabriel, Michał

2015-03-01

Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-02-01

A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK® platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 4: Concepts selection, conceptual designs, recommendations

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

A study was conducted by NASA Lewis Research Center for the Triagency SP-100 program office. The objective was to determine which reactor, conversion and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. The requirement was 10 megawatts for 5 years of full power operation and 10 years system life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study: (1) a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heatpipe and pumped tube fin rejection, (2) a Lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator,(3) a Lithium cooled reactor with a Potassium Rankine turbine-alternator and heat pipe radiator, and (4) a Lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the Lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the Lithium cooled incore thermionic reactor with heat pipe radiator.

77 FR 39521 - Application for a License To Export Nuclear Reactor Major Components and Equipment

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-03

... LLC reactor coolant equipment for four constructing four plant May 14, 2012 pumps with motors, APR1400... Emirates. XR176 monitoring and plant in Braka. 110060011 control equipment, auxiliary equipment and... NUCLEAR REGULATORY COMMISSION Application for a License To Export Nuclear Reactor Major Components...

Testing of Liquid Metal Components for Nuclear Surface Power Systems

NASA Technical Reports Server (NTRS)

Polzin, Kurt A.; Godfroy, Thomas J.; Pearson, J. Boise

2010-01-01

The Early Flight Fission Test Facility (EFF-TF) was established by the Marshall Space Flight Center (MSFC) to provide a capability for performing hardware-directed activities to support multiple in-space nuclear reactor concepts by using a non-nuclear test methodology. This includes fabrication and testing at both the module/component level and near prototypic reactor configurations. The EFF-TF is currently supporting an effort to develop an affordable fission surface power (AFSP) system that could be deployed on the Lunar surface. The AFSP system is presently based on a pumped liquid metal-cooled (Sodium-Potassium eutectic, NaK-78) reactor design. This design was derived from the only fission system that the United States has deployed for space operation, the Systems for Nuclear Auxiliary Power (SNAP) 10A reactor, which was launched in 1965. Two prototypical components recently tested at MSFC were a pair of Stirling power conversion units that would be used in a reactor system to convert heat to electricity, and an annular linear induction pump (ALIP) that uses travelling electromagnetic fields to pump the liquid metal coolant through the reactor loop. First ever tests were conducted at MSFC to determine baseline performance of a pair of 1 kW Stirling convertors using NaK as the hot side working fluid. A special test rig was designed and constructed and testing was conducted inside a vacuum chamber at MSFC. This test rig delivered pumped NaK for the hot end temperature to the Stirlings and water as the working fluid on the cold end temperature. These test were conducted through a hot end temperature range between 400 to 550C in increments of 50 C and a cold end temperature range from 30 to 70 C in 20 C increments. Piston amplitudes were varied from 6 to 1 1mm in .5 mm increments. A maximum of 2240 Watts electric was produced at the design point of 550 hot end, 40 C cold end with a piston amplitude of 10.5mm. This power level was reached at a gross thermal efficiency of 28%. A baseline performance map was established for the pair of 1kW Stirling convertors. The performance data will then be used for design modification to the Stirling convertors. The ALIP tested at MSFC has no moving parts and no direct electrical connections to the liquid metal containing components. Pressure is developed by the interaction of the magnetic field produced by the stator and the current which flows as a result of the voltage induced in the liquid metal contained in the pump duct. Flow is controlled by variation of the voltage supplied to the pump windings. Under steady-state conditions, pump performance is measured for flow rates from 0.5-4.3 kg/s. The pressure rise developed by the pump to support these flow rates is roughly 5-65 kPa. The RMS input voltage (phase-to-phase voltage) ranges from 5-120 V, while the frequency can be varied arbitrarily up to 60 Hz. Performance is quantified at different loop temperature levels from 50 C up to 650 C, which is the peak operating temperature of the proposed AFSP reactor. The transient response of the pump is also evaluated to determine its behavior during startup and shut-down procedures.

Effect of the self-pumped limiter concept on the tritium fuel cycle

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

Finn, P.A.; Sze, D.K.; Hassanein, A.

1988-09-01

The self-pumped limiter concept was the impurity control system for the reactor in the Tokamak Power Systems Study (TPSS). The use of a self-pumped limiter had a major impact on the design of the tritium systems of the TPSS fusion reactor. The self-pumped limiter functions by depositing the helium ash under a layer of metal (vanadium). The majority of the hydrogen species are recycled at the plasma edge; a small fraction permeates to the blanket/coolant which was lithium in TPSS. Use of the self-pumped limiter results in the elimination of the plasma processing system. Thus, the blanket tritium processing systemmore » becomes the major tritium system. The main advantages achieved for the tritium systems with a self-pumped limiter are a reduction in the capital cost of tritium processing equipment as well as a reduction in building space, a reduced tritium inventory for processing and for reserve storage, an increase in the inherent safety of the fusion plant and an improvement in economics for a fusion world economy.« less

Novel limiter pump topologies

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

Schultz, J.H.

1981-01-01

The use of limiter pumps as the principle plasma exhaust system of a magnetic confinement fusion device promises significant simplification, when compared to previously investigating divertor based systems. Further simplifications, such as the integration of the exhaust system with a radio frequency heating system and with the main reactor shield and structure are investigated below. The integrity of limiters in a reactor environment is threatened by many mechanisms, the most severe of which may be erosion by sputtering. Two novel topolgies are suggested which allow high erosion without limiter failure.

DEMINERALIZER BUILDING,TRA608. CAMERA FACES EAST ALONG SOUTH WALL. INSTRUMENT PANEL ...

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

DEMINERALIZER BUILDING,TRA-608. CAMERA FACES EAST ALONG SOUTH WALL. INSTRUMENT PANEL BOARD IS IN RIGHT HALF OF VIEW, WITH FOUR PUMPS BEYOND. SMALLER PUMPS FILL DEMINERALIZED WATER TANK ON SOUTH SIDE OF BUILDING. CARD IN LOWER RIGHT WAS INSERTED BY INL PHOTOGRAPHER TO COVER AN OBSOLETE SECURITY RESTRICTION PRINTED ON ORIGINAL NEGATIVE. INL NEGATIVE NO. 3997A. Unknown Photographer, 12/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Stainless Steel NaK-Cooled Circuit (SNaKC) Fabrication and Assembly

NASA Technical Reports Server (NTRS)

Godfroy, Thomas J.

2007-01-01

An actively pumped Stainless Steel NaK Circuit (SNaKC) has been designed and fabricated by the Early Flight Fission Test Facility (EFF-TF) team at NASA's Marshall Space Flight Center. This circuit uses the eutectic mixture of sodium and potassium (NaK) as the working fluid building upon the experience and accomplishments of the SNAP reactor program from the late 1960's The SNaKC enables valuable experience and liquid metal test capability to be gained toward the goal of designing and building an affordable surface power reactor. The basic circuit components include a simulated reactor core a NaK to gas heat exchanger, an electromagnetic (EM) liquid metal pump, a liquid metal flow meter, an expansion reservoir and a drain/fill reservoir To maintain an oxygen free environment in the presence of NaK, an argon system is utilized. A helium and nitrogen system are utilized for core, pump, and heat exchanger operation. An additional rest section is available to enable special component testing m an elevated temperature actively pumped liquid metal environment. This paper summarizes the physical build of the SNaKC the gas and pressurization systems, vacuum systems, as well as instrumentation and control methods.

Process for photosynthetically splitting water

DOEpatents

Greenbaum, Elias

1984-01-01

The invention is an improved process for producing gaseous hydrogen and oxygen from water. The process is conducted in a photolytic reactor which contains a water-suspension of a photoactive material containing a hydrogen-liberating catalyst. The reactor also includes a volume for receiving gaseous hydrogen and oxygen evolved from the liquid phase. To avoid oxygen-inactivation of the catalyst, the reactor is evacuated continuously by an external pump which circulates the evolved gases through means for selectively recovering hydrogen therefrom. The pump also cools the reactor by evaporating water from the liquid phase. Preferably, product recovery is effected by selectively diffusing the hydrogen through a heated semipermeable membrane, while maintaining across the membrane a magnetic field gradient which biases the oxygen away from the heated membrane. This promotes separation, minimizes the back-reaction of hydrogen and oxygen, and protects the membrane.

POWER REACTOR

DOEpatents

Zinn, W.H.

1958-07-01

A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

Cryosorption Pumps for a Neutral Beam Injector Test Facility

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

Dremel, M.; Mack, A.; Day, C.

2006-04-27

We present the experiences of the manufacturing and the operating of a system of two identical cryosorption pumps used in a neutral beam injector test facility for fusion reactors. Calculated and measured heat loads of the cryogenic liquid helium and liquid nitrogen circuits of the cryosorption pumps are discussed. The design calculations concerning the thermo-hydraulics of the helium circuit are compared with experiences from the operation of the cryosorption pumps. Both cryopumps are integrated in a test facility of a neutral beam injector that will be used to heat the plasma of a nuclear fusion reactor with a beam ofmore » deuterium or hydrogen molecules. The huge gas throughput into the vessel of the test facility results in challenging needs on the cryopumping system.The developed cryosorption pumps are foreseen to pump a hydrogen throughput of 20 - 30 mbar{center_dot}l/s. To establish a mean pressure of several 10-5 mbar in the test vessel a pumping speed of about 350 m3/s per pump is needed. The pressure conditions must be maintained over several hours pumping without regeneration of the cryopanels, which necessitates a very high pumping capacity. A possibility to fulfill these requirements is the use of charcoal coated cryopanels to pump the gasloads by adsorption. For the cooling of the cryopanels, liquid helium at saturation pressure is used and therefore a two-phase forced flow in the cryopump system must be controlled.« less

Heat Radiators for Electromagnetic Pumps

NASA Technical Reports Server (NTRS)

Campana, R. J.

1986-01-01

Report proposes use of carbon/carbon composite radiators in electromagnetic coolant pumps of nuclear reactors on spacecraft. Carbon/carbon composite materials function well at temperatures in excess of 2,200 K. Aluminum has melting temperature of only 880 K.

Studies of PuF sub 6 and transplutonic materials' critical properties for space high power nuclear pumped lasers

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

Gu, A.G.; Miller, M.S.

1991-01-01

All space missions require a reliable, compact source of energy. This paper describes preliminary neutronics studies of pocket'' reactor concepts employing PuF{sub 6} and transplutonic materials as fuels for space high power/energy Nuclear Pumped Lasers (NPLs). Previous research has studied NPL reactor concepts with thin fuel layers, aerosol fuels and gaseous UF{sub 6}. The total reactor volumes for compact reactors with these types of fuels typically range from 3 m{sup 3} to 50 m{sup 3}. By employing PuF{sub 6} and transplutonic fuels at the same low densities, a calculated value for Keff of 1.2 has been achieved for conditions ofmore » 900 K and 5 atm, with total reactor volumes of 1.5 m{sup 3} for PuF{sub 6}, 0.51 m{sup 3} for Am-242m, 0.58 m{sup 3} for Cm-245 and 0.63 m{sup 3} for Cf-249.« less

Design and Analysis of Embedded I&C for a Fully Submerged Magnetically Suspended Impeller Pump

DOE PAGES

Melin, Alexander M.; Kisner, Roger A.

2018-04-03

Improving nuclear reactor power system designs and fuel-processing technologies for safer and more efficient operation requires the development of new component designs. In particular, many of the advanced reactor designs such as the molten salt reactors and high-temperature gas-cooled reactors have operating environments beyond the capability of most currently available commercial components. To address this gap, new cross-cutting technologies need to be developed that will enable design, fabrication, and reliable operation of new classes of reactor components. The Advanced Sensor Initiative of the Nuclear Energy Enabling Technologies initiative is investigating advanced sensor and control designs that are capable of operatingmore » in these extreme environments. Under this initiative, Oak Ridge National Laboratory (ORNL) has been developing embedded instrumentation and control (I&C) for extreme environments. To develop, test, and validate these new sensing and control techniques, ORNL is building a pump test bed that utilizes submerged magnetic bearings to levitate the shaft. The eventual goal is to apply these techniques to a high-temperature (700°C) canned rotor pump that utilizes active magnetic bearings to eliminate the need for mechanical bearings and seals. The technologies will benefit the Next Generation Power Plant, Advanced Reactor Concepts, and Small Modular Reactor programs. In this paper, we will detail the design and analysis of the embedded I&C test bed with submerged magnetic bearings, focusing on the interplay between the different major systems. Then we will analyze the forces on the shaft and their role in the magnetic bearing design. Next, we will develop the radial and thrust bearing geometries needed to meet the operational requirements of the test bed. In conclusion, we will present some initial system identification results to validate the theoretical models of the test bed dynamics.« less

Design and Analysis of Embedded I&C for a Fully Submerged Magnetically Suspended Impeller Pump

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

Melin, Alexander M.; Kisner, Roger A.

Improving nuclear reactor power system designs and fuel-processing technologies for safer and more efficient operation requires the development of new component designs. In particular, many of the advanced reactor designs such as the molten salt reactors and high-temperature gas-cooled reactors have operating environments beyond the capability of most currently available commercial components. To address this gap, new cross-cutting technologies need to be developed that will enable design, fabrication, and reliable operation of new classes of reactor components. The Advanced Sensor Initiative of the Nuclear Energy Enabling Technologies initiative is investigating advanced sensor and control designs that are capable of operatingmore » in these extreme environments. Under this initiative, Oak Ridge National Laboratory (ORNL) has been developing embedded instrumentation and control (I&C) for extreme environments. To develop, test, and validate these new sensing and control techniques, ORNL is building a pump test bed that utilizes submerged magnetic bearings to levitate the shaft. The eventual goal is to apply these techniques to a high-temperature (700°C) canned rotor pump that utilizes active magnetic bearings to eliminate the need for mechanical bearings and seals. The technologies will benefit the Next Generation Power Plant, Advanced Reactor Concepts, and Small Modular Reactor programs. In this paper, we will detail the design and analysis of the embedded I&C test bed with submerged magnetic bearings, focusing on the interplay between the different major systems. Then we will analyze the forces on the shaft and their role in the magnetic bearing design. Next, we will develop the radial and thrust bearing geometries needed to meet the operational requirements of the test bed. In conclusion, we will present some initial system identification results to validate the theoretical models of the test bed dynamics.« less

Initial conceptual design study of self-critical nuclear pumped laser systems

NASA Technical Reports Server (NTRS)

Rodgers, R. J.

1979-01-01

An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.

Hydraulic Actuator for Ganged Control Rods

NASA Technical Reports Server (NTRS)

Thompson, D. C.; Robey, R. M.

1986-01-01

Hydraulic actuator moves several nuclear-reactor control rods in unison. Electromagnetic pump pushes liquid lithium against ends of control rods, forcing them out of or into nuclear reactor. Color arrows show lithium flow for reactor startup and operation. Flow reversed for shutdown. Conceived for use aboard spacecraft, actuator principle applied to terrestrial hydraulic machinery involving motion of ganged rods.

Non-equilibrium radiation nuclear reactor

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T. (Inventor)

1978-01-01

An externally moderated thermal nuclear reactor is disclosed which is designed to provide output power in the form of electromagnetic radiation. The reactor is a gaseous fueled nuclear cavity reactor device which can operate over wide ranges of temperature and pressure, and which includes the capability of processing and recycling waste products such as long-lived transuranium actinides. The primary output of the device may be in the form of coherent radiation, so that the reactor may be utilized as a self-critical nuclear pumped laser.

Fission-powered in-core thermoacoustic sensor

DOE PAGES

Garrett, Steven L.; Smith, James A.; Smith, Robert W. M.; ...

2016-04-07

A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. Furthermore, these signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

Fission-powered in-core thermoacoustic sensor

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

Garrett, Steven L.; Smith, James A.; Smith, Robert W. M.

2016-04-04

A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

Dependence of Initial Oxygen Concentration on Ozone Yield Using Inductive Energy Storage System Pulsed Power Generator

NASA Astrophysics Data System (ADS)

Go, Tomio; Tanaka, Yasushi; Yamazaki, Nobuyuki; Mukaigawa, Seiji; Takaki, Koichi; Fujiwara, Tamiya

Dependence of initial oxygen concentration on ozone yield using streamer discharge reactor driven by an inductive energy storage system pulsed power generator is described in this paper. Fast recovery type diodes were employed as semiconductor opening switch to interrupt a circuit current within 100 ns. This rapid current change produced high-voltage short pulse between a secondary energy storage inductor. The repetitive high-voltage short pulse was applied to a 1 mm diameter center wire electrode placed in a cylindrical pulse corona reactor. The streamer discharge successfully occurred between the center wire electrode and an outer cylinder ground electrode of 2 cm inner diameter. The ozone was produced with the streamer discharge and increased with increasing pulse repetition rate. The ozone yield changed in proportion to initial oxygen concentration contained in the injected gas mixture at 800 ns forward pumping time of the current. However, the decrease of the ozone yield by decreasing oxygen concentration in the gas mixture at 180 ns forward pumping time of the current was lower than the decrease at 800 ns forward pumping time of the current. This dependence of the initial oxygen concentration on ozone yield at 180 ns forward pumping time is similar to that of dielectric barrier discharge reactor.

Conceptual study of the cryocascade for pumping, separation and recycling of ITER torus exhaust

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

Mack, A.; Perinic, D.

1994-12-31

For pumping, separation and recycling of the ITER plasma exhaust, a pumping system working reliably under ambient and operating conditions of the ITER reactor is required. The pump is exposed to a magnetic field of about 3 T and has to be resistant to radioactive irradiation of 10{sup 9} rad. In the burn and dwell phase, a gas mixture consisting of hydrogen isotopes, helium ash and impurities has to be pumped at the pressure range of 10{sup {minus}2} - 10{sup {minus}1} mbar. Within the framework of the European Fusion Technology Programme, the concept of a primary cryopump for use inmore » ITER is being prepared at KfK. The cryocascade concept is planned to include three pump stages. These pump stages, which are connected in series, consist of individual chambers that may be separated from each other by means of cold valves. In the first stage, the impurities of the reactor exhaust gas are frozen out at 20-30 K. Settling of the hydrogen isotopes H/D/T on the 5 K cryosurfaces takes place in the second stage. This stage is made up of two parallel chambers, which can be switched from the pumping to the regeneration mode or vice versa. The helium fraction is bound in the downstream 5 K adsorption stage.« less

Experimental investigation of dynamo effect in the secondary pumps of the fast breeder reactor Superphenix

NASA Astrophysics Data System (ADS)

Alemany, A.; Marty, Ph.; Plunian, F.; Soto, J.

2000-01-01

The fast breeder reactors (FBR) BN600 (Russia) and Phenix (France) have been the subject of several experimental studies aimed at the observation of dynamo action. Though no dynamo effect has been identified, the possibility was raised for the FBR Superphenix (France) which has an electric power twice that of BN600 and five times larger than Phenix. We present the results of a series of experimental investigations on the secondary pumps of Superphenix. The helical sodium flow inside one pump corresponds to a maximum magnetic Reynolds number (Rm) of 25 in the experimental conditions (low temperature). The magnetic field was recorded in the vicinity of the pumps and no dynamo action has been identified. An estimate of the critical flow rate necessary to reach dynamo action has been found, showing that the pumps are far from producing dynamo action. The magnetic energy spectrum was also recorded and analysed. It is of the form k[minus sign]11/3, suggesting the existence of a large-scale magnetic field. Following Moffatt (1978), this spectrum slope is also justified by a phenomenological approach.

Testing of Liquid Metal Components for Nuclear Surface Power Systems

NASA Technical Reports Server (NTRS)

Polzin, K. A.; Pearson, J. B.; Godfroy, T. J.; Schoenfeld, M.; Webster, K.; Briggs, M. H.; Geng, S. M.; Adkins, H. E.; Werner, J. E.

2010-01-01

The capability to perform testing at both the module/component level and in near prototypic reactor configurations using a non-nuclear test methodology allowed for evaluation of two components critical to the development of a potential nuclear fission power system for the lunar surface. A pair of 1 kW Stirling power convertors, similar to the type that would be used in a reactor system to convert heat to electricity, were integrated into a reactor simulator system to determine their performance using pumped NaK as the hot side working fluid. The performance in the pumped-NaK system met or exceed the baseline performance measurements where the converters were electrically heated. At the maximum hot-side temperature of 550 C the maximum output power was 2375 watts. A specially-designed test apparatus was fabricated and used to quantify the performance of an annular linear induction pump that is similar to the type that could be used to circulate liquid metal through the core of a space reactor system. The errors on the measurements were generally much smaller than the magnitude of the measurements, permitting accurate performance evaluation over a wide range of operating conditions. The pump produced flow rates spanning roughly 0.16 to 5.7 l/s (2.5 to 90 GPM), and delta p levels from less than 1 kPa to 90 kPa (greater than 0.145 psi to roughly 13 psi). At the nominal FSP system operating temperature of 525 C the maximum efficiency was just over 4%.

High pressure/high temperature thermogravimetric apparatus. Final report

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

Calo, J.M.; Suuberg, E.M.

1999-12-01

The purpose of this instrumentation grant was to acquire a state-of-the-art, high pressure, high temperature thermogravimetric apparatus (HP/HT TGA) system for the study of the interactions between gases and carbonaceous solids for the purpose of solving problems related to coal utilization and applications of carbon materials. The instrument that we identified for this purpose was manufactured by DMT (Deutsche Montan Technologies)--Institute of Cokemaking and Coal Chemistry of Essen, Germany. Particular features of note include: Two reactors: a standard TGA reactor, capable of 1100 C at 100 bar; and a high temperature (HT) reactor, capable of operation at 1600 C andmore » 100 bar; A steam generator capable of generating steam to 100 bar; Flow controllers and gas mixing system for up to three reaction gases, plus a separate circuit for steam, and another for purge gas; and An automated software system for data acquisition and control. The HP/TP DMT-TGA apparatus was purchased in 1996 and installed and commissioned during the summer of 1996. The apparatus was located in Room 128 of the Prince Engineering Building at Brown University. A hydrogen alarm and vent system were added for safety considerations. The system has been interfaced to an Ametek quadruple mass spectrometer (MA 100), pumped by a Varian V250 turbomolecular pump, as provided for in the original proposed. With this capability, a number of gas phase species of interest can be monitored in a near-simultaneous fashion. The MS can be used in a few different modes. During high pressure, steady-state gasification experiments, it is used to sample, measure, and monitor the reactant/product gases. It can also be used to monitor gas phase species during nonisothermal temperature programmed reaction (TPR) or temperature programmed desorption (TPD) experiments.« less

Application of nuclear pumped laser to an optical self-powered neutron detector

NASA Astrophysics Data System (ADS)

Yamanaka, N.; Takahashi, H.; Iguchi, T.; Nakazawa, M.; Kakuta, T.; Yamagishi, H.; Katagiri, M.

1996-05-01

A Nuclear Pumped Laser (NPL) using 3He/Ne/Ar gas mixture is investigated for a purpose of applying to an optical self-powered neutron detector. Reactor experiments and simulations on lasing mechanism have been made to estimate the best gas pressure and mixture ratios on the threshold input power density (or thermal neutron flux) in 3He/Ne/Ar mixture. Calculational results show that the best mixture pressure is 3He/Ne/Ar=2280/60/100 Torr and thermal neutron flux threshold 5×1012 n/cm2 sec, while the reactor experiments made in the research reactor ``YAYOI'' of the University of Tokyo and ``JRR-4'' of JAERI also demonstrate that excitational efficiency is maximized in a similar gas mixture predicted by the calculation.

Solution of heat removal from nuclear reactors by natural convection

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav

2014-03-01

This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR).The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor) for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

Assembly for facilitating inservice inspection of a reactor coolant pump rotor

DOEpatents

Veronesi, Luciano

1990-01-01

A reactor coolant pump has an outer casing with an internal cavity holding a coolant and a rotor rotatably mounted in the cavity within the coolant. An assembly for permitting inservice inspection of the pump rotor without first draining the coolant from the casing cavity is attached to an end of the pump. A cylindrical bore is defined through the casing in axial alignment with an end of pump rotor and opening into the internal cavity. An extension attached on the rotor end and rotatable therewith has a cylindrical coupler member extending into the bore. An outer end of the coupler member has an element configured to receive a tool for performance of inservice rotor inspection. A hollow cylindrical member is disposed in the bore and surrounds the coupler member. The cylindrical member is slidably movable relative to the coupler member along the bore between a retracted position wherein the cylindrical member is stored for normal pump operation and an extended position wherein the cylindrical member is extended for permitting inservice rotor inspection. A cover member is detachably and sealably attached to the casing across the bore for closing the bore and retaining the cylindrical member at its retracted position for normal pump operation. Upon detachment of the cover member, the cylindrical member can be extended to permit inservice rotor inspection.

Thermoelectric pump performance analysis computer code

NASA Technical Reports Server (NTRS)

Johnson, J. L.

1973-01-01

A computer program is presented that was used to analyze and design dual-throat electromagnetic dc conduction pumps for the 5-kwe ZrH reactor thermoelectric system. In addition to a listing of the code and corresponding identification of symbols, the bases for this analytical model are provided.

Pressure-Letdown Machine for a Coal Reactor

NASA Technical Reports Server (NTRS)

Perkins, G. S.; Mabe, W. B.

1986-01-01

Pumps operating in reverse generate power. Conceptual pressure-letdown machine for coal-liquefaction system extracts energy from expansion of product fluid. Mud pumps, originally intended for use in oil drilling, operated in reverse so their motors act as generators. Several pumps operated in alternating phase to obtain multiple stages of letdown from inlet pressure to outlet pressure. About 75 percent of work generates inlet pressure recoverable as electrical energy.

Heat pump apparatus

DOEpatents

Nelson, Paul A.; Horowitz, Jeffrey S.

1983-01-01

A heat pump apparatus including a compact arrangement of individual tubular reactors containing hydride-dehydride beds in opposite end sections, each pair of beds in each reactor being operable by sequential and coordinated treatment with a plurality of heat transfer fluids in a plurality of processing stages, and first and second valves located adjacent the reactor end sections with rotatable members having multiple ports and associated portions for separating the hydride beds at each of the end sections into groups and for simultaneously directing a plurality of heat transfer fluids to the different groups. As heat is being generated by a group of beds, others are being regenerated so that heat is continuously available for space heating. As each of the processing stages is completed for a hydride bed or group of beds, each valve member is rotated causing the heat transfer fluid for the heat processing stage to be directed to that bed or group of beds. Each of the end sections are arranged to form a closed perimeter and the valve member may be rotated repeatedly about the perimeter to provide a continuous operation. Both valves are driven by a common motor to provide a coordinated treatment of beds in the same reactors. The heat pump apparatus is particularly suitable for the utilization of thermal energy supplied by solar collectors and concentrators but may be used with any source of heat, including a source of low-grade heat.

A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

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

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2015-03-01

The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoidmore » overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety system components such as the safety relief valve (SRV), the RCIC system, the wet well, and the dry well. The results show reasonable system behaviors while exhibiting rich dynamics such as variable flow rates through RCIC turbine and pump during the SBO transient. The model has the potential to resolve the Fukushima RCIC mystery after adding the off-design two-phase turbine operation model and other additional improvements.« less

An experimental investigation of (UF-235)6 fission nuclear-pumped lasers

NASA Technical Reports Server (NTRS)

Miley, G. H.

1979-01-01

A UF6 handling system was designed for use in conjunction with the existing nuclear-pumped laser vacuum system at a nuclear reactor laboratory to perform the experiments described above. A modification to separate the gas fill system from the vacuum system and thus greatly reduce its volume is described as well as operating procedures for the first controlled nuclear pumping experiments with UF6 vapor contained in the laser cell.

Laminated electromagnetic pump stator core

DOEpatents

Fanning, A.W.

1995-08-08

A stator core for an electromagnetic pump includes a plurality of circumferentially abutting tapered laminations extending radially outwardly from a centerline axis to collectively define a radially inner bore and a radially outer circumference. Each of the laminations includes radially inner and outer edges and has a thickness increasing from the inner edge toward the outer edge to provide a substantially continuous path adjacent the circumference. This pump is used in nuclear fission reactors. 19 figs.

Hydrogen production from high moisture content biomass in supercritical water

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

Antal, M.J. Jr.; Xu, X.

1998-08-01

By mixing wood sawdust with a corn starch gel, a viscous paste can be produced that is easily delivered to a supercritical flow reactor by means of a cement pump. Mixtures of about 10 wt% wood sawdust with 3.65 wt% starch are employed in this work, which the authors estimate to cost about $0.043 per lb. Significant reductions in feed cost can be achieved by increasing the wood sawdust loading, but such an increase may require a more complex pump. When this feed is rapidly heated in a tubular flow reactor at pressures above the critical pressure of water (22more » MPa), the sawdust paste vaporizes without the formation of char. A packed bed of carbon catalyst in the reactor operating at about 650 C causes the tarry vapors to react with water, producing hydrogen, carbon dioxide, and some methane with a trace of carbon monoxide. The temperature and history of the reactor`s wall influence the hydrogen-methane product equilibrium by catalyzing the methane steam reforming reaction. The water effluent from the reactor is clean. Other biomass feedstocks, such as the waste product of biodiesel production, behave similarly. Unfortunately, sewage sludge does not evidence favorable gasification characteristics and is not a promising feedstock for supercritical water gasification.« less

Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

NASA Astrophysics Data System (ADS)

Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

Fast-acting nuclear reactor control device

DOEpatents

Kotlyar, Oleg M.; West, Phillip B.

1993-01-01

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

Submersible sodium pump

DOEpatents

Brynsvold, Glen V.; Lopez, John T.; Olich, Eugene E.; West, Calvin W.

1989-01-01

An electromagnetic submerged pump has an outer cylindrical stator with an inner cylindrical conductive core for the submerged pumping of sodium in the cylindrical interstitial volume defined between the stator and core. The cylindrical interstitial volume is typically vertically oriented, and defines an inlet at the bottom and an outlet at the top. The outer stator generates upwardly conveyed toroidal magnetic fields, which fields convey preferably from the bottom of the pump to the top of the pump liquid sodium in the cold leg of a sodium cooled nuclear reactor. The outer cylindrical stator has a vertically disposed duct surrounded by alternately stacked layers of coil units and laminates.

Submersible sodium pump

DOEpatents

Brynsvold, G.V.; Lopez, J.T.; Olich, E.E.; West, C.W.

1989-11-21

An electromagnetic submerged pump has an outer cylindrical stator with an inner cylindrical conductive core for the submerged pumping of sodium in the cylindrical interstitial volume defined between the stator and core. The cylindrical interstitial volume is typically vertically oriented, and defines an inlet at the bottom and an outlet at the top. The outer stator generates upwardly conveyed toroidal magnetic fields, which fields convey preferably from the bottom of the pump to the top of the pump liquid sodium in the cold leg of a sodium cooled nuclear reactor. The outer cylindrical stator has a vertically disposed duct surrounded by alternately stacked layers of coil units and laminates. 14 figs.

Degradation of TCE using sequential anaerobic biofilm and aerobic immobilized bed reactor

NASA Technical Reports Server (NTRS)

Chapatwala, Kirit D.; Babu, G. R. V.; Baresi, Larry; Trunzo, Richard M.

1995-01-01

Bacteria capable of degrading trichloroethylene (TCE) were isolated from contaminated wastewaters and soil sites. The aerobic cultures were identified as Pseudomonas aeruginosa (four species) and Pseudomonas fluorescens. The optimal conditions for the growth of aerobic cultures were determined. The minimal inhibitory concentration values of TCE for Pseudomonas sps. were also determined. The aerobic cells were immobilized in calcium alginate in the form of beads. Degradation of TCE by the anaerobic and dichloroethylene (DCE) by aerobic cultures was studied using dual reactors - anaerobic biofilm and aerobic immobilized bed reactor. The minimal mineral salt (MMS) medium saturated with TCE was pumped at the rate of 1 ml per hour into the anaerobic reactor. The MMS medium saturated with DCE and supplemented with xylenes and toluene (3 ppm each) was pumped at the rate of 1 ml per hour into the fluidized air-uplift-type reactor containing the immobilized aerobic cells. The concentrations of TCE and DCE and the metabolites formed during their degradation by the anaerobic and aerobic cultures were monitored by GC. The preliminary study suggests that the anaerobic and aerobic cultures of our isolates can degrade TCE and DCE.

Pressurized water nuclear reactor system with hot leg vortex mitigator

DOEpatents

Lau, Louis K. S.

1990-01-01

A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

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

Anderson, J.L.; Jenkins, E.M.; Walthers, C.R.

Compound cryopumps have been added to the Tritium Systems Test Assembly (TSTA) integrated fusion fuel loop. Operations have been performed which closely simulate an actual fusion reactor pumping scenario. In addition, performance data have been taken that support the concept of using coconut charcoal as a sorbent at 4K for pumping helium. Later tests show that coconut charcoal may be used to co-pump D,T and He mixtures on a single 4K panel. Rotary spiral pumps have been used successfully in several applications at TSTA and have acquired more than 9000 hours of maintenance-free operation. Metal bellows pumps have been usedmore » to back the spiral pumps and have been relatively trouble free in loop operations. Bellows pumps also have more than 9000 hours of maintenance-free operation. 5 refs., 6 figs.« less

The 5-kwe reactor thermoelectric system summary

NASA Technical Reports Server (NTRS)

Vanosdol, J. H. (Editor)

1973-01-01

Design of the 5-kwe reactor thermoelectric system was initiated in February 1972 and extended through the conceptual design phase into the preliminary design phase. Design effort was terminated in January, 1973. This report documents the system and component requirements, design approaches, and performance and design characteristics for the 5-kwe system. Included is summary information on the reactor, radiation shields, power conversion systems, thermoelectric pump, radiator/structure, liquid metal components, and the control system.

Flow Components in a NaK Test Loop Designed to Simulate Conditions in a Nuclear Surface Power Reactor

NASA Astrophysics Data System (ADS)

Polzin, Kurt A.; Godfroy, Thomas J.

2008-01-01

A test loop using NaK as the working fluid is presently in use to study material compatibility effects on various components that comprise a possible nuclear reactor design for use on the lunar surface. A DC electromagnetic (EM) pump has been designed and implemented as a means of actively controlling the NaK flow rate through the system and an EM flow sensor is employed to monitor the developed flow rate. These components allow for the matching of the flow rate conditions in test loops with those that would be found in a full-scale surface-power reactor. The design and operating characteristics of the EM pump and flow sensor are presented. In the EM pump, current is applied to a set of electrodes to produce a Lorentz body force in the fluid. A measurement of the induced voltage (back-EMF) in the flow sensor provides the means of monitoring flow rate. Both components are compact, employing high magnetic field strength neodymium magnets thermally coupled to a water-cooled housing. A vacuum gap limits the heat transferred from the high temperature NaK tube to the magnets and a magnetically-permeable material completes the magnetic circuit. The pump is designed to produce a pressure rise of 34.5 kPa, and the flow sensor's predicted output is roughly 20 mV at the loop's nominal flow rate of 0.114 m3/hr.

Flow Components in a NaK Test Loop Designed to Simulate Conditions in a Nuclear Surface Power Reactor

NASA Technical Reports Server (NTRS)

Polzin, Kurt A.; Godfroy, Thomas J.

2008-01-01

A test loop using NaK as the working fluid is presently in use to study material compatibility effects on various components that comprise a possible nuclear reactor design for use on the lunar surface. A DC electromagnetic (EM) pump has been designed and implemented as a means of actively controlling the NaK flow rate through the system and an EM flow sensor is employed to monitor the developed flow rate. These components allow for the matching of the flow rate conditions in test loops with those that would be found in a full-scale surface-power reactor. The design and operating characteristics of the EM pump and flow sensor are presented. In the EM pump, current is applied to a set of electrodes to produce a Lorentz body force in the fluid. A measurement of the induced voltage (back-EMF) in the flow sensor provides the means of monitoring flow rate. Both components are compact, employing high magnetic field strength neodymium magnets thermally coupled to a water-cooled housing. A vacuum gap limits the heat transferred from the high temperature NaK tube to the magnets and a magnetically-permeable material completes the magnetic circuit. The pump is designed to produce a pressure rise of 5 psi, and the flow sensor's predicted output is roughly 20 mV at the loop's nominal flow rate of 0.5 GPM.

Chemical Characterization of Simulated Boiling Water Reactor Coolant

DTIC Science & Technology

1990-05-01

33 Table 3. 1: BCCL Sample Block Design Calculations ........................................... 45 Table 5.1: Gas Absorption...cover gas . The cool, degassed pure water is pumped through a regenerative heat exchanger and then through an electric feedwater heater. The feedwater is...POINTS DWCMRHEAT DOWNOMER---EXCHANGER CHEMICAL GAHP INJECTIOIN PUMP SYSTEM COIVER GAS IN-CLIRE SECTION CAGN TANK RECOMBINER! ______ DEMINERALIZER (Cic

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

NASA Image and Video Library

1964-05-21

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

Experimental investigations of a uranium plasma pertinent to a self-sustaining plasma source

NASA Technical Reports Server (NTRS)

Schneider, R. T.

1971-01-01

The research is pertinent to the realization of a self-sustained fissioning plasma for applications such as nuclear propulsion, closed cycle MHD power generation using a plasma core reactor, and heat engines such as the nuclear piston engine, as well as the direct conversion of fission energy into optical radiation (nuclear pumped lasers). Diagnostic measurement methods and experimental devices simulating plasma core reactor conditions are discussed. Studies on the following topics are considered: (1) ballistic piston compressor (U-235); (2) high pressure uranium plasma (natural uranium); (3) sliding spark discharge (natural uranium); (4) fission fragment interaction (He-3 and U-235); and (5) nuclear pumped lasers (He-3 and U-235).

Tower reactors for bioconversion of lignocellulosic material

DOEpatents

Nguyen, Quang A.

1999-01-01

An apparatus for enzymatic hydrolysis and fermentation of pretreated lignocellulosic material, in the form of a tower bioreactor, having mixers to achieve intermittent mixing of the material. Precise mixing of the material is important for effective heat and mass transfer requirements without damaging or denaturing the enzymes or fermenting microorganisms. The pretreated material, generally in the form of a slurry, is pumped through the bioreactor, either upwards or downwards, and is mixed periodically as it passes through the mixing zones where the mixers are located. For a thin slurry, alternate mixing can be achieved by a pumping loop which also serves as a heat transfer device. Additional heat transfer takes place through the reactor heat transfer jackets.

Tower reactors for bioconversion of lignocellulosic material

DOEpatents

Nguyen, Quang A.

1998-01-01

An apparatus for enzymatic hydrolysis and fermentation of pretreated lignocellulosic material, in the form of a tower bioreactor, having mixers to achieve intermittent mixing of the material. Precise mixing of the material is important for effective heat and mass transfer requirements without damaging or denaturing the enzymes or fermenting microorganisms. The pretreated material, generally in the form of a slurry, is pumped through the bioreactor, either upwards of downwards, and is mixed periodically as it passes through the mixing zones where the mixers are located. For a thin slurry, alternate mixing can be achieved by a pumping loop which also serves as a heat transfer device. Additional heat transfer takes place through the reactor heat transfer jackets.

Multiphysics Modeling of an Annular Linear Induction Pump With Applications to Space Nuclear Power Systems

NASA Technical Reports Server (NTRS)

Kilbane, J.; Polzin, K. A.

2014-01-01

An annular linear induction pump (ALIP) that could be used for circulating liquid-metal coolant in a fission surface power reactor system is modeled in the present work using the computational COMSOL Multiphysics package. The pump is modeled using a two-dimensional, axisymmetric geometry and solved under conditions similar to those used during experimental pump testing. Real, nonlinear, temperature-dependent material properties can be incorporated into the model for both the electrically-conducting working fluid in the pump (NaK-78) and structural components of the pump. The intricate three-phase coil configuration of the pump is implemented in the model to produce an axially-traveling magnetic wave that is qualitatively similar to the measured magnetic wave. The model qualitatively captures the expected feature of a peak in efficiency as a function of flow rate.

DIRECT CURRENT ELECTROMAGNETIC PUMP

DOEpatents

Barnes, A.H.

1957-11-01

An improved d-c electromagnetic pump is presented in which the poles, and consequently the magetic gap at the poles, are tapered to be wider at the upstream end. In addition, the cross section of the tube carryiQ the liquid metal is tapered so that the velocity of the pumped liquid increases in the downstream direction at a rate such that the counter-induced voltage in the liquid metal remains constant as it traverses the region between the poles. This configuration compensates for the distortion of the magnetic field caused by the induced voltage that would otherwise result in the lowering of the pumping capacity. This improved electromagnetic pump as practical application in the pumping of liquid metal coolants for nuclear reactors where conventional positive displacement pumps have proved unsatisfactory due to the high temperatures and the corrosive properties of the liquid metals involved.

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

Logan, B.G.

A recently completed two-year study of a commercial tandem mirror reactor design (Mirror Advanced Reactor Study (MARS)) is briefly reviewed. The end plugs are designed for trapped particle stability, MHD ballooning, balanced geodesic curvature, and small radial electric fields in the central cell. New technologies such as lithium-lead blankets, 24T hybrid coils, gridless direct converters and plasma halo vacuum pumps are highlighted.

Development of coal-feeding systems at the Morgantown Energy Research Center

NASA Technical Reports Server (NTRS)

Hobday, J. M.

1977-01-01

Systems for feeding crushed and pulverized coal into coal conversion reactor vessels are described. Pneumatic methods for feeding pulverized coal, slurry feeders, and coal pumps, methods for steam pickup, and a method for drying a water-coal slurry in a steam fluidized bed subsequent to feeding the coal into a reactor vessel are included.

Shock and vibration tests of a SNAP-8 NaK pump

NASA Technical Reports Server (NTRS)

Stromquist, A. J.; Nelson, R. B.; Hibben, L.

1971-01-01

The pump used for reactor cooling in the SNAP 8 space power system was subjected to the expected vehicle launch vibration, and shock loading in accordance with the SNAP 8 environmental specification. Subsequent disassembly revealed damage to the thrust bearing pins, which should be redesigned and strengthened. The unit was operational, however, when run in a test loop after reassembly.

The WSTIAC Quarterly. Volume 9, Number 3

DTIC Science & Technology

2010-01-25

program .[8] THE THORIUM FUEL CYCLE AND LFTR POWER PLANT The thorium fuel cycle is based on a series of neutron absorp- tion and beta decay processes...the fig- ure is a graphite matrix moderated MSR reactor with fuel salt mixture (ThF4-U233F4) being circulated by a pump through the core and to a...the core as purified salt. As one of the unique safety features, a melt-plug at the reactor bottom would permit the reactor fluid fuel to be drained

System for detecting and limiting electrical ground faults within electrical devices

DOEpatents

Gaubatz, Donald C.

1990-01-01

An electrical ground fault detection and limitation system for employment with a nuclear reactor utilizing a liquid metal coolant. Elongate electromagnetic pumps submerged within the liquid metal coolant and electrical support equipment experiencing an insulation breakdown occasion the development of electrical ground fault current. Without some form of detection and control, these currents may build to damaging power levels to expose the pump drive components to liquid metal coolant such as sodium with resultant undesirable secondary effects. Such electrical ground fault currents are detected and controlled through the employment of an isolated power input to the pumps and with the use of a ground fault control conductor providing a direct return path from the affected components to the power source. By incorporating a resistance arrangement with the ground fault control conductor, the amount of fault current permitted to flow may be regulated to the extent that the reactor may remain in operation until maintenance may be performed, notwithstanding the existence of the fault. Monitors such as synchronous demodulators may be employed to identify and evaluate fault currents for each phase of a polyphase power, and control input to the submerged pump and associated support equipment.

Enhanced Component Performance Study: Motor-Driven Pumps 1998–2014

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

Schroeder, John Alton

2016-02-01

This report presents an enhanced performance evaluation of motor-driven pumps at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience failure reports from fiscal year 1998 through 2014 for the component reliability as reported in the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The motor-driven pump failure modes considered for standby systems are failure to start, failure to run less than or equal to one hour, and failure to run more than one hour; for normally running systems, the failure modes considered are failure to start and failure tomore » run. An eight hour unreliability estimate is also calculated and trended. The component reliability estimates and the reliability data are trended for the most recent 10-year period while yearly estimates for reliability are provided for the entire active period. Statistically significant increasing trends were identified in pump run hours per reactor year. Statistically significant decreasing trends were identified for standby systems industry-wide frequency of start demands, and run hours per reactor year for runs of less than or equal to one hour.« less

Integrated electrochemical-biological process as an alternative mean for ammonia monitoring during anaerobic digestion of organic wastes.

PubMed

Zhao, Nannan; Li, Xiaohu; Jin, Xiangdan; Angelidaki, Irini; Zhang, Yifeng

2018-03-01

Ammonia monitoring is important to control anaerobic digestion (AD) process due to inhibition effect. Here, an electrolysis cell (EC) was integrated with a complete nitrification reactor as an alternative approach for online monitoring of ammonia during AD processes. The AD effluent was pumped into nitrification reactor to convert ammonia to nitrate, followed by the introduction of nitrate-rich effluent to EC cathode. It was first evaluated with synthetic ammonia-rich digesters and was observed that the current at 5 min were linearly corresponding to the ammonia levels (from 0 to 7.5 mM NH 4 + -N, R 2  = 0.9673). The linear relationship was always observed regardless of different wastewater pH and external voltage. Pre-removal of other electron acceptors from digestate at cathode could eliminate their disturbances to sensor performance. Finally, the accuracy of biosensor was verified with real digestate test. The simple and reliable biosensor showed great promising for online ammonia monitoring of AD processes. Copyright © 2017 Elsevier Ltd. All rights reserved.

Simulation of sodium pumps for nuclear power plants. Technical report 1 Oct 80-1 May 81

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

Boadu, H.O.

1981-05-01

A single-phase pump model for analysis of transients in sodium cooled fast breeder nuclear power plants has been presented, where homologous characteristic curves are used to predict the behavior of the pump during operating transients. The pump model has been incorporated into BRENDA and FFTF; two system cases to simulate Clinch River Breeder Reactor Plant (CRBRP) and the Fast Flux Test Facility (FFTF) respectively. Two simulation test results for BRENDA which is one loop representation of a three loop plant have been presented. They are: (1) Primary pump coastdown to natural circulation coupled with scram failure, and (2) 10 percentmore » deviation of primary speed with plant controllers incorporated.« less

Compact, Lightweight Electromagnetic Pump for Liquid Metal

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Palzin, Kurt

2010-01-01

A proposed direct-current electromagnetic pump for circulating a molten alkali metal alloy would be smaller and lighter and would demand less input power, relative to currently available pumps of this type. (Molten alkali metals are used as heat-transfer fluids in high-temperature stages of some nuclear reactors.) The principle of operation of this or any such pump involves exploitation of the electrical conductivity of the molten metal: An electric current is made to pass through the liquid metal along an axis perpendicular to the longitudinal axis of the flow channel, and a magnetic field perpendicular to both the longitudinal axis and the electric current is superimposed on the flowchannel region containing the electric current. The interaction between the electric current and the magnetic field produces the pumping force along the longitudinal axis. The advantages of the proposed pump over other such pumps would accrue from design features that address overlapping thermal and magnetic issues.

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

A compact spin-exchange optical pumping system for 3He polarization based on a solenoid coil, a VBG laser diode, and a cosine theta RF coil

NASA Astrophysics Data System (ADS)

Lee, Sungman; Kim, Jongyul; Moon, Myung Kook; Lee, Kye Hong; Lee, Seung Wook; Ino, Takashi; Skoy, Vadim R.; Lee, Manwoo; Kim, Guinyun

2013-02-01

For use as a neutron spin polarizer or analyzer in the neutron beam lines of the HANARO (High-flux Advanced Neutron Application ReactOr) nuclear research reactor, a 3He polarizer was designed based on both a compact solenoid coil and a VBG (volume Bragg grating) diode laser with a narrow spectral linewidth of 25 GHz. The nuclear magnetic resonance (NMR) signal was measured and analyzed using both a built-in cosine radio-frequency (RF) coil and a pick-up coil. Using a neutron transmission measurement, we estimated the polarization ratio of the 3He cell as 18% for an optical pumping time of 8 hours.

Tower reactors for bioconversion of lignocellulosic material

DOEpatents

Nguyen, Q.A.

1998-03-31

An apparatus is disclosed for enzymatic hydrolysis and fermentation of pretreated lignocellulosic material. The apparatus consists of a tower bioreactor which has mixers to achieve intermittent mixing of the material. Precise mixing of the material is important for effective heat and mass transfer requirements without damaging or denaturing the enzymes or fermenting microorganisms. The pretreated material, generally in the form of a slurry, is pumped through the bioreactor, either upwards or downwards, and is mixed periodically as it passes through the mixing zones where the mixers are located. For a thin slurry, alternate mixing can be achieved by a pumping loop which also serves as a heat transfer device. Additional heat transfer takes place through the reactor heat transfer jackets. 5 figs.

Tower reactors for bioconversion of lignocellulosic material

DOEpatents

Nguyen, Q.A.

1999-03-30

An apparatus is described for enzymatic hydrolysis and fermentation of pretreated lignocellulosic material, in the form of a tower bioreactor, having mixers to achieve intermittent mixing of the material. Precise mixing of the material is important for effective heat and mass transfer requirements without damaging or denaturing the enzymes or fermenting microorganisms. The pretreated material, generally in the form of a slurry, is pumped through the bioreactor, either upwards or downwards, and is mixed periodically as it passes through the mixing zones where the mixers are located. For a thin slurry, alternate mixing can be achieved by a pumping loop which also serves as a heat transfer device. Additional heat transfer takes place through the reactor heat transfer jackets. 5 figs.

Experimental investigation of a new method for advanced fast reactor shutdown cooling

NASA Astrophysics Data System (ADS)

Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

2017-07-01

We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

71. ARAII. Construction progress at SL1 site near end of ...

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

71. ARA-II. Construction progress at SL-1 site near end of 1957. Buildings from right to left are guard house, support building, reactor building, water tank and pump house. Construction was 23 percent complete. December 20, 1957. Ineel photo no. 57-6224. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Self-pumping impurity control

DOEpatents

Brooks, Jeffrey N.; Mattas, Richard F.

1991-01-01

Apparatus for removing the helium ash from a fusion reactor having a D-T plasma comprises a helium trapping site within the reactor plasma confinement device, said trapping site being formed of a trapping material having negligible helium solubility and relatively high hydrogen solubility; and means for depositing said trapping material on said site at a rate sufficient to prevent saturation of helium trapping.

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

A plasma arc reactor for fullerene research

NASA Astrophysics Data System (ADS)

Anderson, T. T.; Dyer, P. L.; Dykes, J. W.; Klavins, P.; Anderson, P. E.; Liu, J. Z.; Shelton, R. N.

1994-12-01

A modified Krätschmer-Huffman reactor for the mass production of fullerenes is presented. Fullerene mass production is fundamental for the synthesis of higher and endohedral fullerenes. The reactor employs mechanisms for continuous graphite-rod feeding and in situ slag removal. Soot collects into a Soxhlet extraction thimble which serves as a fore-line vacuum pump filter, thereby easing fullerene separation from soot. Thermal gravimetric analysis (TGA) for yield determination is reported. This TGA method is faster and uses smaller samples than Soxhlet extraction methods which rely on aromatic solvents. Production of 10 g of soot per hour is readily achieved utilizing this reactor. Fullerene yields of 20% are attained routinely.

Comprehensive work plan for Building 3001 storage canal at the Oak Ridge National Laboratory, Oak Ridge, Tennessee

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

NONE

1997-01-01

This Comprehensive Work Plan describes the method of accomplishment to replace the shielding protection of the water in the canal with a controlled low strength material (CLSM) 4. The canal was used during the operation of the Oak Ridge Graphite Reactor in the 1940s and 1950s to transport spent fuel slugs and irradiated test materials from the reactor, under water to the hot cell in Building 3019 for further processing, packaging, and handling. After the reactor was shut down, the canal was used until 1990 to store some irradiated materials until they could be transferred to a Solid Waste Storagemore » Area. This task has the following objectives and components: (1) minimize potential future risk to human health and the environment; (2) reduce surveillance and maintenance cost of the canal; (3) perform site preparation activities; (4) replace the water in the canal with a solid CLSM; (5) pump the water to the Process Waste Treatment System (PWTS) for further processing at the same rate that the CLSM is pumped under the water; (6) remove the water using a process that will protect the workers and the public in the visitors area from contamination while the CLSM is being pumped underneath the water; (7) painting a protective coating material over the CLSM after the CLSM has cured.« less

Space power reactor in-core thermionic multicell evolutionary (S-prime) design

NASA Astrophysics Data System (ADS)

Determan, William R.; Van Hagan, Tom H.

1993-01-01

A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m2 heat pipe space radiator.

REACTOR

DOEpatents

Spitzer, L. Jr.

1962-01-01

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

Method and system to directly produce electrical power within the lithium blanket region of a magnetically confined, deuterium-tritium (DT) fueled, thermonuclear fusion reactor

DOEpatents

Woolley, Robert D.

1999-01-01

A method for integrating liquid metal magnetohydrodynamic power generation with fusion blanket technology to produce electrical power from a thermonuclear fusion reactor located within a confining magnetic field and within a toroidal structure. A hot liquid metal flows from a liquid metal blanket region into a pump duct of an electromagnetic pump which moves the liquid metal to a mixer where a gas of predetermined pressure is mixed with the pressurized liquid metal to form a Froth mixture. Electrical power is generated by flowing the Froth mixture between electrodes in a generator duct. When the Froth mixture exits the generator the gas is separated from the liquid metal and both are recycled.

122. ARAI Pump House (ARA629). Drawing shows north, south, east, ...

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

122. ARA-I Pump House (ARA-629). Drawing shows north, south, east, and west elevations, floor plan, foundation plan, and other details. Note small enclosure at southwest corner of building to contain chlorination equipment. Norman Engineering Company 961-area/SF-629-A-1. Date: January 1959. Ineel index code no. 068-0629-00-613-102774. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Capabilities and Testing of the Fission Surface Power Primary Test Circuit (FSP-PTC)

NASA Technical Reports Server (NTRS)

Garber, Anne E.

2007-01-01

An actively pumped alkali metal flow circuit, designed and fabricated at the NASA Marshall Space Flight Center, is currently undergoing testing in the Early Flight Fission Test Facility (EFF-TF). Sodium potassium (NaK), which was used in the SNAP-10A fission reactor, was selected as the primary coolant. Basic circuit components include: simulated reactor core, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, liquid metal flowmeter, load/drain reservoir, expansion reservoir, test section, and instrumentation. Operation of the circuit is based around a 37-pin partial-array core (pin and flow path dimensions are the same as those in a full core), designed to operate at 33 kWt. NaK flow rates of greater than 1 kg/sec may be achieved, depending upon the power applied to the EM pump. The heat exchanger provides for the removal of thermal energy from the circuit, simulating the presence of an energy conversion system. The presence of the test section increases the versatility of the circuit. A second liquid metal pump, an energy conversion system, and highly instrumented thermal simulators are all being considered for inclusion within the test section. This paper summarizes the capabilities and ongoing testing of the Fission Surface Power Primary Test Circuit (FSP-PTC).

Effect of the self-pumped limiter concept on the tritium fuel cycle

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

Finn, P.A.; Sze, D.K.; Hassanein, A.

1988-01-01

The self-pumped limiter concept for impurity control of the plasma of a fusion reactor has a major impact on the design of the tritium systems. To achieve a sustained burn, conventional limiters and divertors remove large quantities of unburnt tritium and deuterium from the plasma which must be then recycled using a plasma processing system. The self-pumped limiter which does not remove the hydrogen species, does not require any plasma processing equipment. The blanket system and the coolant processing systems acquire greater importance with the use of this unconventional impurity control system. 3 refs., 2 figs.

Surveillance application using patten recognition software at the EBR-II Reactor Facility

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

Olson, D.L.

1992-01-01

The System State Analyzer (SSA) is a software based pattern recognition system. For the past several year this system has been used at Argonne National Laboratory's Experimental Breeder Reactor 2 (EBR-2) reactor for detection of degradation and other abnormalities in plant systems. Currently there are two versions of the SSA being used at EBR-2. One version of SSA is used for daily surveillance and trending of the reactor delta-T and startups of the reactor. Another version of the SSA is the QSSA which is used to monitor individual systems of the reactor such as the Secondary Sodium System, Secondary Sodiummore » Pumps, and Steam Generator. This system has been able to detect problems such as signals being affected by temperature variations due to a failing temperature controller.« less

Surveillance application using patten recognition software at the EBR-II Reactor Facility

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

Olson, D.L.

1992-05-01

The System State Analyzer (SSA) is a software based pattern recognition system. For the past several year this system has been used at Argonne National Laboratory`s Experimental Breeder Reactor 2 (EBR-2) reactor for detection of degradation and other abnormalities in plant systems. Currently there are two versions of the SSA being used at EBR-2. One version of SSA is used for daily surveillance and trending of the reactor delta-T and startups of the reactor. Another version of the SSA is the QSSA which is used to monitor individual systems of the reactor such as the Secondary Sodium System, Secondary Sodiummore » Pumps, and Steam Generator. This system has been able to detect problems such as signals being affected by temperature variations due to a failing temperature controller.« less

77 FR 60039 - Non-Power Reactor License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-02

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [NRC-2011-0087] RIN 3150-AI96 Non-Power Reactor... the final regulatory basis for rulemaking to streamline non-power reactor license renewal. This final... Reactor (RTR) License Renewal Process. This contemplated rulemaking also recommends conforming changes to...

Passive shut-down heat removal system

DOEpatents

Hundal, Rolv; Sharbaugh, John E.

1988-01-01

An improved shut-down heat removal system for a liquid metal nuclear reactor of the type having a vessel for holding hot and cold pools of liquid sodium is disclosed herein. Generally, the improved system comprises a redan or barrier within the reactor vessel which allows an auxiliary heat exchanger to become immersed in liquid sodium from the hot pool whenever the reactor pump fails to generate a metal-circulating pressure differential between the hot and cold pools of sodium. This redan also defines an alternative circulation path between the hot and cold pools of sodium in order to equilibrate the distribution of the decay heat from the reactor core. The invention may take the form of a redan or barrier that circumscribes the inner wall of the reactor vessel, thereby defining an annular space therebetween. In this embodiment, the bottom of the annular space communicates with the cold pool of sodium, and the auxiliary heat exchanger is placed in this annular space just above the drawn-down level that the liquid sodium assumes during normal operating conditions. Alternatively, the redan of the invention may include a pair of vertically oriented, concentrically disposed standpipes having a piston member disposed between them that operates somewhat like a pressure-sensitive valve. In both embodiments, the cessation of the pressure differential that is normally created by the reactor pump causes the auxiliary heat exchanger to be immersed in liquid sodium from the hot pool. Additionally, the redan in both embodiments forms a circulation flow path between the hot and cold pools so that the decay heat from the nuclear core is uniformly distributed within the vessel.

Nuclear Thermal Propulsion: Past, Present, and a Look Ahead

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.

2014-01-01

NTR: High thrust high specific impulse (2 x LOXLH2 chemical) engine uses high power density fission reactor with enriched uranium fuel as thermal power source. Reactor heat is removed using H2 propellant which is then exhausted to produce thrust. Conventional chemical engine LH2 tanks, turbo pumps, regenerative nozzles and radiation-cooled shirt extensions used -- NTR is next evolutionary step in high performance liquid rocket engines.

PRESSURE SYSTEM CONTROL

DOEpatents

Esselman, W.H.; Kaplan, G.M.

1961-06-20

The control of pressure in pressurized liquid systems, especially a pressurized liquid reactor system, may be achieved by providing a bias circuit or loop across a closed loop having a flow restriction means in the form of an orifice, a storage tank, and a pump connected in series. The subject invention is advantageously utilized where control of a reactor can be achieved by response to the temperature and pressure of the primary cooling system.

Molten salt destruction of energetic waste materials

DOEpatents

Brummond, W.A.; Upadhye, R.S.; Pruneda, C.O.

1995-07-18

A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor. 4 figs.

Molten salt destruction of energetic waste materials

DOEpatents

Brummond, William A.; Upadhye, Ravindra S.; Pruneda, Cesar O.

1995-01-01

A molten salt destruction process is used to treat and destroy energetic waste materials such as high explosives, propellants, and rocket fuels. The energetic material is pre-blended with a solid or fluid diluent in safe proportions to form a fluid fuel mixture. The fuel mixture is rapidly introduced into a high temperature molten salt bath. A stream of molten salt is removed from the vessel and may be recycled as diluent. Additionally, the molten salt stream may be pumped from the reactor, circulated outside the reactor for further processing, and delivered back into the reactor or cooled and circulated to the feed delivery system to further dilute the fuel mixture entering the reactor.

Technical note on results of production test IP-628-AI: Crosstie flow test and flushing calibration, and the earlier production tests IP-573-AI and IP-499-AI

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

Jones, S.S.

1964-06-15

A series of three experimental tests was conducted on the operation and adequacy of the K-Reactors` third, or last-ditch, cooling system. The first test showed considerable line corrosion the second test was performed directly after line cleaning, and the third test showed a significant amount of additional line corrosion after only nine months of service. The present cooling adequacy of this last-ditch system at the KE and KW reactors is summarized these show the power levels for which we have adequate last-ditch cooling as a function of the crosstie coolant temperature. These figures include the effects of increasing the numbermore » of pumps that remain in operation at the other K-Reactor, and various other operating or emergency conditions. These curves are for diesel pump speeds of 750 rpm which are planned for June of this year. In these figures the crosstie temperature is assumed to be a conservative 5{degrees}C above the process water inlet temperature.« less

Research on fission fragment excitation of gases and nuclear pumping of lasers

NASA Technical Reports Server (NTRS)

Schneider, R. T.; Davie, R. N.; Davis, J. F.; Fuller, J. L.; Paternoster, R. R.; Shipman, G. R.; Sterritt, D. E.; Helmick, H. H.

1974-01-01

Experimental investigations of fission fragment excited gases are reported along with a theoretical analysis of population inversions in fission fragment excited helium. Other studies reported include: nuclear augmentation of gas lasers, direct nuclear pumping of a helium-xenon laser, measurements of a repetitively pulsed high-power CO2 laser, thermodynamic properties of UF6 and UF6/He mixtures, and nuclear waste disposal utilizing a gaseous core reactor.

Browns Ferry-1 single-loop operation tests

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

March-Leuba, J.; Wood, R.T.; Otaduy, P.J.

1985-09-01

This report documents the results of the stability tests performed on February 9, 1985, at the Browns Ferry Nuclear Power Plant Unit 1 under single-loop operating conditions. The observed increase in neutron noise during single-loop operation is solely due to an increase in flow noise. The Browns Ferry-1 reactor has been found to be stable in all modes of operation attained during the present tests. The most unstable test plateau corresponded to minimum recirculation pump speed in single-loop operation (test BFTP3). This operating condition had the minimum flow and maximum power-to-flow ratio. The estimated decay ratio in this plateau ismore » 0.53. The decay ratio decreased as the flow was increased during single-loop operation (down to 0.34 for test plateau BFTP6). This observation implies that the core-wide reactor stability follows the same trends in single-loop as it does in two-loop operation. Finally, no local or higher mode instabilities were found in the data taken from local power range monitors. The decay ratios estimated from the local power range monitors were not significantly different from those estimated from the average power range monitors.« less

Reactor application of an improved bundle divertor

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

Yang, T.F.; Ruck, G.W.; Lee, A.Y.

1978-11-01

A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supportedmore » by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW.« less

Pressure Loss Predictions of the Reactor Simulator Subsystem at NASA GRC

NASA Technical Reports Server (NTRS)

Reid, Terry V.

2015-01-01

Testing of the Fission Power System (FPS) Technology Demonstration Unit (TDU) is being conducted at NASA GRC. The TDU consists of three subsystems: the Reactor Simulator (RxSim), the Stirling Power Conversion Unit (PCU), and the Heat Exchanger Manifold (HXM). An Annular Linear Induction Pump (ALIP) is used to drive the working fluid. A preliminary version of the TDU system (which excludes the PCU for now), is referred to as the RxSim subsystem and was used to conduct flow tests in Vacuum Facility 6 (VF 6). In parallel, a computational model of the RxSim subsystem was created based on the CAD model and was used to predict loop pressure losses over a range of mass flows. This was done to assess the ability of the pump to meet the design intent mass flow demand. Measured data indicates that the pump can produce 2.333 kg/sec of flow, which is enough to supply the RxSim subsystem with a nominal flow of 1.75 kg/sec. Computational predictions indicated that the pump could provide 2.157 kg/sec (using the Spalart-Allmaras turbulence model), and 2.223 kg/sec (using the k-? turbulence model). The computational error of the predictions for the available mass flow is -0.176 kg/sec (with the S-A turbulence model) and -0.110 kg/sec (with the k-epsilon turbulence model) when compared to measured data.

Reactor Simulator Testing

NASA Technical Reports Server (NTRS)

Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

2013-01-01

As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

Reactor Simulator Integration and Testing

NASA Technical Reports Server (NTRS)

Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

2013-01-01

As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

Partially ionized plasmas including the third symposium on uranium plasmas

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

Krishnan, M.

1976-09-01

Separate abstracts are included for 28 papers on electrically generated plasmas, fission generated plasmas, nuclear pumped lasers, gaseous fuel reactor research, and applications. Five papers have been previously abstracted and included in ERA.

Heterogeneous Reaction gaseous chlorine nitrate and solid sodium chloride

NASA Technical Reports Server (NTRS)

Timonen, Raimo S.; Chu, Liang T.; Leu, Ming-Taun

1994-01-01

The heterogeneous reaction of gaseous chlorine nitrate and solid sodium chloride was investigated over a temperature range of 220 - 300 K in a flow-tube reactor interfaced with a differentially pumped quadrupole mass spectrometer.

Controlled-Turbulence Bioreactors

NASA Technical Reports Server (NTRS)

Wolf, David A.; Schwartz, Ray; Trinh, Tinh

1989-01-01

Two versions of bioreactor vessel provide steady supplies of oxygen and nutrients with little turbulence. Suspends cells in environment needed for sustenance and growth, while inflicting less damage from agitation and bubbling than do propeller-stirred reactors. Gentle environments in new reactors well suited to delicate mammalian cells. One reactor kept human kidney cells alive for as long as 11 days. Cells grow on carrier beads suspended in liquid culture medium that fills cylindrical housing. Rotating vanes - inside vessel but outside filter - gently circulates nutrient medium. Vessel stationary; magnetic clutch drives filter cylinder and vanes. Another reactor creates even less turbulence. Oxygen-permeable tubing wrapped around rod extending along central axis. Small external pump feeds oxygen to tubing through rotary coupling, and oxygen diffuses into liquid medium.

Inherently Safe Fission Power System for Lunar Outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.; El-Genk, Mohamed S.

2013-09-01

This paper presents the Solid Core-Sectored Compact Reactor (SC-SCoRe) and power system for future lunar outposts. The power system nominally provides 38 kWe continuously for 21 years, employs static components and has no single point failures in reactor cooling or power generation. The reactor core has six sectors, each has a separate pair of primary and secondary loops with liquid NaK-56 working fluid, thermoelectric (TE) power conversion and heat-pipes radiator panels. The electromagnetic (EM) pumps in the primary and secondary loops, powered with separate TE power units, ensure operation reliability and passive decay heat removal from the reactor after shutdown. The reactor poses no radiological concerns during launch, and remains sufficiently subcritical, with the radial reflector dissembled, when submerged in wet sand and the core flooded with seawater, following a launch abort accident. After 300 years of storage below grade on the Moon, the total radioactivity in the post-operation reactor drops below 164 Ci, a low enough radioactivity for a recovery and safe handling of the reactor.

Reactor-pumped laser facility at DOE's Nevada Test Site

NASA Astrophysics Data System (ADS)

Lipinski, Ronald J.

1994-05-01

The Nevada Test Site (NTS) is one excellent possibility for a laser power beaming site. It is in the low latitudes of the U.S., is in an exceptionally cloud-free area of the southwest, is already an area of restricted access (which enhances safety considerations), and possesses a highly skilled technical team with extensive engineering and research capabilities from underground testing of our nation's nuclear deterrence. The average availability of cloud-free clear line of site to a given point in space is about 84%. With a beaming angle of +/- 60 degree(s) from the zenith, about 52 geostationary-orbit (GEO) satellites could be accessed continuously from NTS. In addition, the site would provide an average view factor of about 10% for orbital transfer from low earth orbit to GEO. One of the major candidates for a long-duration, high- power laser is a reactor-pumped laser being developed by DOE. The extensive nuclear expertise at NTS makes this site a prime candidate for utilizing the capabilities of a rector pumped laser for power beaming. The site then could be used for many dual-use roles such as industrial material processing research, defense testing, and removing space debris.

Development of flow systems by direct-milling on poly(methyl methacrylate) substrates using UV-photopolymerization as sealing process.

PubMed

Rodrigues, Eunice R G O; Lapa, Rui A S

2009-03-01

An alternative process for the design and construction of fluidic devices is presented. Several sealing processes were studied, as well as the hydrodynamic characteristics of the proposed fluidic devices. Manifolds were imprinted on polymeric substrates by direct-write milling, according to Computer Assisted Design (CAD) data. Poly(methyl methacrylate) (PMMA) was used as substrate due to its physical and chemical properties. Different bonding approaches for the imprinted channels were evaluated and UV-photopolymerization of acrylic acid (AA) was selected. The hydrodynamic characteristics of the proposed flow devices were assessed and compared to those obtained in similar flow systems using PTFE reactors and micro-pumps as propulsion units (multi-pumping approach). The applicability of the imprinted reactors was evaluated in the sequential determination of calcium and magnesium in water samples. Results obtained were in good agreement with those obtained by the reference procedure.

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

Steiner, J.L.; Lime, J.F.; Elson, J.S.

One dimensional TRAC transient calculations of the process inherent ultimate safety (PIUS) advanced reactor design were performed for a pump-trip SCRAM. The TRAC calculations showed that the reactor power response and shutdown were in qualitative agreement with the one-dimensional analyses presented in the PIUS Preliminary Safety Information Document (PSID) submitted by Asea Brown Boveri (ABB) to the US Nuclear Regulatory Commission for preapplication safety review. The PSID analyses were performed with the ABB-developed RIGEL code. The TRAC-calculated phenomena and trends were also similar to those calculated with another one-dimensional PIUS model, the Brookhaven National Laboratory developed PIPA code. A TRACmore » pump-trip SCRAM transient has also been calculated with a TRAC model containing a multi-dimensional representation of the PIUS intemal flow structures and core region. The results obtained using the TRAC fully one-dimensional PIUS model are compared to the RIGEL, PIPA, and TRAC multi-dimensional results.« less

TEM Pump With External Heat Source And Sink

NASA Technical Reports Server (NTRS)

Nesmith, Bill J.

1991-01-01

Proposed thermoelectric/electromagnetic (TEM) pump driven by external source of heat and by two or more heat pipe radiator heat sink(s). Thermoelectrics generate electrical current to circulate liquid metal in secondary loop of two-fluid-loop system. Intended for use with space and terrestrial dual loop liquid metal nuclear reactors. Applications include spacecraft on long missions or terrestrial beacons or scientific instruments having to operate in remote areas for long times. Design modified to include multiple radiators, converters, and ducts, as dictated by particular application.

Ion cyclotron production by a four-wave interaction with a helicon pump.

PubMed

Sutherland, O; Giles, M; Boswell, R

2005-05-27

Ion cyclotron waves at approximately 0.7 the ion gyrofrequency have been observed experimentally in the large volume helicon reactor WOMBAT. These waves are highly localized along the axis of the device where a 8 cm diameter, 2 m long. Ar II plasma column is produced. Spectral measurements reveal a four-wave interaction where energy is down-converted to the ion cyclotron mode from the helicon pump. The experimental results are explained in terms of a filamentation type instability.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Experimental investigation on charcoal adsorption for cryogenic pump application

NASA Astrophysics Data System (ADS)

Scannapiego, Matthieu; Day, Christian

2017-12-01

Fusion reactors are generating energy by nuclear fusion between deuterium and tritium. In order to evacuate the high gas throughputs from the plasma exhaust, large pumping speed systems are required. Within the European Fusion Programme, the Karlsruhe Institute of Technology (KIT) has taken the lead to design a three-stage cryogenic pump that can provide a separation function of hydrogen isotopes from the remaining gases; hence limiting the tritium inventory in the machine. A primary input parameter for the detailed design of a cryopump is the sticking coefficient between the gas and the pumping surface. For this purpose, the so-called TIMO open panel pump experiment was conducted in the TIMO-2 test facility at KIT in order to measure pumping speeds on an activated carbon surface cooled at temperatures between 6 K and 22 K, for various pure gases and gas mixtures, under fusion relevant gas flow conditions, and for two different geometrical pump configurations. The influences of the panel temperature, the gas throughput and the intake gas temperature on the pumping speed have been characterized, providing valuable qualitative results for the design of the three-stage cryopump. In a future work, supporting Monte Carlo simulations should allow for derivation of the sticking coefficients.

Documentation of Stainless Steel Lithium Circuit Test Section Design. Suppl

NASA Technical Reports Server (NTRS)

Godfroy, Thomas J. (Compiler); Martin, James J.

2010-01-01

The Early Flight Fission-Test Facilities (EFF-TF) team was tasked by Naval Reactors Prime Contract Team (NRPCT) to design, fabricate, and test an actively pumped lithium (Li) flow circuit. This Li circuit takes advantage of work in progress at the EFF TF on a stainless steel sodium/potassium (NaK) circuit. The effort involved modifying the original stainless steel NaK circuit such that it could be operated with Li in place of NaK. This new design considered freeze/thaw issues and required the addition of an expansion tank and expansion/extrusion volumes in the circuit plumbing. Instrumentation has been specified for Li and circuit heaters have been placed throughout the design to ensure adequate operational temperatures and no uncontrolled freezing of the Li. All major components have been designed and fabricated prior to circuit redesign for Li and were not modified. Basic circuit components include: reactor segment, Li to gas heat exchanger, electromagnetic liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and trace heaters. The reactor segment, based on a Los Alamos National Laboratory 100-kW design study with 120 fuel pins, is the only prototypic component in the circuit. However, due to earlier funding constraints, a 37-pin partial-array of the core, including the central three rings of fuel pins (pin and flow path dimensions are the same as those in the full design), was selected for fabrication and test. This Technical Publication summarizes the design and integration of the pumped liquid metal Li flow circuit as of May 1, 2005. This supplement contains drawings, analysis, and calculations

Documentation of Stainless Steel Lithium Circuit Test Section Design

NASA Technical Reports Server (NTRS)

Godfroy, T. J.; Martin, J. J.; Stewart, E. T.; Rhys, N. O.

2010-01-01

The Early Flight Fission-Test Facilities (EFF-TF) team was tasked by Naval Reactors Prime Contract Team (NRPCT) to design, fabricate, and test an actively pumped lithium (Li) flow circuit. This Li circuit takes advantage of work in progress at the EFF TF on a stainless steel sodium/potassium (NaK) circuit. The effort involved modifying the original stainless steel NaK circuit such that it could be operated with Li in place of NaK. This new design considered freeze/thaw issues and required the addition of an expansion tank and expansion/extrusion volumes in the circuit plumbing. Instrumentation has been specified for Li and circuit heaters have been placed throughout the design to ensure adequate operational temperatures and no uncontrolled freezing of the Li. All major components have been designed and fabricated prior to circuit redesign for Li and were not modified. Basic circuit components include: reactor segment, Li to gas heat exchanger, electromagnetic liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and trace heaters. The reactor segment, based on a Los Alamos National Laboratory 100-kW design study with 120 fuel pins, is the only prototypic component in the circuit. However, due to earlier funding constraints, a 37-pin partial-array of the core, including the central three rings of fuel pins (pin and flow path dimensions are the same as those in the full design), was selected for fabrication and test. This Technical Publication summarizes the design and integration of the pumped liquid metal Li flow circuit as of May 1, 2005.

ETR COMPLEX. CAMERA FACING SOUTH. FROM BOTTOM OF VIEW TO ...

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

ETR COMPLEX. CAMERA FACING SOUTH. FROM BOTTOM OF VIEW TO TOP: MTR, MTR SERVICE BUILDING, ETR CRITICAL FACILITY, ETR CONTROL BUILDING (ATTACHED TO ETR), ETR BUILDING (HIGH-BAY), COMPRESSOR BUILDING (ATTACHED AT LEFT OF ETR), HEAT EXCHANGER BUILDING (JUST BEYOND COMPRESSOR BUILDING), COOLING TOWER PUMP HOUSE, COOLING TOWER. OTHER BUILDINGS ARE CONTRACTORS' CONSTRUCTION BUILDINGS. INL NEGATIVE NO. 56-4105. Unknown Photographer, ca. 1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Metal halides vapor lasers with inner reactor and small active volume.

NASA Astrophysics Data System (ADS)

Shiyanov, D. V.; Sukhanov, V. B.; Evtushenko, G. S.

2018-04-01

Investigation of the energy characteristics of copper, manganese, lead halide vapor lasers with inner reactor and small active volume 90 cm3 was made. The optimal operating pulse repetition rates, temperatures, and buffer gas pressure for gas discharge tubes with internal and external electrodes are determined. Under identical pump conditions, such systems are not inferior in their characteristics to standard metal halide vapor lasers. It is shown that the use of a zeolite halogen generator provides lifetime laser operation.

Comparative performance of fixed-film biological filters: Application of reactor theory

USGS Publications Warehouse

Watten, B.J.; Sibrell, P.L.

2006-01-01

Nitrification is classified as a two-step consecutive reaction where R1 represents the rate of formation of the intermediate product NO2-N and R2 represents the rate of formation of the final product NO3-N. The relative rates of R1 and R2 are influenced by reactor type characterized hydraulically as plug-flow, plug-flow with dispersion and mixed-flow. We develop substrate conversion models for fixed-film biofilters operating in the first-order kinetic regime based on application of chemical reactor theory. Reactor type, inlet conditions and the biofilm kinetic constants Ki (h-1) are used to predict changes in NH4-N, NO2-N, NO3-N and BOD5. The inhibiting effects of the latter on R1 and R2 were established based on the ?? relation, e.g.:{A formula is presented}where BOD5,max is the concentration that causes nitrification to cease and N is a variable relating Ki to increasing BOD5. Conversion models were incorporated in spreadsheet programs that provided steady-state concentrations of nitrogen and BOD5 at several points in a recirculating aquaculture system operating with input values for fish feed rate, reactor volume, microscreen performance, make-up and recirculating flow rates. When rate constants are standardized, spreadsheet use demonstrates plug-flow reactors provide higher rates of R1 and R2 than mixed-flow reactors thereby reducing volume requirements for target concentrations of NH4-N and NO2-N. The benefit provided by the plug-flow reactor varies with hydraulic residence time t as well as the effective vessel dispersion number, D/??L. Both reactor types are capable of providing net increases in NO2-N during treatment but the rate of decrease in the mixed-flow case falls well behind that predicted for plug-flow operation. We show the potential for a positive net change in NO2-N increases with decreases in the dimensionless ratios K2, (R2 )/K1,( R1 ) and [NO2-N]/[NH4-N] and when the product K1, (R1) t provides low to moderate NH4-N conversions. Maintaining high levels of the latter reduces the effective reactor utilization rate (%) defined here as (RNavg/RNmax)100 where RNavg is the mean reactive nitrogen concentration ([NH4-N] + [NO2-N]) within the reactor, and RNmax represents the feed concentration of the same. Low utilization rates provide a hedge against unexpected increases in substrate loading and reduce water pumping requirements but force use of elevated reactor volumes. Further ?? effects on R1 and R2 can be reduced through use of a tanks-in-series versus a single mixed-flow reactor configuration and by improving the solids removal efficiency of microscreen treatment.

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

Goetsch, D.; Bieniussa, K.; Schulz, H.

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branchingmore » pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.« less

Pressure Loss Predictions of the Reactor Simulator Subsystem at NASA Glenn Research Center

NASA Technical Reports Server (NTRS)

Reid, Terry V.

2016-01-01

Testing of the Fission Power System (FPS) Technology Demonstration Unit (TDU) is being conducted at NASA Glenn Research Center. The TDU consists of three subsystems: the reactor simulator (RxSim), the Stirling Power Conversion Unit (PCU), and the heat exchanger manifold (HXM). An annular linear induction pump (ALIP) is used to drive the working fluid. A preliminary version of the TDU system (which excludes the PCU for now) is referred to as the "RxSim subsystem" and was used to conduct flow tests in Vacuum Facility 6 (VF 6). In parallel, a computational model of the RxSim subsystem was created based on the computer-aided-design (CAD) model and was used to predict loop pressure losses over a range of mass flows. This was done to assess the ability of the pump to meet the design intent mass flow demand. Measured data indicates that the pump can produce 2.333 kg/sec of flow, which is enough to supply the RxSim subsystem with a nominal flow of 1.75 kg/sec. Computational predictions indicated that the pump could provide 2.157 kg/sec (using the Spalart-Allmaras (S?A) turbulence model) and 2.223 kg/sec (using the k- turbulence model). The computational error of the predictions for the available mass flow is ?0.176 kg/sec (with the S-A turbulence model) and -0.110 kg/sec (with the k- turbulence model) when compared to measured data.

Reactor Simulator Testing

NASA Technical Reports Server (NTRS)

Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise Jon

2013-01-01

As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz. Keywords: fission, space power, nuclear, liquid metal, NaK.

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

Baylor, Larry R.; Meitner, Steven J.

Magnetically confined fusion plasmas generate energy from deuterium-tritium (DT) fusion reactions that produce energetic 3.5 MeV alpha particles and 14 MeV neutrons. Since the DT fusion reaction rate is a strong function of plasma density, an efficient fueling source is needed to maintain high plasma density in such systems. Energetic ions in fusion plasmas are able to escape the confining magnetic fields at a much higher rate than the fusion reactions occur, thus dictating the fueling rate needed. These lost ions become neutralized and need to be pumped away as exhaust gas to be reinjected into the plasma as fuelmore » atoms.The technology to fuel and pump fusion plasmas has to be inherently compatible with the tritium fuel. An ideal holistic solution would couple the pumping and fueling such that the pump exhaust is directly fed back into pellet formation without including impurity gases. This would greatly reduce the processing needs for the exhaust. Concepts to accomplish this are discussed along with the fueling and pumping needs for a DT fusion reactor.« less

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, M.M.; Lau, L.K.; Schulz, T.L.

1993-12-14

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

Fuel cycle for a fusion neutron source

NASA Astrophysics Data System (ADS)

Ananyev, S. S.; Spitsyn, A. V.; Kuteev, B. V.

2015-12-01

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion-fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium-tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium-tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

Charge exchange system

DOEpatents

Anderson, Oscar A.

1978-01-01

An improved charge exchange system for substantially reducing pumping requirements of excess gas in a controlled thermonuclear reactor high energy neutral beam injector. The charge exchange system utilizes a jet-type blanket which acts simultaneously as the charge exchange medium and as a shield for reflecting excess gas.

Intermediate-Size Inducer Pump design report. [LMFBR

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

Boardman, T.J.

1979-06-15

This report summarizes the mechanical, structural, and hydrodynamic design of the Intermediate-Size Inducer Pump (ISIP). The design was performed under Atomics International's DOE Base Technology Program by the Atomics International and Rocketdyne Divisions of Rockwell International. The pump was designed to utilize the FFTF prototype pump frame as a test vehicle to test the inducer, impeller, and diffuser plus necessary adapter hardware under simulated Large Scale Liquid Metal Fast Breeder Reactor service conditions. The report describes the design requirements including the purpose and objectives, and discusses those design efforts and considerations made to meet the requirements. Included in the reportmore » are appendices showing calculative methods and results. Also included are overall assembly and layout drawings plus some details used as illustrations for discussion of the design results and the results of water tests performed on a model of the inducer.« less

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.

An experimental investigation of 235 sub UF sub 6 fission produced plasmas. [gas handling system for use with nuclear pumped laser experiments

NASA Technical Reports Server (NTRS)

Miley, G. H.

1981-01-01

A gas handling system capable of use with uranium fluoride was designed and constructed for use with nuclear pumped laser experiments using the TRIGA research reactor. By employing careful design and temperature controls, the UF6 can be first transported into the irradiation chamber, and then, at the conclusion of the experiment, returned to gas cylinders. The design of the system is described. Operating procedures for the UF6 and gas handling systems are included.

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

James K. Neathery; Gary Jacobs; Amitava Sarkar

In the previous reporting period, modifications were completed for integrating a continuous wax filtration system for a 4 liter slurry bubble column reactor. During the current reporting period, a shakedown of the system was completed. Several problems were encountered with the progressive cavity pump used to circulate the wax/catalyst slurry though the cross-flow filter element and reactor. During the activation of the catalyst with elevated temperature (> 270 C) the elastomer pump stator released sulfur thereby totally deactivating the iron-based catalyst. Difficulties in maintaining an acceptable leak rate from the pump seal and stator housing were also encountered. Consequently, themore » system leak rate exceeded the expected production rate of wax; therefore, no online filtration could be accomplished. Work continued regarding the characterization of ultra-fine catalyst structures. The effect of carbidation on the morphology of iron hydroxide oxide particles was the focus of the study during this reporting period. Oxidation of Fe (II) sulfate results in predominantly {gamma}-FeOOH particles which have a rod-shaped (nano-needles) crystalline structure. Carbidation of the prepared {gamma}-FeOOH with CO at atmospheric pressure produced iron carbides with spherical layered structure. HRTEM and EDS analysis revealed that carbidation of {gamma}-FeOOH particles changes the initial nano-needles morphology and generates ultrafine carbide particles with irregular spherical shape.« less

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

DOEpatents

Furgerson, W.T.

1963-12-17

A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

LCRE and SNAP 50-DR-1 programs. Engineering progress report, October 1, 1962--December 31, 1962

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

None

Declassified 5 Sep 1973. Information is presented concerning LCRE specifications, reactor kinetics, fuel elements, primary coolant circuit, secondary coolant circuit, materials development, and fabrication; and SNAP50-DR- 1 specifications, primary pump, and materials development. (DCC)

Atmospheric pressure flow reactor: Gas phase chemical kinetics under tropospheric conditions without wall effects

NASA Technical Reports Server (NTRS)

Koontz, Steven L. (Inventor); Davis, Dennis D. (Inventor)

1991-01-01

A flow reactor for simulating the interaction in the troposphere is set forth. A first reactant mixed with a carrier gas is delivered from a pump and flows through a duct having louvers therein. The louvers straighten out the flow, reduce turbulence and provide laminar flow discharge from the duct. A second reactant delivered from a source through a pump is input into the flowing stream, the second reactant being diffused through a plurality of small diffusion tubes to avoid disturbing the laminar flow. The commingled first and second reactants in the carrier gas are then directed along an elongated duct where the walls are spaced away from the flow of reactants to avoid wall interference, disturbance or turbulence arising from the walls. A probe connected with a measuring device can be inserted through various sampling ports in the second duct to complete measurements of the first and second reactants and the product of their reaction at selected XYZ locations relative to the flowing system.

Solar solids reactor

DOEpatents

Yudow, B.D.

1986-02-24

A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

Solar solids reactor

DOEpatents

Yudow, Bernard D.

1987-01-01

A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

Multiphysics analysis of liquid metal annular linear induction pumps: A project overview

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

Maidana, Carlos Omar; Nieminen, Juha E.

Liquid metal-cooled fission reactors are both moderated and cooled by a liquid metal solution. These reactors are typically very compact and they can be used in regular electric power production, for naval and space propulsion systems or in fission surface power systems for planetary exploration. The coupling between the electromagnetics and thermo-fluid mechanical phenomena observed in liquid metal thermo-magnetic systems for nuclear and space applications gives rise to complex engineering magnetohydrodynamics and numerical problems. It is known that electromagnetic pumps have a number of advantages over rotating mechanisms: absence of moving parts, low noise and vibration level, simplicity of flowmore » rate regulation, easy maintenance and so on. However, while developing annular linear induction pumps, we are faced with a significant problem of magnetohydrodynamic instability arising in the device. The complex flow behavior in this type of devices includes a time-varying Lorentz force and pressure pulsation due to the time-varying electromagnetic fields and the induced convective currents that originates from the liquid metal flow, leading to instability problems along the device geometry. The determinations of the geometry and electrical configuration of liquid metal thermo-magnetic devices give rise to a complex inverse magnetohydrodynamic field problem were techniques for global optimization should be used, magnetohydrodynamics instabilities understood –or quantified- and multiphysics models developed and analyzed. Lastly, we present a project overview as well as a few computational models developed to study liquid metal annular linear induction pumps using first principles and the a few results of our multi-physics analysis.« less

Multiphysics analysis of liquid metal annular linear induction pumps: A project overview

DOE PAGES

Maidana, Carlos Omar; Nieminen, Juha E.

2016-03-14

Liquid metal-cooled fission reactors are both moderated and cooled by a liquid metal solution. These reactors are typically very compact and they can be used in regular electric power production, for naval and space propulsion systems or in fission surface power systems for planetary exploration. The coupling between the electromagnetics and thermo-fluid mechanical phenomena observed in liquid metal thermo-magnetic systems for nuclear and space applications gives rise to complex engineering magnetohydrodynamics and numerical problems. It is known that electromagnetic pumps have a number of advantages over rotating mechanisms: absence of moving parts, low noise and vibration level, simplicity of flowmore » rate regulation, easy maintenance and so on. However, while developing annular linear induction pumps, we are faced with a significant problem of magnetohydrodynamic instability arising in the device. The complex flow behavior in this type of devices includes a time-varying Lorentz force and pressure pulsation due to the time-varying electromagnetic fields and the induced convective currents that originates from the liquid metal flow, leading to instability problems along the device geometry. The determinations of the geometry and electrical configuration of liquid metal thermo-magnetic devices give rise to a complex inverse magnetohydrodynamic field problem were techniques for global optimization should be used, magnetohydrodynamics instabilities understood –or quantified- and multiphysics models developed and analyzed. Lastly, we present a project overview as well as a few computational models developed to study liquid metal annular linear induction pumps using first principles and the a few results of our multi-physics analysis.« less

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

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

Baek J.; Diamond D.; Cuadra, A.

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a modelmore » of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.« less

BWR Anticipated Transients Without Scram Leading to Instability

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

Cheng L. Y.; Baek J.; Cuadra, A.

2013-11-10

Anticipated transients without scram (ATWS) in aboiling water reactor (BWR) were simulated in order to understand reactor response and determine the effectiveness of automatic and operator actions to mitigate this beyond-design-basis accident. The events of interest herein are initiated by a turbine trip when the reactor is operating in the expanded operating domainMELLLA+ [maximum extended load line limit plus]. In these events the reactor may initially be at up to 120% of the original licensed thermal power (OLTP) and at flow rates as low as 80% of rated.For these (and similar) ATWS events the concern isthat when the reactor powermore » decreases in response to a dual recirculation pump trip, the core will become unstable and large amplitude oscillations will begin. The occurrence of these power oscillations, if left unmitigated, may result in fuel damage, and the amplitude of the poweroscillations may hamper the effectiveness of the injection of dissolved neutron absorber through the standby liquid control system (SLCS).« less

Novel Solar Photocatalytic Reactor for Wastewater Treatment

NASA Astrophysics Data System (ADS)

Sutisna; Rokhmat, M.; Wibowo, E.; Murniati, R.; Khairurrijal; Abdullah, M.

2017-07-01

A new solar photocatalytic reactor (photoreactor) using TiO2 nanoparticles coated onto plastic granules has been designed. Catalyst granules are placed into the cavity of a reactor panel made of glass. A pump is used to circulate wastewater in the photoreactor. Methylene blue (MB) dissolved in water was chosen as the wastewater model. The performance of the photoreactor was evaluated based on changes in MB concentration with respect to time. The photoreactor showed a good performance by degrading 10 L of MB solution up to 96.54% after 48 h of solar irradiation. The photoreactor was scaled up by enlarging the panel area to twice its original size. The increase in the surface area of the reactor panel and therefore of the mass of catalyst granules and reactor volume led to a three-fold increase of the photodegradation rate. In addition, the MB degradation kinetics were also studied. Data analysis confirmed the applicability of the pseudo-first-order Langmuir-Hinshelwood model. The proposed photoreactor has great potential for use in large-scale wastewater treatment.

Modeling and simulation of flow field in giant magnetostrictive pump

NASA Astrophysics Data System (ADS)

Zhao, Yapeng; Ren, Shiyong; Lu, Quanguo

2017-09-01

Recent years, there has been significant research in the design and analysis of giant magnetostrictive pump. In this paper, the flow field model of giant magnetostrictive pump was established and the relationship between pressure loss and working frequency of piston was studied by numerical simulation method. Then, the influence of different pump chamber height on pressure loss in giant magnetostrictive pump was studied by means of flow field simulation. Finally, the fluid pressure and velocity vector distribution in giant magnetostrictive pump chamber were simulated.

ETR AND MTR COMPLEXES IN CONTEXT. CAMERA FACING NORTHERLY. FROM ...

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

ETR AND MTR COMPLEXES IN CONTEXT. CAMERA FACING NORTHERLY. FROM BOTTOM TO TOP: ETR COOLING TOWER, ELECTRICAL BUILDING AND LOW-BAY SECTION OF ETR BUILDING, HEAT EXCHANGER BUILDING (WITH U SHAPED YARD), COMPRESSOR BUILDING. MTR REACTOR SERVICES BUILDING IS ATTACHED TO SOUTH WALL OF MTR. WING A IS ATTACHED TO BALCONY FLOOR OF MTR. NEAR UPPER RIGHT CORNER OF VIEW IS MTR PROCESS WATER BUILDING. WING B IS AT FAR WEST END OF COMPLEX. NEAR MAIN GATE IS GAMMA FACILITY, WITH "COLD" BUILDINGS BEYOND: RAW WATER STORAGE TANKS, STEAM PLANT, MTR COOLING TOWER PUMP HOUSE AND COOLING TOWER. INL NEGATIVE NO. 56-4101. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

HEAVY WATER MODERATED NEUTRONIC REACTOR

DOEpatents

Szilard, L.

1958-04-29

A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

Analysis of steam generator tube rupture transients with single failure

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

Trambauer, K.

The Gesellschaft fuer Reaktorsicherheit is engaged in the collection and evaluation of light water reactor operating experience as well as analyses for the risk study of the pressurized water reactor (PWR). Within these activities, thermohydraulic calculations have been performed to show the influence of different boundary conditions and disturbances on the steam generator tube rupture (SGTR) transients. The analyses of these calculations have focused on the measures and systems needed to cope with an SGTR. The reference plant for this analysis is a 1300-MW(e) PWR of Kraftwerk Union design with four loops, each containing a U-tube steam generator (SG) andmore » a reactor cooling pump (RCP). The thermal-hydraulic code DRUFAN-02 was used for the transient calculations.« less

40 CFR 63.11398 - What definitions apply to this subpart?

Code of Federal Regulations, 2013 CFR

2013-07-01

... acrylonitrile units: acrylic fiber or modacrylic fiber. Acrylonitrile solution polymerization means a process where acrylonitrile and comonomers are dissolved in a solvent to form a polymer solution (typically... resulting reactor polymer solution (spin dope) is filtered and pumped directly to the fiber spinning process...

40 CFR 63.11398 - What definitions apply to this subpart?

Code of Federal Regulations, 2014 CFR

2014-07-01

... acrylonitrile units: acrylic fiber or modacrylic fiber. Acrylonitrile solution polymerization means a process where acrylonitrile and comonomers are dissolved in a solvent to form a polymer solution (typically... resulting reactor polymer solution (spin dope) is filtered and pumped directly to the fiber spinning process...

Performance analysis of solar-assisted chemical heat-pump dryer

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

Fadhel, M.I.; Faculty of Engineering and Technology, Multimedia University, Jalan Ayer Keroh Lama, 75450, Melaka; Sopian, K.

2010-11-15

A solar-assisted chemical heat-pump dryer has been designed, fabricated and tested. The performance of the system has been studied under the meteorological conditions of Malaysia. The system consists of four main components: solar collector (evacuated tubes type), storage tank, solid-gas chemical heat pump unit and dryer chamber. A solid-gas chemical heat pump unit consists of reactor, condenser and evaporator. The reaction used in this study (CaCl2-NH{sub 3}). A simulation has been developed, and the predicted results are compared with those obtained from experiments. The maximum efficiency for evacuated tubes solar collector of 80% has been predicted against the maximum experimentmore » of 74%. The maximum values of solar fraction from the simulation and experiment are 0.795 and 0.713, respectively, whereas the coefficient of performance of chemical heat pump (COP{sup h}) maximum values 2.2 and 2 are obtained from simulation and experiments, respectively. The results show that any reduction of energy at condenser as a result of the decrease in solar radiation will decrease the coefficient of performance of chemical heat pump as well as decrease the efficiency of drying. (author)« less

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

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

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

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

Passive filtration of air egressing from nuclear containment

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

Malloy, III, John D

2017-09-26

A nuclear reactor includes a reactor core comprising fissile material disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor. A containment compartment contains the radiological containment. A heat sink includes a chimney configured to develop an upward-flowing draft in response to heated fluid flowing into a lower portion of the chimney. A fluid conduit is arranged to receive fluid from the containment compartment and to discharge into the chimney. A filter may be provided, with the fluid conduit including a first fluid conduit arranged to receive fluid from the containment compartment and to discharge into anmore » inlet of the filter, and a second fluid conduit arranged to receive fluid from an outlet of the filter and to discharge into the chimney. As the draft is developed passively, there is no need for a blower or pump configured to move fluid through the fluid conduit.« less

INDEPENDENT CONFIRMATORY SURVEY REPORT FOR THE REACTOR BUILDING, HOT LABORATORY, PRIMARY PUMP HOUSE, AND LAND AREAS AT THE PLUM BROOK REACTOR FACILITY, SANDUSKY, OHIO

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

Erika N. Bailey

2011-10-10

In 1941, the War Department acquired approximately 9,000 acres of land near Sandusky, Ohio and constructed a munitions plant. The Plum Brook Ordnance Works Plant produced munitions, such as TNT, until the end of World War II. Following the war, the land remained idle until the National Advisory Committee for Aeronautics later called the National Aeronautics and Space Administration (NASA) obtained 500 acres to construct a nuclear research reactor designed to study the effects of radiation on materials used in space flight. The research reactor was put into operation in 1961 and was the first of fifteen test facilities eventuallymore » built by NASA at the Plum Brook Station. By 1963, NASA had acquired the remaining land at Plum Brook for these additional test facilities« less

Thermonuclear inverse magnetic pumping power cycle for stellarator reactor

DOEpatents

Ho, Darwin D.; Kulsrud, Russell M.

1991-01-01

The plasma column in a stellarator is compressed and expanded alternatively in minor radius. First a plasma in thermal balance is compressed adiabatically. The volume of the compressed plasma is maintained until the plasma reaches a new thermal equilibrium. The plasma is then expanded to its original volume. As a result of the way a stellarator works, the plasma pressure during compression is less than the corresponding pressure during expansion. Therefore, negative work is done on the plasma over a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils. Direct electrical energy is obtained from this voltage. Alternatively, after the compression step, the plasma can be expanded at constant pressure. The cycle can be made self-sustaining by operating a system of two stellarator reactors in tandem. Part of the energy derived from the expansion phase of a first stellarator reactor is used to compress the plasma in a second stellarator reactor.

Electromagnetic Pumps for Conductive-Propellant Feed Systems

NASA Technical Reports Server (NTRS)

Markusic, T. E.; Polzin, K. A.

2005-01-01

There has been a recent, renewed interest in high-power electric thrusters for application in nuclear-electric propulsion systems. Two of the most promising thrusters utilize liquid metal propellants: the lithium-fed magnetoplasmadynamic thruster and the bismuth-fed Hall thruster. An important element of part of the maturation of these thrusters will be the development of compact, reliable conductive-propellant feed system components. In the present paper we provide design considerations and experimental calibration data for electromagnetic (EM) pumps. The role of an electromagnetic pump in a liquid metal feed system is to establish a pressure gradient between the propellant reservoir and the thruster - to establish the requisite mass flow rate. While EM pumps have previously been used to a limited extent in nuclear reactor cooling loops, they have never been implemented in electric propulsion (EP) systems. The potential benefit of using EM pumps for EP are reliability (no moving parts) and the ability to precisely meter the propellant flow rate. We have constructed and tested EM pumps that use gallium, lithium, and bismuth propellants. Design details, test results (pressure developed versus current), and material compatibility issues are reported. It is concluded that EM pumps are a viable technology for application in both laboratory and flight EP conductive-propellant feed systems.

Pump and Signal Taper for Airclad Fibers

DTIC Science & Technology

2006-01-05

as follows: Crystal Fibre A/S will develop a taper/coupler solution to interface between a new polarization maintaining/polarizing amplifier fiber ...MM) pump combiner with a high NA air-clad output. The input side of the combiner is 7 individual MM pump delivery solid all- glass fibers . The NA of...pump combiner. MOTIVATION FINAL REPORT ITEM 0002 In a typical standard fused fiber coupler a number of all- glass 0.22 NA pump

New Discrete Fibonacci Charge Pump Design, Evaluation and Measurement

NASA Astrophysics Data System (ADS)

Matoušek, David; Hospodka, Jiří; Šubrt, Ondřej

2017-06-01

This paper focuses on the practical aspects of the realisation of Dickson and Fibonacci charge pumps. Standard Dickson charge pump circuit solution and new Fibonacci charge pump implementation are compared. Both charge pumps were designed and then evaluated by LTspice XVII simulations and realised in a discrete form on printed circuit board (PCB). Finally, the key parameters as the output voltage, efficiency, rise time, variable power supply and clock frequency effects were measured.

Component and Technology Development for Advanced Liquid Metal Reactors

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

Anderson, Mark

2017-01-30

The following report details the significant developments to Sodium Fast Reactor (SFR) technologies made throughout the course of this funding. This report will begin with an overview of the sodium loop and the improvements made over the course of this research to make it a more advanced and capable facility. These improvements have much to do with oxygen control and diagnostics. Thus a detailed report of advancements with respect to the cold trap, plugging meter, vanadium equilibration loop, and electrochemical oxygen sensor is included. Further analysis of the university’s moving magnet pump was performed and included in a section ofmore » this report. A continuous electrical resistance based level sensor was built and tested in the sodium with favorable results. Materials testing was done on diffusion bonded samples of metal and the results are presented here as well. A significant portion of this work went into the development of optical fiber temperature sensors which could be deployed in an SFR environment. Thus, a section of this report presents the work done to develop an encapsulation method for these fibers inside of a stainless steel capillary tube. High temperature testing was then done on the optical fiber ex situ in a furnace. Thermal response time was also explored with the optical fiber temperature sensors. Finally these optical fibers were deployed successfully in a sodium environment for data acquisition. As a test of the sodium deployable optical fiber temperature sensors they were installed in a sub-loop of the sodium facility which was constructed to promote the thermal striping effect in sodium. The optical fibers performed exceptionally well, yielding unprecedented 2 dimensional temperature profiles with good temporal resolution. Finally, this thermal striping loop was used to perform cross correlation velocimetry successfully over a wide range of flow rates.« less

Commercialization of PV-powered pumping systems for use in utility PV service programs. Final report

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

NONE

1997-03-01

The project described in this report was a commercialization effort focused on cost-effective remote water pumping systems for use in utility-based photovoltaic (PV) service programs. The project combined a commercialization strategy tailored specifically for electric utilities with the development of a PV-powered pumping system that operates conventional ac pumps rather than relying on the more expensive and less reliable PV pumps on the market. By combining these two attributes, a project goal was established of creating sustained utility purchases of 250 PV-powered water pumping systems per year. The results of each of these tasks are presented in two parts containedmore » in this Final Summary Report. The first part summarizes the results of the Photovoltaic Services Network (PSN) as a new business venture, while the second part summarizes the results of the Golden Photon system installations. Specifically, results and photographs from each of the system installations are presented in this latter part.« less

Performance Testing of a Liquid Metal Pump for In-Space Power Systems

NASA Technical Reports Server (NTRS)

Polzin, Kurt

2011-01-01

Fission surface power (FSP) systems could be used to provide power on the surface of the moon, Mars, or other planets and moons of our solar system. Fission power systems could provide excellent performance at any location, including those near the poles or other permanently shaded regions, and offer the capability to provide on demand power at any time, even at large distances from the sun. Fission-based systems also offer the potential for outposts, crew and science instruments to operate in a power-rich environment. NASA has been exploring technologies with the goal of reducing the cost and technical risk of employing FSP systems. A reference 40 kWe option has been devised that is cost-competitive with alternatives while providing more power for less mass anywhere on the lunar surface. The reference FSP system is also readily extensible for use on Mars, where it would be capable of operating through global dust storms and providing year-round power at any Martian latitude. Detailed development of the FSP concept and the reference mission are documented in various other reports. The development discussed in this paper prepares the way for testing of the Technology Demonstration Unit (TDU), which is a 10 kWe end-to-end test of FSP technologies intended to raise the entire FSP system to technology readiness level (TRL) 6. The Early Flight Fission Test Facility (EFF-TF) was established by NASA s Marshall Space Flight Center (MSFC) to provide a capability for performing hardware-directed activities to support multiple in-space nuclear reactor concepts by using a nonnuclear test methodology. This includes fabrication and testing at both the module/component level and at near prototypic reactor components and configurations allowing for realistic thermal-hydraulic evaluations of systems. The liquid-metal pump associated with the FSP system must be compatible with the liquid NaK coolant and have adequate performance to enable a viable flight system. Idaho National Laboratory (INL) was tasked with the modeling, design, and fabrication of an ALIP suitable for the FSP reference mission. A prototypic ALIP was fabricated under the direction of INL and shipped to MSFC for inclusion in the Technology Demonstration Unit (TDU), a quarter-scale end-to-end reactor simulator system that is scheduled for testing at NASA-GRC. Before inclusion in the TDU, the ALIP was tested in the ALIP test circuit (ATC), which is a rig developed and operated at MSFC for the specific purpose of providing accurate quantification of liquid metal pump performance. Data showing the pump performance curves (pressure, flowrate, and pump efficiency) are presented for various operating power levels, demonstrating the full performance envelope of the pump.

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

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

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

Mars Propellant Production with Ionic Liquids Project

NASA Technical Reports Server (NTRS)

Falker, John; Thompson, Karen; Zeitlin, Nancy; Muscatello, Anthony

2015-01-01

This project seeks to develop a single vessel for carbon dioxide (CO2) capture and electrolysis for in situ Mars propellant production by eliminating several steps of CO2 processing, two cryocoolers, a high temperature reactor, a recycle pump, and a water condenser; thus greatly reducing mass, volume, and power.

Liquid oxygen/liquid hydrogen boost/vane pump for the advanced orbit transfer vehicles auxiliary propulsion system

NASA Technical Reports Server (NTRS)

Gluzek, F.; Mokadam, R. G.; To, I. H.; Stanitz, J. D.; Wollschlager, J.

1979-01-01

A rotating, positive displacement vane pump with an integral boost stage was designed to pump saturated liquid oxygen and liquid hydrogen for auxiliary propulsion system of orbit transfer vehicle. This unit is designed to ingest 10% vapor by volume, contamination free liquid oxygen and liquid hydrogen. The final pump configuration and the predicted performance are included.

Stainless Steel NaK Circuit Integration and Fill Submission

NASA Technical Reports Server (NTRS)

Garber, Anne E.

2006-01-01

The Early Flight Fission Test Facilities (EFF-TF) team has been tasked by the Marshall Space Flight Center Nuclear Systems Office to design, fabricate, and test an actively pumped alkali metal flow circuit. The system, which was originally designed to hold a eutectic mixture of sodium potassium (NaK), was redesigned to hold lithium; but due to a shift in focus, it is once again being prepared for use with NaK. Changes made to the actively pumped, high temperature loop include the replacement of the expansion reservoir, addition of remotely operated valves, and modification of the support table. Basic circuit components include: reactor segment, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and a spill reservoir. A 37-pin partial-array core (pin and flow path dimensions are the same as those in a full design) was selected for fabrication and test. This document summarizes the integration and fill of the pumped liquid metal NaK flow circuit.

A Course in Chemical Reactor Design.

ERIC Educational Resources Information Center

Takoudis, Christos G.

1983-01-01

Presents course outline, topics covered, and final project (doubling as a take home final exam) for a one-semester, interdisciplinary course on the design and behavior of chemical reactors. Interplay of chemical and physical rate processes is stressed in the course. (JM)

Heat exchanger with auxiliary cooling system

DOEpatents

Coleman, John H.

1980-01-01

A heat exchanger with an auxiliary cooling system capable of cooling a nuclear reactor should the normal cooling mechanism become inoperable. A cooling coil is disposed around vertical heat transfer tubes that carry secondary coolant therethrough and is located in a downward flow of primary coolant that passes in heat transfer relationship with both the cooling coil and the vertical heat transfer tubes. A third coolant is pumped through the cooling coil which absorbs heat from the primary coolant which increases the downward flow of the primary coolant thereby increasing the natural circulation of the primary coolant through the nuclear reactor.

Topological Floquet-Thouless Energy Pump

NASA Astrophysics Data System (ADS)

Kolodrubetz, Michael H.; Nathan, Frederik; Gazit, Snir; Morimoto, Takahiro; Moore, Joel E.

2018-04-01

We explore adiabatic pumping in the presence of a periodic drive, finding a new phase in which the topologically quantized pumped quantity is energy rather than charge. The topological invariant is given by the winding number of the micromotion with respect to time within each cycle, momentum, and adiabatic tuning parameter. We show numerically that this pump is highly robust against both disorder and interactions, breaking down at large values of either in a manner identical to the Thouless charge pump. Finally, we suggest experimental protocols for measuring this phenomenon.

Four wave mixing oscillation in a semiconductor microcavity: generation of two correlated polariton populations.

PubMed

Romanelli, M; Leyder, C; Karr, J Ph; Giacobino, E; Bramati, A

2007-03-09

We demonstrate a novel kind of polariton four wave mixing oscillation. Two pump polaritons scatter towards final states that emit two beams of equal intensity, separated both spatially and in polarization with respect to the pumps. The measurement of the intensity fluctuations of the emitted light demonstrates that the final states are strongly correlated.

Fluidized bed coupled rotary reactor for nanoparticles coating via atomic layer deposition

NASA Astrophysics Data System (ADS)

Duan, Chen-Long; Liu, Xiao; Shan, Bin; Chen, Rong

2015-07-01

A fluidized bed coupled rotary reactor has been designed for coating on nanoparticles (NPs) via atomic layer deposition. It consists of five major parts: reaction chamber, dosing and fluidizing section, pumping section, rotary manipulator components, as well as a double-layer cartridge for the storage of particles. In the deposition procedure, continuous fluidization of particles enlarges and homogenizes the void fraction in the particle bed, while rotation enhances the gas-solid interactions to stabilize fluidization. The particle cartridge presented here enables both the fluidization and rotation acting on the particle bed, demonstrated by the analysis of pressure drop. Moreover, enlarged interstitials and intense gas-solid contact under sufficient fluidizing velocity and proper rotation speed facilitate the precursor delivery throughout the particle bed and consequently provide a fast coating process. The cartridge can ensure precursors flowing through the particle bed exclusively to achieve high utilization without static exposure operation. By optimizing superficial gas velocities and rotation speeds, minimum pulse time for complete coating has been shortened in experiment, and in situ mass spectrometry showed the precursor usage can reach 90%. Inductively coupled plasma-optical emission spectroscopy results suggested a saturated growth of nanoscale Al2O3 films on spherical SiO2 NPs. Finally, the uniformity and composition of the shells were characterized by high angle annular dark field-transmission electron microscopy and energy dispersive X-ray spectroscopy.

Hardware Progress Made in the Early Flight Fission Test Facilities (EFF-TF) To Support Near-Term Space Fission Systems

NASA Astrophysics Data System (ADS)

Van Dyke, Melissa; Martin, James

2005-02-01

The NASA Marshall Space Flight Center's Early Flight Fission Test Facility (EFF-TF), provides a facility to experimentally evaluate nuclear reactor related thermal hydraulic issues through the use of non-nuclear testing. This facility provides a cost effective method to evaluate concepts/designs and support mitigation of developmental risk. Electrical resistance thermal simulators can be used to closely mimic the heat deposition of the fission process, providing axial and radial profiles. A number of experimental and design programs were underway in 2004 which include the following. Initial evaluation of the Department of Energy Los Alamos National Laboratory 19 module stainless steel/sodium heat pipe reactor with integral gas heat exchanger was operated at up to 17.5 kW of input power at core temperatures of 1000 K. A stainless steel sodium heat pipe module was placed through repeated freeze/thaw cyclic testing accumulating over 200 restarts to a temperature of 1000 K. Additionally, the design of a 37- pin stainless steel pumped sodium/potassium (NaK) loop was finalized and components procured. Ongoing testing at the EFF-TF is geared towards facilitating both research and development necessary to support future decisions regarding potential use of space nuclear systems for space exploration. All efforts are coordinated with DOE laboratories, industry, universities, and other NASA centers. This paper describes some of the 2004 efforts.

Grout Isolation and Stabilization of Structures and Materials within Nuclear Facilities at the U.S. Department of Energy, Hanford Site, Summary - 12309

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

Phillips, S.J.; Phillips, M.; Etheridge, D.

2012-07-01

Per regulatory agreement and facility closure design, U.S. Department of Energy Hanford Site nuclear fuel cycle structures and materials require in situ isolation in perpetuity and/or interim physicochemical stabilization as a part of final disposal or interim waste removal, respectively. To this end, grout materials are being used to encase facilities structures or are being incorporated within structures containing hazardous and radioactive contaminants. Facilities where grout materials have been recently used for isolation and stabilization include: (1) spent fuel separations, (2) uranium trioxide calcining, (3) reactor fuel storage basin, (4) reactor fuel cooling basin transport rail tanker cars and casks,more » (5) cold vacuum drying and reactor fuel load-out, and (6) plutonium fuel metal finishing. Grout components primarily include: (1) portland cement, (2) fly ash, (3) aggregate, and (4) chemical admixtures. Mix designs for these typically include aggregate and non aggregate slurries and bulk powders. Placement equipment includes: (1) concrete piston line pump or boom pump truck for grout slurry, (2) progressive cavity and shearing vortex pump systems, and (3) extendable boom fork lift for bulk powder dry grout mix. Grout slurries placed within the interior of facilities were typically conveyed utilizing large diameter slick line and the equivalent diameter flexible high pressure concrete conveyance hose. Other facilities requirements dictated use of much smaller diameter flexible grout conveyance hose. Placement required direct operator location within facilities structures in most cases, whereas due to radiological dose concerns, placement has also been completed remotely with significant standoff distances. Grout performance during placement and subsequent to placement often required unique design. For example, grout placed in fuel basin structures to serve as interim stabilization materials required sufficient bearing i.e., unconfined compressive strength, to sustain heavy equipment yet, low breakout force to permit efficient removal by track hoe bucket or equivalent construction equipment. Further, flow of slurries through small orifice geometries of moderate head pressures was another typical design requirement. Phase separation of less than 1 percent was a typical design requirement for slurries. On the order of 30,000 cubic meters of cementitious grout have recently been placed in the above noted U.S. Department of Energy Hanford Site facilities or structures. Each has presented a unique challenge in mix design, equipment, grout injection or placement, and ultimate facility or structure performance. Unconfined compressive and shear strength, flow, density, mass attenuation coefficient, phase separation, air content, wash-out, parameters and others, unique to each facility or structure, dictate the grout mix design for each. Each mix design was tested under laboratory and scaled field conditions as a precursor to field deployment. Further, after injection or placement of each grout formulation, the material was field inspected either by standard laboratory testing protocols, direct physical evaluation, or both. (authors)« less

Direct liquefaction Proof-of-Concept facility. Final technical progress report

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

Comolli, A.G.; Lee, L.K.; Pradhan, V.R.

1995-08-01

This report presents the results of work which included extensive modifications to HRI`s existing 3 ton per day Process Development Unit (PDU) and completion of the first PDU run. The 58-day Run 1 demonstrated scale-up of the Catalytic Two-Stage Liquefaction (CTSL Process) on Illinois No. 6 coal to produce distillate liquid products at a rate of up to 5 barrels per to of moisture-ash-free coal. The Kerr McGee Rose-SR unit from Wilsonville was redesigned and installed next to the US Filter installation to allow a comparison of the two solids removal systems. Also included was a new enclosed reactor tower,more » upgraded computer controls and a data acquisition system, an alternate power supply, a newly refurbished reactor, an in-line hydrotreater, interstage sampling system, coal handling unit, a new ebullating pump, load cells and improved controls and remodeled preheaters. Distillate liquid yields of 5 barrels/ton of moisture ash free coal were achieved. Coal slurry recycle rates were reduced from the 2--2.5 to 1 ratio demonstrated at Wilsonville to as low as 0.9 to 1. Coal feed rates were increased during the test by 50% while maintaining process performance at a marginally higher reactor severity. Sulfur in the coal was reduced from 4 wt% to ca. 0.02 wt% sulfur in the clean distillate fuel product. More than 3,500 gallons of distillate fuels were collected for evaluation and upgrading studies. The ROSE-SR Process was operated for the first time with a pentane solvent in a steady-state model. The energy rejection of the ash concentrate was consistently below prior data, being as low as 12%, allowing improved liquid yields and recovery.« less

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

Spatial atomic layer deposition for coating flexible porous Li-ion battery electrodes

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

Yersak, Alexander S.; Sharma, Kashish; Wallas, Jasmine M.

Ultrathin atomic layer deposition (ALD) coatings on the electrodes of Li-ion batteries can enhance the capacity stability of the Li-ion batteries. To commercialize ALD for Li-ion battery production, spatial ALD is needed to decrease coating times and provide a coating process compatible with continuous roll-to-roll (R2R) processing. The porous electrodes of Li-ion batteries provide a special challenge because higher reactant exposures are needed for spatial ALD in porous substrates. This work utilized a modular rotating cylinder spatial ALD reactor operating at rotation speeds up to 200 revolutions/min (RPM) and substrate speeds up to 200 m/min. The conditions for spatial ALDmore » were adjusted to coat flexible porous substrates. The reactor was initially used to characterize spatial Al2O3 and ZnO ALD on flat, flexible metalized polyethylene terephthalate foils. These studies showed that slower rotation speeds and spacers between the precursor module and the two adjacent pumping modules could significantly increase the reactant exposure. The modular rotating cylinder reactor was then used to coat flexible, model porous anodic aluminum oxide (AAO) membranes. The uniformity of the ZnO ALD coatings on the porous AAO membranes was dependent on the aspect ratio of the pores and the reactant exposures. Larger reactant exposures led to better uniformity in the pores with higher aspect ratios. The reactant exposures were increased by adding spacers between the precursor module and the two adjacent pumping modules. The modular rotating cylinder reactor was also employed for Al2O3 ALD on porous LiCoO2 (LCO) battery electrodes. Uniform Al coverages were obtained using spacers between the precursor module and the two adjacent pumping modules at rotation speeds of 25 and 50 RPM. The LCO electrodes had a thickness of ~49 um and pores with aspect ratios of ~12-25. Coin cells were then constructed using the ALD-coated LCO electrodes and were tested to determine their battery performance. The capacity of the Al2O3 ALD-coated LCO battery electrodes was measured versus the number of charge-discharge cycles. Both temporal and spatial ALD processing methods led to higher capacity stability compared with uncoated LCO battery electrodes. The results for improved battery performance were comparable for temporal and spatial ALD-coated electrodes. The next steps are also presented for scale-up to R2R spatial ALD using the modular rotating cylinder reactor.« less

76 FR 58050 - Tennessee Valley Authority, Bellefonte Nuclear Power Plant, Unit 1; Environmental Assessment and...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-19

...--federal candidate) were identified in the TVA Biological Assessment (BA) as occurring in areas potentially... major pumps, motors, heat exchangers, tanks, and piping; refurbish major equipment, such as reactor... actions with SIPs, amends 40 CFR part 51, Subpart W, and specifically identifies tribal agencies as...

76 FR 52699 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-23

... controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the... New analysis limit 10 CFR 50.67 limit EAB 4.56 rem TEDE Main steam line isolation. Closing the pressurizer power-operated relief valves block valves. Securing all reactor coolant pumps. Feedwater isolation...

76 FR 50759 - Applications and Amendments to Facility Operating Licenses Involving Proposed No Significant...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-16

... addition, the requestor/petitioner shall provide a brief explanation of the bases for the contention and a... trip, various pumps and coolers to start, and various valves to open and close), Containment Isolation (closes valves to isolate the Reactor Building interior from the environment), Steam isolation (close all...

Treatment of smuggled cigarette tobacco by composting process in facultative reactors.

PubMed

Zittel, Rosimara; Pinto da Silva, Cleber; Domingues, Cinthia Eloise; de Oliveira Stremel, Tatiana Roselena; de Almeida, Thiago Eduardo; Vieira Damiani, Gislaine; Xavier de Campos, Sandro

2018-01-01

This paper presents a study on the degradation of smuggled cigarette tobacco combined with domestic organic waste and sawdust or wood chips, using facultative reactor. Four reactors with different amounts of residue were assembled. For the study of the quality of the compost obtained, physicochemical, phytotoxicity and microbiological analyses were carried out. The mixture with wood chips presented the best temperature conditions and pH variation optimizing the degradation. The final germination index (GI) values of all treatments were above the recommended GI value (50%) and the final C/N ratio between 8 and 13 indicated a mature compost. The concentration of metals under study was below the limit allowed for the commercialization. The composting carried out in all facultative reactors provided ideal conditions for the total sterilization of the final compost. Therefore, the treatment of smuggled cigarettes through facultative reactors was efficient to produce stable and mature compost. Copyright © 2017 Elsevier Ltd. All rights reserved.

Centrifugal Force Based Magnetic Micro-Pump Driven by Rotating Magnetic Fields

NASA Astrophysics Data System (ADS)

Kim, S. H.; Hashi, S.; Ishiyama, K.

2011-01-01

This paper presents a centrifugal force based magnetic micro-pump for the pumping of blood. Most blood pumps are driven by an electrical motor with wired control. To develop a wireless and battery-free blood pump, the proposed pump is controlled by external rotating magnetic fields with a synchronized impeller. Synchronization occurs because the rotor is divided into multi-stage impeller parts and NdFeB permanent magnet. Finally, liquid is discharged by the centrifugal force of multi-stage impeller. The proposed pump length is 30 mm long and19 mm in diameter which much smaller than currently pumps; however, its pumping ability satisfies the requirement for a blood pump. The maximum pressure is 120 mmHg and the maximum flow rate is 5000ml/min at 100 Hz. The advantage of the proposed pump is that the general mechanical problems of a normal blood pump are eliminated by the proposed driving mechanism.

A monolithic lipase reactor for biodiesel production by transesterification of triacylglycerides into fatty acid methyl esters

PubMed Central

Urban, Jiri; Svec, Frantisek; Fréchet, Jean M.J.

2011-01-01

An enzymatic reactor with lipase immobilized on a monolithic polymer support has been prepared and used to catalyze the transesterification of triacylglycerides into the fatty acid methyl esters commonly used for biodiesel. A design of experiments procedure was used to optimize the monolithic reactor with variables including control of the surface polarity of the monolith via variations in the length of the hydrocarbon chain in alkyl methacrylate monomer, time of grafting of 1-vinyl-4,4-dimethylazlactone used to activate the monolith, and time used for the immobilization of porcine lipase. Optimal conditions involved the use of a poly(stearyl methacrylate-co-ethylene dimethacrylate) monolith, grafted first with vinylazlactone, then treated with lipase for 2 h to carry out the immobilization of the enzyme. Best conditions for the transesterification of glyceryl tributyrate included a temperature of 37°C and a 10 min residence time of the substrate in the bioreactor. The reactor did not lose its activity even after pumping through it a solution of substrate equaling 1,000 reactor volumes. This enzymatic reactor was also used for the transesterification of triacylglycerides from soybean oil to fatty acid methyl esters thus demonstrating the ability of the reactor to produce biodiesel. PMID:21915852

A monolithic lipase reactor for biodiesel production by transesterification of triacylglycerides into fatty acid methyl esters.

PubMed

Urban, Jiri; Svec, Frantisek; Fréchet, Jean M J

2012-02-01

An enzymatic reactor with lipase immobilized on a monolithic polymer support has been prepared and used to catalyze the transesterification of triacylglycerides into the fatty acid methyl esters commonly used for biodiesel. A design of experiments procedure was used to optimize the monolithic reactor with variables including control of the surface polarity of the monolith via variations in the length of the hydrocarbon chain in alkyl methacrylate monomer, time of grafting of 1-vinyl-4,4-dimethylazlactone used to activate the monolith, and time used for the immobilization of porcine lipase. Optimal conditions involved the use of a poly(stearyl methacrylate-co-ethylene dimethacrylate) monolith, grafted first with vinylazlactone, then treated with lipase for 2 h to carry out the immobilization of the enzyme. Best conditions for the transesterification of glyceryl tributyrate included a temperature of 37°C and a 10 min residence time of the substrate in the bioreactor. The reactor did not lose its activity even after pumping through it a solution of substrate equaling 1,000 reactor volumes. This enzymatic reactor was also used for the transesterification of triacylglycerides from soybean oil to fatty acid methyl esters thus demonstrating the ability of the reactor to produce biodiesel. Copyright © 2011 Wiley Periodicals, Inc.

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

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

Gorzel, A.

2006-07-01

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a Germanmore » BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)« less

Motorcycle emission control demonstration. Final report April--November 1977

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

Ullman, T.L.; Hare, C.T.

1977-12-01

The testing of ten motorcycles for exhaust emissions of hydrocarbons (HC), carbon monoxide (CO), and oxides of nitrogen (NOx) is described. Emission rates in grams per kilometer are presented for the 1975 Federal Test Procedure and the Highway Fuel Economy Test, as well as in percent or parts per million for several steady-state conditions. The test motorcycles were equipped with 2-stroke, 4-stroke, and rotary engines ranging from about 100 to 1000 cc displacement. Several exhaust emission control techniques were applied, including minor and major carburetor enleanment with and without capacitive discharge ignition, secondary air injection with both pumps and reedmore » valves, port liners, thermal reactors, oxidation catalysts, rotary valve modification, a spark ignited afterburner, and a short-circuited air-fuel mixture extraction system. Driveability, performance, and maladjustment (stock only) testing were also performed. All testing was performed with low accumulated distance on the motorcycles. (Portions of this document are not fully legible)« less

75 FR 57302 - Advisory Committee on Reactor Safeguards; Public Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-20

... Monday, October 14, 2009, (74 FR 52829-52830). Thursday, October 7, 2010, Conference Room T2-B1, Two....: Final Safety Evaluation Report Associated with the Economic Simplified Boiling Water Reactor (ESBWR..., Inc. regarding the final Safety Evaluation Report associated with the ESBWR design certification...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

Multi-Function Gas Fired Heat Pump

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

Abu-Heiba, Ahmad; Vineyard, Edward Allan

2015-11-01

The aim of this project was to design a residential fuel fired heat pump and further improve efficiency in collaboration with an industry partner – Southwest Gas, the developer of the Nextaire commercial rooftop fuel-fired heat pump. Work started in late 2010. After extensive search for suitable engines, one manufactured by Marathon was selected. Several prototypes were designed and built over the following four years. Design changes were focused on lowering the cost of components and the cost of manufacturing. The design evolved to a final one that yielded the lowest cost. The final design also incorporates noise and vibrationmore » reduction measures that were verified to be effective through a customer survey. ETL certification is currently (as of November 2015) underway. Southwest Gas is currently in talks with GTI to reach an agreement through which GTI will assess the commercial viability and potential of the heat pump. Southwest Gas is searching for investors to manufacture the heat pump and introduce it to the market.« less

NASA Ames Hosts Viewing Party for Final Shuttle Launch (Reporter Package)

NASA Image and Video Library

2011-07-12

The public was invited to NASA's Ames Research Center to observe a live televised broadcast of the final space shuttle launch on July 8, 2011. The STS-135 mission is the final flight of NASA's Space Shuttle Program. The orbiter Atlantis is carrying a system to investigate the potential for robotically refueling existing spacecraft and bring back a failed ammonia pump to help NASA better understand and improve pump designs for future systems. It also will deliver spare parts to sustain space station operations after the shuttles retire from service.

76 FR 8637 - Medical Devices; Medical Device Data Systems

Federal Register 2010, 2011, 2012, 2013, 2014

2011-02-15

... would alter the parameters on an infusion pump. The MDDS could pass that control signal to the infusion... proposed Sec. 880.6310(a). Commenters also asked whether a system that sends data to an infusion pump to... signal to an infusion pump to control the flow rate would not be an MDDS because, as the revised final...

Diode-pumped Tunable 3 Micron Laser Sources

DTIC Science & Technology

2000-02-21

DoD Ballistic Missile Defense Organization U.S. Army Space and Missile Defense Command SBIR Phase I Final Report AC Materials, Inc. 2721 Forsyth...pumped tunable 3 micron laser sources 6. AUTHORISI Arlete Cassanho, Hans Jenssen 7. PERFORMING ORGANIZATION NAME(S) AND ADDRESS(ES) AC Materials, Inc...impurities in the final crystal, starting materials for the crystal growth were prepared at AC Materials from optical grade barium fluoride and

GaAs laser diode pumped Nd:YAG laser

NASA Technical Reports Server (NTRS)

Conant, L. C.; Reno, C. W.

1974-01-01

A 1.5-mm by 3-cm neodymium-ion doped YAG laser rod has been side pumped using a GaAs laser diode array tuned to the 8680-A absorption line, achieving a multimode average output power of 120 mW for a total input power of 20 W to the final-stage laser diode drivers. The pumped arrangement was designed to take advantage of the high brightness of a conventional GaAs array as a linear source by introducing the pump light through a slit into a close-wrapped gold coated pump cavity. This cavity forms an integrating chamber for the pump light.

Inherent Safety Characteristics of Advanced Fast Reactors

NASA Astrophysics Data System (ADS)

Bochkarev, A. S.; Korsun, A. S.; Kharitonov, V. S.; Alekseev, P. N.

2017-01-01

The study presents SFR transient performance for ULOF events initiated by pump trip and pump seizure with simultaneous failure of all shutdown systems in both cases. The most severe cases leading to the pin cladding rupture and possible sodium boiling are demonstrated. The impact of various features on SFR inherent safety performance for ULOF events was analysed. The decrease in hydraulic resistance of primary loop and increase in primary pump coast down time were investigated. Performing analysis resulted in a set of recommendations to varying parameters for the purpose of enhancing the inherent safety performance of SFR. In order to prevent the safety barrier rupture for ULOF events the set of thermal hydraulic criteria defining the ULOF transient processes dynamics and requirements to these criteria were recommended based on achieved results: primary sodium flow dip under the natural circulation asymptotic level and natural circulation rise time.

Experimental investigations of helium cryotrapping by argon frost

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

Mack, A.; Perinic, D.; Murdoch, D.

1992-03-01

At the Karlsruhe Nuclear Research Centre (KfK) cryopumping techniques are being investigated by which the gaseous exhausts from the NET/ITER reactor can be pumped out during the burn-and dwell-times. Cryosorption and cryotrapping are techniques which are suitable for this task. It is the target of the investigations to test the techniques under NET/ITER conditions and to determine optimum design data for a prototype. They involve measurement of the pumping speed as a function of the gas composition, gas flow and loading condition of the pump surfaces. The following parameters are subjected to variations: Ar/He ratio, specific helium volume flow rate,more » cryosurface temperature, process gas composition, impurities in argon trapping gas, three-stage operation and two-stage operation. This paper is a description of the experiments on argon trapping techniques started in 1990. Eleven tests as well as the results derived from them are described.« less

Final Technical Report: Electromagnetic Pump Insulation Materials Development and Testing (PLM-DOC-0005-2465) Report # DOEGEHB00613

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

Krahn, John; Reed, Claude; Loewen, Eric

Final Technical Report: Electromagnetic Pump Insulation Materials Development and Testing (Report # DOEGEHB00613) summarizes the information gathered from the analysis of the 160 m3/min EM Pump insulation that was tested in 2000-2002 and additional evaluations of new resilient, engineered insulation system evaluated and tested at both GRC and ANL. This report provides information on Tasks 1 and 2 of the entire project. This report also provides information in three broad areas: Historical and current data; Conclusions based on test data; and Insulation specifications for use in EM Pumps. The research for Task 2 builds upon Task 1: Update EM Pumpmore » Databank, which is summarized within this report. Where research for Task 3 and 4 Next-Generation EM Pump Analysis Tools identified parameters or analysis model that benefits Task 2 research, those items are noted within this report. The important design variables for the manufacture and operation of an EM Pump that the insulation research can evaluate are: space constraints; voltage capability of insulation system; maximum flux density through iron; flow rate and outlet pressure; efficiency and manufacturability. The development summary of the Electromagnetic Pump Insulation Materials Development and Testing was completed to include: Historical and current data; Conclusions based on test data; and Insulation specifications for use in EM Pumps.« less

ETR COMPLEX. CAMERA FACING EAST. FROM LEFT TO RIGHT: ETRCRITICAL ...

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

ETR COMPLEX. CAMERA FACING EAST. FROM LEFT TO RIGHT: ETR-CRITICAL FACILITY BUILDING, ETR CONTROL BUILDING (ATTACHED TO HIGH-BAY ETR), ETR, ONE-STORY SECTION OF ETR BUILDING, ELECTRICAL BUILDING, COOLING TOWER PUMP HOUSE, COOLING TOWER. COMPRESSOR AND HEAT EXCHANGER BUILDING ARE PARTLY IN VIEW ABOVE ETR. DARK-COLORED DUCTS PROCEED FROM GROUND CONNECTION TO ETR WASTE GAS STACK. OTHER STACK IS MTR STACK WITH FAN HOUSE IN FRONT OF IT. RECTANGULAR STRUCTURE NEAR TOP OF VIEW IS SETTLING BASIN. INL NEGATIVE NO. 56-4102. Unknown Photographer, ca. 1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Fast Flux Test Facility thermal and pressure transient events during Cycle 11

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

Ahrens, D. M.

1992-03-01

This report documents the thermal and pressure transients experienced by the Reactor Heat Transport System (RHTS) during Cycle 11 which included Cycles 11A, 11B-1, 11B-2 and 11C (i.e. 4 startups and 4 shutdowns). Cycle 11 consisted of a refueling period that began on March 14, 1989 and power operation which began on May 3, 1989 and ended on October 27, 1990. Transients resulted from secondary pump starts/stops while at refueling conditions. The major causes of transients at power were five unplanned reactor scrams from 100% power and problems with Loop 2 DHX Fan Controls During 11A.

Hydrocarbon polymeric binder for advanced solid propellant

NASA Technical Reports Server (NTRS)

Potts, J. E. (Editor); Ashcraft, A. C., Jr.; Wise, E. W.

1971-01-01

Various experimental factors were examined to determine the source of difficulty in an isoprene polymerization in the 5-gallon reactor which gave a non-uniform product of low functionality. It was concluded that process improvements relating to initiator and monomer purity were desirable, but that the main difficulty was in the initiator feed system. A new pumping system was installed and an analog simulation of the reactor, feed system and initiator decomposition kinetics was devised which permits the selection of initial initiator concentrations and feed rates to use to give a nearly uniform initiator concentration throughout a polymerization run. An isoprene polymerization was run in which the process improvements were implemented.

Large-scale breeder reactor prototype mechanical pump conceptual design study, hot leg

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

Not Available

1976-09-01

Due to the extensive nature of this study, the report is presented as a series of small reports. The complete design analysis is placed in a separate section. The drawings and tabulations are in the back portion of the report. Other topics are enumerated and located as shown in the table of contents.

77 FR 7184 - Entergy Nuclear Indian Point 2, LLC; Entergy Nuclear Operations, Inc.; Indian Point Nuclear...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-10

...-FR- H.1, ``Response To Loss Of Secondary Heat Sink.'' The NRC does not consider implementing 2-FR-H.1 an OMA, as actions to establish reactor coolant system decay heat removal can be performed from the... status trees if Auxiliary Feed necessary to Pump Building. establish alternate secondary heat sink...

15. The 183Filter Plant with settling basins in January 1945. ...

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

15. The 183-Filter Plant with settling basins in January 1945. The 182-B Reservoir and Pump House is on the left in the background, and the coal storage pond for the 184-B Power House is in the upper right. View is to the northwest. P-8012 - B Reactor, Richland, Benton County, WA

2 kW pump-light-stripper-free distributed side-coupled cladding-pumped fiber oscillator

NASA Astrophysics Data System (ADS)

Ying, Hanyuan; Yu, Yu; Cao, Jianqiu; Huang, Zhihe; Pan, Zhiyong; Wang, Zefeng; Chen, Jinbao

2017-06-01

A 2 kW pump-light-stripper-free all-fiber distributed-pumping oscillator fabricated with the distributed side-coupled cladding-pumped Yb-doped fiber is demonstrated for the first time, to the best of our knowledge. An output power of 1969 W with a slope efficiency of 72.2% is obtained. By utilizing the final-section counter-pumping scheme, pure output spectra free from residual pump light are obtained without using any pump light stripper, which demonstrates that the pump light stripper is not indispensable for this configuration. Besides, no stimulated Raman scattering component is observed in the output spectra. The laser has the M 2 factor ranging from 2.0 to 2.6. We believe that the pertinent results are helpful and valuable for designing high-power fiber laser systems.

Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Dickens, Ricky; Houts, Michael; Pearson, Boise; Webster, Kenny; Gibson, Marc; Qualls, Lou; Poston, Dave; Werner, Jim; Radel, Ross

2011-01-01

The Nuclear Systems Team at NASA Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and Mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program, which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter for tests at MSFC. When tested at NASA Glenn Research Center (GRC) the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumentation (temperature, pressure, flow) for data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

NASA Astrophysics Data System (ADS)

Godfroy, T.; Dickens, R.; Houts, M.; Pearson, B.; Webster, K.; Gibson, M.; Qualls, L.; Poston, D.; Werner, J.; Radel, R.

The Nuclear Systems Team at Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter when being tested at MSFC. When tested at GRC the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumenta- tion (temperature, pressure, flow) data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

Design, Fabrication and Integration of a NaK-Cooled Circuit

NASA Technical Reports Server (NTRS)

Garber, Anne; Godfroy, Thomas

2006-01-01

The Early Flight Fission Test Facilities (EFF-TF) team has been tasked by the NASA Marshall Space Flight Center Nuclear Systems Office to design, fabricate, and test an actively pumped alkali metal flow circuit. The system, which was originally designed for use with a eutectic mixture of sodium potassium (NaK), was redesigned to for use with lithium. Due to a shi$ in focus, it is once again being prepared for use with NaK. Changes made to the actively pumped, high temperature circuit include the replacement of the expansion reservoir, addition of remotely operated valves, and modification of the support table. Basic circuit components include: reactor segment, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and a spill reservoir. A 37-pin partial-array core (pin and flow path dimensions are the same as those in a fill design) was selected for fabrication and test. This paper summarizes the integration and preparations for the fill of the pumped liquid metal NaK flow circuit.

Status of Kilowatt-Class Stirling Power Conversion Using a Pumped NaK Loop for Thermal Input

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Geng, Steven M.; Robbie, Malcolm G.

2010-01-01

Free-piston Stirling power conversion has been identified as a viable option for potential Fission Surface Power (FSP) systems on the Moon and Mars. Proposed systems consist of two or more Stirling convertors, in a dual-opposed configuration, coupled to a low-temperature uranium-dioxide-fueled, liquid-metal-cooled reactor. To reduce developmental risks associated with liquid-metal loop integration, a test rig has been built to evaluate the performance of a pair of 1-kW free-piston Stirling convertors using a pumped sodium-potassium (NaK) loop for thermal energy input. Baseline performance maps have been generated at the Glenn Research Center (GRC) for these 1-kW convertors operating with an electric heat source. Each convertor was then retrofitted with a custom-made NaK heater head and integrated into a pumped NaK system at the Marshall Space Flight Center (MSFC). This paper documents baseline testing at GRC as well as the progress made in integrating the Stirling convertors into the pumped NaK loop.

[Design and verification of Luo-Ye pump-based stress formation for cultivation of tissue-engineered blood vessel].

PubMed

Liao, Wen-Jun; Chen, Wan-Wen; Wen, Zhang; Wu, Yue-Heng; Li, Dong-Feng; Zhou, Jia-Hui; Zheng, Jian-Yi; Lin, Zhan-Yi

2016-06-20

To improve Luo-Ye pump-based stress-forming system and optimize the stimulating effect on smooth muscle cells during cultivation of tissue-engineered blood vessels (TEBV). A new Luo-Ye pump-based TEBV 3D culture system was developed by adding an air pump to the output of the bioreactor. A pressure guide wire was used to measure the stress at different points of the silicone tube inside the TEBV bio-reactor, and fitting curves of the stress changes over time was created using Origin 8.0 software. The TEBVs were constructed by seeding vascular smooth muscle cells (VSMCs) isolated from human umbilical artery on polyglycolic acid (PGA) and cultured under dynamic conditions with 40 mmHg resistance (improved group), dynamic conditions without resistance (control group) or static condition (static group) for 4 weeks. The harvested TEBVs were then examined with HE staining, masson staining, α-SMA immunohistochemical staining, and scanning and transmission electron microscopy with semi-quantitative analysis of collagen content and α-SMA expression. The measured stress values and the fitting curves showed that the stress stimuli from the Luo-Ye pump were enhanced by adding an air pump to the output of the bioreactor. Histological analysis revealed improved VSMC density, collagen content and α-SMA expression in the TEBVs constructed with the improved method as compared with those in the control and static groups. Adding an air pump to the Luo-Ye pump significantly enhances the stress stimulation in the TEBV 3-D culture system to promote the secretion function of VSMCs.

76 FR 52715 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-23

... review Draft Final Regulatory Guide (RG) 1.93, ``Availability of Electric Power Sources,'' Revision 1 and new Draft Final RG 1.218, ``Condition Monitoring Techniques for Electric Cables Used in Nuclear Power... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS...

76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-16

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0055] Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for the economic...

Highly Efficient Nd:yag Lasers for Free-space Optical Communications

NASA Technical Reports Server (NTRS)

Sipes, D. L., Jr.

1985-01-01

A highly efficient Nd:YAG laser end-pumped by semiconductor lasers as a possible free-space optical communications source is discussed. Because this concept affords high pumping densities, a long absorption length, and excellent mode-matching characteristics, it is estimated that electrical-to-optical efficiencies greater than 5% could be achieved. Several engineering aspects such as resonator size and configuration, pump collecting optics, and thermal effects are also discussed. Finally, possible methods for combining laser-diode pumps to achieve higher output powers are illustrated.

LLNL Experimental Test Site (Site 300) Potable Water System Operations Plan

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

Ocampo, R. P.; Bellah, W.

The existing Lawrence Livermore National Laboratory (LLNL) Site 300 drinking water system operation schematic is shown in Figures 1 and 2 below. The sources of water are from two Site 300 wells (Well #18 and Well #20) and San Francisco Public Utilities Commission (SFPUC) Hetch-Hetchy water through the Thomas shaft pumping station. Currently, Well #20 with 300 gallons per minute (gpm) pump capacity is the primary source of well water used during the months of September through July, while Well #18 with 225 gpm pump capacity is the source of well water for the month of August. The well watermore » is chlorinated using sodium hypochlorite to provide required residual chlorine throughout Site 300. Well water chlorination is covered in the Lawrence Livermore National Laboratory Experimental Test Site (Site 300) Chlorination Plan (“the Chlorination Plan”; LLNL-TR-642903; current version dated August 2013). The third source of water is the SFPUC Hetch-Hetchy Water System through the Thomas shaft facility with a 150 gpm pump capacity. At the Thomas shaft station the pumped water is treated through SFPUC-owned and operated ultraviolet (UV) reactor disinfection units on its way to Site 300. The Thomas Shaft Hetch- Hetchy water line is connected to the Site 300 water system through the line common to Well pumps #18 and #20 at valve box #1.« less

Lawrence Livermore National Laboratory Experimental Test Site (Site 300) Potable Water System Operations Plan

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

Ocampo, Ruben P.; Bellah, Wendy

The existing Lawrence Livermore National Laboratory (LLNL) Site 300 drinking water system operation schematic is shown in Figures 1 and 2 below. The sources of water are from two Site 300 wells (Well #18 and Well #20) and San Francisco Public Utilities Commission (SFPUC) Hetch-Hetchy water through the Thomas shaft pumping station. Currently, Well #20 with 300 gallons per minute (gpm) pump capacity is the primary source of well water used during the months of September through July, while Well #18 with 225 gpm pump capacity is the source of well water for the month of August. The well watermore » is chlorinated using sodium hypochlorite to provide required residual chlorine throughout Site 300. Well water chlorination is covered in the Lawrence Livermore National Laboratory Experimental Test Site (Site 300) Chlorination Plan (“the Chlorination Plan”; LLNL-TR-642903; current version dated August 2013). The third source of water is the SFPUC Hetch-Hetchy Water System through the Thomas shaft facility with a 150 gpm pump capacity. At the Thomas shaft station the pumped water is treated through SFPUC-owned and operated ultraviolet (UV) reactor disinfection units on its way to Site 300. The Thomas Shaft Hetch- Hetchy water line is connected to the Site 300 water system through the line common to Well pumps #18 and #20 at valve box #1.« less

Convective cooling in a pool-type research reactor

NASA Astrophysics Data System (ADS)

Sipaun, Susan; Usman, Shoaib

2016-01-01

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

DOEpatents

Dickinson, R.W.

1963-03-01

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

The Experimental Breeder Reactor II seismic probabilistic risk assessment

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

Roglans, J; Hill, D J

1994-02-01

The Experimental Breeder Reactor II (EBR-II) is a US Department of Energy (DOE) Category A research reactor located at Argonne National Laboratory (ANL)-West in Idaho. EBR-II is a 62.5 MW-thermal Liquid Metal Reactor (LMR) that started operation in 1964 and it is currently being used as a testbed in the Integral Fast Reactor (IFR) Program. ANL has completed a Level 1 Probabilistic Risk Assessment (PRA) for EBR-II. The Level 1 PRA for internal events and most external events was completed in June 1991. The seismic PRA for EBR-H has recently been completed. The EBR-II reactor building contains the reactor, themore » primary system, and the decay heat removal systems. The reactor vessel, which contains the core, and the primary system, consisting of two primary pumps and an intermediate heat exchanger, are immersed in the sodium-filled primary tank, which is suspended by six hangers from a beam support structure. Three systems or functions in EBR-II were identified as the most significant from the standpoint of risk of seismic-induced fuel damage: (1) the reactor shutdown system, (2) the structural integrity of the passive decay heat removal systems, and (3) the integrity of major structures, like the primary tank containing the reactor that could threaten both the reactivity control and decay heat removal functions. As part of the seismic PRA, efforts were concentrated in studying these three functions or systems. The passive safety response of EBR-II reactor -- both passive reactivity shutdown and passive decay heat removal, demonstrated in a series of tests in 1986 -- was explicitly accounted for in the seismic PRA as it had been included in the internal events assessment.« less

Application of LIF technique for the space- and time-resolved monitoring of pollutant gas decomposition in nonthermal plasma reactors

NASA Astrophysics Data System (ADS)

Mizeraczyk, Jerzy; Ohkubo, Toshikazu; Kanazawa, Seiji; Kocik, Marek

2003-10-01

Laser-induced fluorescence (LIF) technique aided by intensified CCD light signal detection and fast digital image processing is demonstrated to be a useful diagnostic method for in-situ observation of the discharge-induced plasma-chemistry processes responsible for NOx(NO + NO2) decomposition occurring in non-thermal plasma reactors. In this paper a method and results of the LIF measurement of two-dimensional distribution of the ground-state NO molecule density inside a DC positive streamer corona reactor during NO removal from a flue gas simulator [air/NO(up to 300 ppm)] are presented. Either a needle-to-plate or nozzle-to-plate electrode system, having an electrode gap of 30-50 mm was used for generating the corona discharge in the reactor. The LIF monitoring of NO molecules was carried out under the steady-state DC corona discharge condition. The laser-induced fluorescence on the transition NO X2Π(v"=0)<--A2Σ+(v'=0) at λ=226nm was chosen for monitoring ground-state NO molecules in the reactor. This transition was induced by irradiation of the NO molecules with UV laser pulses generated by a laser system consisted of a XeF excimer laser, dye laser and BBO crystal. The laser pulses from the XeF excimer laser (Lambda Physik, Complex 150, λ=351 nm) pumped the dye laser (Lambda Physik, Scanmate) with Coumarin 47 as a dye, which generated the laser beam of a wavelength turned around λ=450 nm. Then, the tuned dye laser beam pumped the BBO crystal in which the second harmonic radiation of a wavelength correspondingly tuned around λ=226 nm was generated. The 226-nm UV laser pulses of energy of 0.8-2 mJ and duration of about 20 ns were transformed into the form of the so-called laser sheet (width of 1 mm, height of 30-50 mm) which passed between the electrodes through the operating gas. The obtained results, presented in the form of images, which illustrated the two-dimensional distributions of NO molecule concentration in the non-thermal reactor, showed that the corona discharge-induced removal of NO molecules occurred not only in the vicinity of the plasma region formed by the corona discharge-induced removal of NO molecules occurred not only in the vicinity of the plasma region formed by the corona streamers and in the downstream region of the reactor but also in the upstream region of the reactor, i.e. before the flue gas simulator has entered the plasma region. This information obtained owing to the LIF technique, is important for the understanding of the plasma-chemistry processes responsible for NOx decomposition in non-thermal plasma reactors and for optimising their performance.

Free-piston Stirling Engine system considerations for various space power applications

NASA Technical Reports Server (NTRS)

Dochat, George R.; Dhar, Manmohan

1991-01-01

Free-Piston Stirling Engines (FPSE) have the potential to provide high reliability, long life, and efficient operation. Therefore, they are excellent candidates for the dynamic power conversion module of a space-based, power-generating system. FPSE can be coupled with many potential heat sources (radioisotope, solar, or nuclear reactor), various heat input systems (pumped loop, heat pipe), heat rejection (pumped loop or heat pipe), and various power management and distribution systems (ac, dc, high or low voltage, and fixed or variable load). This paper reviews potential space missions that can be met using free-piston Stirling engines and discusses options of various system integration approaches. This paper briefly outlines the program and recent progress.

The Dynomak: An advanced spheromak reactor system with imposed-dynamo current drive and next-generation nuclear power technologies

NASA Astrophysics Data System (ADS)

Sutherland, D. A.; Jarboe, T. R.; Marklin, G.; Morgan, K. D.; Nelson, B. A.

2013-10-01

A high-beta spheromak reactor system has been designed with an overnight capital cost that is competitive with conventional power sources. This reactor system utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER developed cryogenic pumping systems were implemented in this design on the basis of technological feasibility. A tritium breeding ratio of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High-temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%. A paper concerning the Dynomak reactor design is currently being reviewed for publication.

Gaseous-fuel nuclear reactor research for multimegawatt power in space

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T.; Helmick, H. H.

1977-01-01

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

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

Oktamuliani, Sri, E-mail: srioktamuliani@ymail.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id

A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation atmore » inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.« less

Nuclear thermal propulsion engine system design analysis code development

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.; Ivanenok, Joseph F.

1992-01-01

A Nuclear Thermal Propulsion (NTP) Engine System Design Analyis Code has recently been developed to characterize key NTP engine system design features. Such a versatile, standalone NTP system performance and engine design code is required to support ongoing and future engine system and vehicle design efforts associated with proposed Space Exploration Initiative (SEI) missions of interest. Key areas of interest in the engine system modeling effort were the reactor, shielding, and inclusion of an engine multi-redundant propellant pump feed system design option. A solid-core nuclear thermal reactor and internal shielding code model was developed to estimate the reactor's thermal-hydraulic and physical parameters based on a prescribed thermal output which was integrated into a state-of-the-art engine system design model. The reactor code module has the capability to model graphite, composite, or carbide fuels. Key output from the model consists of reactor parameters such as thermal power, pressure drop, thermal profile, and heat generation in cooled structures (reflector, shield, and core supports), as well as the engine system parameters such as weight, dimensions, pressures, temperatures, mass flows, and performance. The model's overall analysis methodology and its key assumptions and capabilities are summarized in this paper.

Cooling molten salt reactors using "gas-lift"

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav; Klimko, Marek

2014-08-01

This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a "Two-phase flow demonstrator" (TFD) used for experimental study of the "gas-lift" system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for "gas-lift" (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

NASA Astrophysics Data System (ADS)

Marcille, T. F.; Dixon, D. D.; Fischer, G. A.; Doherty, S. P.; Poston, D. I.; Kapernick, R. J.

2006-01-01

In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

Closure head for a nuclear reactor

DOEpatents

Wade, Elman E.

1980-01-01

A closure head for a nuclear reactor includes a stationary outer ring integral with the reactor vessel with a first rotatable plug disposed within the stationary outer ring and supported from the stationary outer ring by a bearing assembly. A sealing system is associated with the bearing assembly to seal the annulus defined between the first rotatable plug and the stationary outer ring. The sealing system comprises tubular seal elements disposed in the annulus with load springs contacting the tubular seal elements so as to force the tubular seal elements against the annulus in a manner to seal the annulus. The sealing system also comprises a sealing fluid which is pumped through the annulus and over the tubular seal elements causing the load springs to compress thereby reducing the friction between the tubular seal elements and the rotatable components while maintaining a gas-tight seal therebetween.

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.

SEPARATING LIQUID MODERATOR FROM A SLURRY TYPE REACTOR

DOEpatents

Vernon, H.C.

1961-07-01

A system for evaporating moderator such as D/sub 2/O from an irradiated slurry or sloution characterized by two successive evaproators is described. In the first of these the most troublesome radioactivity dissipates before the slurry becomes too thick to be pumped out; in the second the slurry, now easier to handle, can be safely reduced to a sludge.

Reactor Simulator Testing Overview

NASA Technical Reports Server (NTRS)

Schoenfeld, Michael P.

2013-01-01

Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Pearson, J. Boise; Godfroy, Thomas J.

2011-01-01

Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology demonstration unit (TDU). In November, 2011 testing of a 37-pin core simulator (designed in conjunction with Los Alamos National Laboratory) for use with the TDU will occur. Previous testing at the EFFTF has included the thermal and mechanical coupling of a pumped NaK loop to Stirling engines (provided by GRC). Testing related to heat pipe cooled systems, gas cooled systems, heat exchangers, and other technologies has also been performed. Integrated TDU testing will begin at GRC in 2013. Thermal simulators developed at the EFF-TF are capable of operating over the temperature and power range typically of interest to compact reactors. Small and large diameter simulators have been developed, and simulators (coupled with the facility) are able to closely match the axial and radial power profile of all potential systems of interest. A photograph of the TDU core simulator during assembly is provided in Figure 2.

Two-Step Multi-Physics Analysis of an Annular Linear Induction Pump for Fission Power Systems

NASA Technical Reports Server (NTRS)

Geng, Steven M.; Reid, Terry V.

2016-01-01

One of the key technologies associated with fission power systems (FPS) is the annular linear induction pump (ALIP). ALIPs are used to circulate liquid-metal fluid for transporting thermal energy from the nuclear reactor to the power conversion device. ALIPs designed and built to date for FPS project applications have not performed up to expectations. A unique, two-step approach was taken toward the multi-physics examination of an ALIP using ANSYS Maxwell 3D and Fluent. This multi-physics approach was developed so that engineers could investigate design variations that might improve pump performance. Of interest was to determine if simple geometric modifications could be made to the ALIP components with the goal of increasing the Lorentz forces acting on the liquid-metal fluid, which in turn would increase pumping capacity. The multi-physics model first calculates the Lorentz forces acting on the liquid metal fluid in the ALIP annulus. These forces are then used in a computational fluid dynamics simulation as (a) internal boundary conditions and (b) source functions in the momentum equations within the Navier-Stokes equations. The end result of the two-step analysis is a predicted pump pressure rise that can be compared with experimental data.

CONTEXTUAL AERIAL VIEW OF "COLD" NORTH HALF OF MTR COMPLEX. ...

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

CONTEXTUAL AERIAL VIEW OF "COLD" NORTH HALF OF MTR COMPLEX. CAMERA FACING EASTERLY. FOREGROUND CORNER CONTAINS OIL STORAGE TANKS. WATER TANKS AND WELL HOUSES ARE BEYOND THEM TO THE LEFT. LARGE LIGHT-COLORED BUILDING IN CENTER OF VIEW IS STEAM PLANT. DEMINERALIZER AND WATER STORAGE TANK ARE BEYOND. SIX-CELL COOLING TOWER AND ITS PUMP HOUSE ARE ABOVE IT IN VIEW. SERVICE BUILDINGS INCLUDING CANTEEN ARE ON NORTH SIDE OF ROAD. "EXCLUSION" AREA IS BEYOND ROAD. COMPARE LOCATION OF EXCLUSION-AREA GATE WITH PHOTO ID-33-G-202. INL NEGATIVE NO. 3608. Unknown Photographer, 10/30/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Temperature Swing Adsorption Compressor Development

NASA Technical Reports Server (NTRS)

Finn, John E.; Mulloth, Lila M.; Affleck, Dave L.

2001-01-01

Closing the oxygen loop in an air revitalization system based on four-bed molecular sieve and Sabatier reactor technology requires a vacuum pump-compressor that can take the low-pressure CO, from the 4BMS and compress and store for use by a Sabatier reactor. NASA Ames Research Center proposed a solid-state temperature-swing adsorption (TSA) compressor that appears to meet performance requirements, be quiet and reliable, and consume less power than a comparable mechanical compressor/accumulator combination. Under this task, TSA compressor technology is being advanced through development of a complete prototype system. A liquid-cooled TSA compressor has been partially tested, and the rest of the system is being fabricated. An air-cooled TSA compressor is also being designed.

Impact of input FBG reflectivity and forward pump power on RIN transfer in ultralong Raman laser amplifiers.

PubMed

Rizzelli, Giuseppe; Iqbal, Md Asif; Gallazzi, Francesca; Rosa, Paweł; Tan, Mingming; Ania-Castañón, Juan Diego; Krzczanowicz, Lukasz; Corredera, Pedro; Phillips, Ian; Forysiak, Wladek; Harper, Paul

2016-12-12

Relative intensity noise transfer from the pump to the signal in 2nd-order ultra-long Raman laser amplifiers for telecommunications is characterized numerically and experimentally. Our results showcase the need for careful adjustment of the front FBG reflectivity and the relative contribution of forward pump power, and their impact on performance. Finally, our analysis is verified through a 10 × 30 GBaud DP-QPSK transmission experiment, showing a large Q factor penalty associated with the combination of high forward pumping and high reflectivities.

Optimization of end-pumped, actively Q-switched quasi-III-level lasers.

PubMed

Jabczynski, Jan K; Gorajek, Lukasz; Kwiatkowski, Jacek; Kaskow, Mateusz; Zendzian, Waldemar

2011-08-15

The new model of end-pumped quasi-III-level laser considering transient pumping processes, ground-state-depletion and up-conversion effects was developed. The model consists of two parts: pumping stage and Q-switched part, which can be separated in a case of active Q-switching regime. For pumping stage the semi-analytical model was developed, enabling the calculations for final occupation of upper laser level for given pump power and duration, spatial profile of pump beam, length and dopant level of gain medium. For quasi-stationary inversion, the optimization procedure of Q-switching regime based on Lagrange multiplier technique was developed. The new approach for optimization of CW regime of quasi-three-level lasers was developed to optimize the Q-switched lasers operating with high repetition rates. Both methods of optimizations enable calculation of optimal absorbance of gain medium and output losses for given pump rate. © 2011 Optical Society of America

Radiation intensification of the reactor pressure vessels recovery by low temperature heat treatment (wet annealing)

NASA Astrophysics Data System (ADS)

Krasikov, E.

2015-04-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of NPP safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. There are two approaches to annealing. The first one is so-called «dry» high temperature (∼475°C) annealing. It allows obtaining practically complete recovery, but requires the removal of the reactor core and internals. External heat source (furnace) is required to carry out RPV heat treatment. The alternative approach is to anneal RPV at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps while operating within the RPV design limits. This low temperature «wet» annealing, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible.

Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

1976-01-01

The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

NASA Reactor Facility Hazards Summary. Volume 1

NASA Technical Reports Server (NTRS)

1959-01-01

The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2014 CFR

2014-01-01

...), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation... 10 Energy 1 2014-01-01 2014-01-01 false Finality of partial decision on site suitability issues in a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION AGENCY RULES...

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2013 CFR

2013-01-01

...), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation... 10 Energy 1 2013-01-01 2013-01-01 false Finality of partial decision on site suitability issues in a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION AGENCY RULES...

Iowa Hill Pumped Storage Project Investigations - Final Technical Report

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

Hanson, David

2016-07-01

This Final Technical Report is a summary of the activities and outcome of the Department of Energy (DOE) Assistance Agreement DE-EE0005414 with the Sacramento Municipal Utility District (SMUD). The Assistance Agreement was created in 2012 to support investigations into the Iowa Hill Pumped-storage Project (Project), a new development that would add an additional 400 MW of capacity to SMUD’s existing 688MW Upper American River Hydroelectric Project (UARP) in the Sierra Nevada mountains east of Sacramento, California.

Design of the helium cooled lithium lead breeding blanket in CEA: from TBM to DEMO

NASA Astrophysics Data System (ADS)

Aiello, G.; Aubert, J.; Forest, L.; Jaboulay, J.-C.; Li Puma, A.; Boccaccini, L. V.

2017-04-01

The helium cooled lithium lead (HCLL) blanket concept was originally developed in CEA at the beginning of 2000: it is one of the two European blanket concepts to be tested in ITER in the form of a test blanket module (TBM) and one of the four blanket concepts currently being considered for the DEMOnstration reactor that will follow ITER. The TBM is a highly optimized component for the ITER environment that will provide crucial information for the development of the DEMO blanket, but its design needs to be adapted to the DEMO reactor. With respect to the TBM design, reduction of the steel content in the breeding zone (BZ) is sought in order to maximize tritium breeding reactions. Different options are being studied, with the potential of reaching tritium breeding ratio (TBR) values up to 1.21. At the same time, the design of the back supporting structure (BSS), which is a DEMO specific component that has to support the blanket modules inside the vacuum vessel (VV), is ongoing with the aim of maximizing the shielding power and minimizing pumping power. This implies a re-engineering of the modules’ attachment system. Design changes however, will have an impact on the manufacturing and assembly sequences that are being developed for the HCLL-TBM. Due to the differences in joint configurations, thicknesses to be welded, heat dissipation and the various technical constraints related to the accessibility of the welding tools and implementation of non-destructive examination (NDE), the manufacturing procedure should be adapted and optimized for DEMO design. Laser welding instead of TIG could be an option to reduce distortions. The time-of-flight diffraction (TOFD) technique is being investigated for NDE. Finally, essential information expected from the HCLL-TBM program that will be needed to finalize the DEMO design is discussed.

NORTHROP REACTOR. REVISION NO. 1 TO THE FINAL SAFEGUARDS REPORT

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

Duncan, J.M.; Shimizu, B.; Romine, R.A.

1962-10-01

Additions and changes related to the original application concerning construction and operation of the Northrop Reactor are given. Answers to 13 questions relative to the Final Safeguards Report are included. Answers are also included concerning 12 questions regarding receipt, possession, and storage of fuel elements. Other information is included concerning personnel changes and changes from Al-clad fuel elements to improved stainless steelclad hydride elements. It is concluded that the improved Northrop Reactor does not present any undue hazard to the health and safety of the operating personnel or the public. (J.R.D.)

Fukushima Daiichi Nuclear Accident; based on the Final Report of Atomic Energy Society of Japan

NASA Astrophysics Data System (ADS)

Sekimura, Naoto

2014-09-01

The Atomic Energy Society of Japan (AESJ) published the Final Report of the AESJ Investigation Committee on Fukushima Daiichi NPS Accident in March 2014. The AESJ is responsible to identify the underlying root causes of the accident through technical surveys and analyses, and to offer solutions for nuclear safety. At the Fukushima Daiichi, Units 1 to 3, which were under operation, were automatically shut down at 14:46 on March 11, 2011 by the Tohoku District-off the Pacific Ocean Earthquake. About 50 minutes later, the tsunami flooded and destroyed the emergency diesel generators, the seawater cooling pumps, the electric wiring system and the DC power for Units 1, 2 and 4, resulting in loss of all power except for an air-cooled emergency diesel generator at Unit 6. Unit 3 lost all AC power, and later lost DC before dawn of March 13. Cooling the reactors and monitoring the results were heavily dependent on electricity for high-pressure water injection, depressurizing the reactor, low pressure water injection, and following continuous cooling. In Unit 3, for example, recent re-evaluation in August 2014 by TEPCO shows that no cooling water was injected into the reactor core region after 8 PM on March 12, leading to the fuel melting from 5:30 AM on March 13. Even though seawater was injected from fire engines afterwards, the rupture of pressure vessel was caused and the majority of melted fuel dropped into the containment vessel of Unit 3. The estimation of amount of radioactive materials such as Xe-133, I-131, Cs-137 and Cs-134, emitted to the environment from Units 1 to 3 is discussed in the presentation. Direct causes of the accident identified in the AESJ Report were, 1) inadequate tsunami measures, 2) inadequate severe accident management measures and 3) inadequate emergency response, post-accident management/mitigation, and recovery measures. These were caused by the following underlying factors, i.e., a) lack of awareness on the roles and responsibilities by experts, b) shortfalls in establishing safety measures and fostering safety awareness by utilities, c) lack of safety awareness by the regulatory body, d) inadequacies in attitude of learning from efforts and collaborations in the international community, and e) shortage of qualified personnel to ensure safety and inadequacies in organization and management framework.

PMMA/PDMS valves and pumps for disposable microfluidics.

PubMed

Zhang, Wenhua; Lin, Shuichao; Wang, Chunming; Hu, Jia; Li, Cong; Zhuang, Zhixia; Zhou, Yongliang; Mathies, Richard A; Yang, Chaoyong James

2009-11-07

Poly(methyl methacrylate) (PMMA) is gaining in popularity in microfluidic devices because of its low cost, excellent optical transparency, attractive mechanical/chemical properties, and simple fabrication procedures. It has been used to fabricate micromixers, PCR reactors, CE and many other microdevices. Here we present the design, fabrication, characterization and application of pneumatic microvalves and micropumps based on PMMA. Valves and pumps are fabricated by sandwiching a PDMS membrane between PMMA fluidic channel and manifold wafers. Valve closing or opening can be controlled by adjusting the pressure in a displacement chamber on the pneumatic layer via a computer regulated solenoid. The valve provides up to 15.4 microL s(-1) at 60 kPa fluid pressure and seals reliably against forward fluid pressure as high as 60 kPa. A PMMA diaphragm pump can be assembled by simply connecting three valves in series. By varying valve volume or opening time, pumping rates ranging from nL to microL per second can be accurately achieved. The PMMA based valves and pumps were further tested in a disposable automatic nucleic acid extraction microchip to extract DNA from human whole blood. The DNA extraction efficiency was about 25% and the 260 nm/280 nm UV absorption ratio for extracted DNA was 1.72. Because of its advantages of inexpensive, facile fabrication, robust and easy integration, the PMMA valve and pump will find their wide application for fluidic manipulation in portable and disposable microfluidic devices.

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

Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time stepmore » of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

The slightly-enriched spectral shift control reactor. Final report, September 30, 1988--September 30, 1991

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

Martin, W.R.; Lee, J.C.; Larsen, E.W.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

Initial Testing of the Stainless Steel NaK-Cooled Circuit (SNaKC)

NASA Technical Reports Server (NTRS)

Garber, Anne; Godfroy, Thomas

2007-01-01

An actively pumped alkali metal flow circuit, designed and fabricated at the NASA Marshall Space Flight Center, is currently undergoing testing in the Early Flight Fission Test Facility (EFF-TF). Sodium potassium (NaK) was selected as the primary coolant. Basic circuit components include: simulated reactor core, NaK to gas heat exchanger, electromagnetic liquid metal pump, liquid metal flowmeter, load/drain reservoir, expansion reservoir, test section, and instrumentation. Operation of the circuit is based around the 37-pin partial-array core (pin and flow path dimensions are the same as those in a full core), designed to operate at 33 kWt. This presentation addresses the construction, fill and initial testing of the Stainless Steel NaK-Cooled Circuit (SNaKC).

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Technical Reports Server (NTRS)

El-Genk, Mohamed S. (Editor); Hoover, Mark D. (Editor)

1991-01-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects.

Using the sound of nuclear energy

DOE PAGES

Garrett, Steven; Smith, James; Smith, Robert; ...

2016-08-01

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-poweredmore » and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.« less

Using the sound of nuclear energy

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

Garrett, Steven; Smith, James; Smith, Robert

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-poweredmore » and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.« less

Reference reactor module for NASA's lunar surface fission power system

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

Poston, David I; Kapernick, Richard J; Dixon, David D

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on themore » lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.« less

Multi-Megawatt Power System Trade Study

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

Longhurst, Glen Reed; Schnitzler, Bruce Gordon; Parks, Benjamin Travis

2001-11-01

As part of a larger task, the Idaho National Engineering and Environmental Laboratory (INEEL) was tasked to perform a trade study comparing liquid-metal cooled reactors having Rankine power conversion systems with gas-cooled reactors having Brayton power conversion systems. This report summarizes the approach, the methodology, and the results of that trade study. Findings suggest that either approach has the possibility to approach the target specific mass of 3-5 kg/kWe for the power system, though it appears either will require improvements to achieve that. Higher reactor temperatures have the most potential for reducing the specific mass of gas-cooled reactors but domore » not necessarily have a similar effect for liquid-cooled Rankine systems. Fuels development will be the key to higher reactor operating temperatures. Higher temperature turbines will be important for Brayton systems. Both replacing lithium coolant in the primary circuit with gallium and replacing potassium with sodium in the power loop for liquid systems increase system specific mass. Changing the feed pump turbine to an electric motor in Rankine systems has little effect. Key technologies in reducing specific mass are high reactor and radiator operating temperatures, low radiator areal density, and low turbine/generator system masses. Turbine/generator mass tends to dominate overall power system mass for Rankine systems. Radiator mass was dominant for Brayton systems.« less

Convective cooling in a pool-type research reactor

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

Sipaun, Susan, E-mail: susan@nm.gov.my; Usman, Shoaib, E-mail: usmans@mst.edu

2016-01-22

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to passmore » through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.« less

A low-cost municipal sewage treatment system with a combination of UASB and the "fourth-generation" downflow hanging sponge reactors.

PubMed

Tandukar, M; Uemura, S; Machdar, I; Ohashi, A; Harada, H

2005-01-01

This paper presents an evaluation of the process performance of a pilot-scale "fourth generation" downflow hanging sponge (DHS) post-treatment system combined with a UASB pretreatment unit treating municipal wastewater. After the successful operation of the second- and third-generation DHS reactors, the fourth-generation DHS reactor was developed to overcome a few shortcomings of its predecessors. This reactor was designed to further enhance the treatment efficiency and simplify the construction process in real scale, especially for the application in developing countries. Configuration of the reactor was modified to enhance the dissolution of air into the wastewater and to avert the possible clogging of the reactor especially during sudden washout from the UASB reactor. The whole system was operated at a total hydraulic retention time (HRT) of 8 h (UASB: 6 h and DHS: 2 h) for a period of over 600 days. The combined system was able to remove 96% of unfiltered BOD with only 9 mg/L remaining in the final effluent. Likewise, F. coli were removed by 3.45 log with the final count of 10(3) to 10(4) MPN/100 ml. Nutrient removal by the system was also satisfactory.

CFD Design and Analysis of a Passively Suspended Tesla Pump Left Ventricular Assist Device

PubMed Central

Medvitz, Richard B.; Boger, David A.; Izraelev, Valentin; Rosenberg, Gerson; Paterson, Eric G.

2012-01-01

This paper summarizes the use of computational fluid dynamics (CFD) to design a novelly suspended Tesla LVAD. Several design variants were analyzed to study the parameters affecting device performance. CFD was performed at pump speeds of 6500, 6750 and 7000 RPM and at flow rates varying from 3 to 7 liter-per-minute (LPM). The CFD showed that shortening the plates nearest the pump inlet reduced the separations formed beneath the upper plate leading edges and provided a more uniform flow distribution through the rotor gaps, both of which positively affected the device hydrodynamic performance. The final pump design was found to produce a head rise of 77 mmHg with a hydraulic efficiency of 16% at the design conditions of 6 LPM throughflow and a 6750 RPM rotation rate. To assess the device hemodynamics the strain rate fields were evaluated. The wall shear stresses demonstrated that the pump wall shear stresses were likely adequate to inhibit thrombus deposition. Finally, an integrated field hemolysis model was applied to the CFD results to assess the effects of design variation and operating conditions on the device hemolytic performance. PMID:21595722

Hydrodynamic cavitation as a novel approach for pretreatment of oily wastewater for anaerobic co-digestion with waste activated sludge.

PubMed

Habashi, Nima; Mehrdadi, Nasser; Mennerich, Artur; Alighardashi, Abolghasem; Torabian, Ali

2016-07-01

Application of hydrodynamic cavitation (HC) was investigated with the objective of biogas production enhancement from co-digestion of oily wastewater (OWW) and waste activated sludge (WAS). Initially, the effect of HC on the OWW was evaluated in terms of energy consumption and turbidity increase. Then, several mixtures of OWW (with and without HC pretreatment) and WAS with the same concentration of total volatile solid were prepared as a substrate for co-digestion. Following, several batch co-digestion trials were conducted. To compare the biogas production, a number of digestion trials were also conducted with a mono substrate (OWW or WAS alone). The best operating condition of HC was achieved in the shortest retention time (7.5 min) with the application of 3mm diameter orifice and maximum pump rotational speed. Biogas production from all co-digestion reactors was higher than the WAS mono substrate reactors. Moreover, biogas production had a direct relationship with OWW ratio and no major inhibition was observed in any of the reactors. The biogas production was also enhanced by HC pretreatment and almost all of the reactors with HC pretreatment had higher reaction rates than the reactors without pretreatment. Copyright © 2016 Elsevier B.V. All rights reserved.

Thermal Stratification Analysis for Sodium Fast Reactors

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

Schneider, James; Anderson, Mark; Baglietto, Emilio

The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool ismore » desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.« less

The assembly and use of continuous flow systems for chemical synthesis.

PubMed

Britton, Joshua; Jamison, Timothy F

2017-11-01

The adoption of and opportunities in continuous flow synthesis ('flow chemistry') have increased significantly over the past several years. Continuous flow systems provide improved reaction safety and accelerated reaction kinetics, and have synthesised several active pharmaceutical ingredients in automated reconfigurable systems. Although continuous flow platforms are commercially available, systems constructed 'in-lab' provide researchers with a flexible, versatile, and cost-effective alternative. Herein, we describe the assembly and use of a modular continuous flow apparatus from readily available and affordable parts in as little as 30 min. Once assembled, the synthesis of a sulfonamide by reacting 4-chlorobenzenesulfonyl chloride with dibenzylamine in a single reactor coil with an in-line quench is presented. This example reaction offers the opportunity to learn several important skills including reactor construction, charging of a back-pressure regulator, assembly of stainless-steel syringes, assembly of a continuous flow system with multiple junctions, and yield determination. From our extensive experience of single-step and multistep continuous flow synthesis, we also describe solutions to commonly encountered technical problems such as precipitation of solids ('clogging') and reactor failure. Following this protocol, a nonspecialist can assemble a continuous flow system from reactor coils, syringes, pumps, in-line liquid-liquid separators, drying columns, back-pressure regulators, static mixers, and packed-bed reactors.

Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

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

Baek, Joo S.; Diamond, David

A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in themore » analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.« less

Integration of On-Chip Peristaltic Pumps and Injection Valves with Microchip Electrophoresis and Electrochemical Detection

PubMed Central

Bowen, Amanda L; Martin, R. Scott

2010-01-01

A microfluidic approach that integrates peristaltic pumping from an on-chip reservoir with injection valves, microchip electrophoresis and electrochemical detection is described. Fabrication and operation of both the peristaltic pumps and injection valves were optimized to ensure efficient pumping and discrete injections. The final device uses the peristaltic pumps to continuously direct sample from a reservoir containing a mixture of analytes to injection valves that are coupled with microchip electrophoresis and amperometric detection. The separation and direct detection of dopamine and norepinephrine were possible with this approach and the utility of the device was demonstrated by monitoring the stimulated release of these neurotransmitters from a layer of cells introduced into the microchip. It is also shown that this pumping/reservoir approach can be expanded to multiple reservoirs and pumps, where one reservoir can be addressed individually or multiple reservoirs sampled simultaneously. PMID:20665914

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

A novel combined solar pasteurizer/TiO2 continuous-flow reactor for decontamination and disinfection of drinking water.

PubMed

Monteagudo, José María; Durán, Antonio; Martín, Israel San; Acevedo, Alba María

2017-02-01

A new combined solar plant including an annular continuous-flow compound parabolic collector (CPC) reactor and a pasteurization system was designed, built, and tested for simultaneous drinking water disinfection and chemical decontamination. The plant did not use pumps and had no electricity costs. First, water continuously flowed through the CPC reactor and then entered the pasteurizer. The temperature and water flow from the plant effluent were controlled by a thermostatic valve located at the pasteurizer outlet that opened at 80 °C. The pasteurization process was simulated by studying the effect of heat treatment on the death kinetic parameters (D and z values) of Escherichia coli K12 (CECT 4624). 99.1% bacteria photo-inactivation was reached in the TiO 2 -CPC system (0.60 mg cm -2 TiO 2 ), and chemical decontamination in terms of antipyrine degradation increased with increasing residence time in the TiO 2 -CPC system, reaching 70% degradation. The generation of hydroxyl radicals (between 100 and 400 nmol L -1 ) was a key factor in the CPC system efficiency. Total thermal bacteria inactivation was attained after pasteurization in all cases. Chemical degradation and bacterial photo-inactivation in the TiO 2 -CPC system were improved with the addition of 150 mg L -1 of H 2 O 2 , which generated approximately 2000-2300 nmol L -1 of HO ● radicals. Finally, chemical degradation and bacterial photo-inactivation kinetic modelling in the annular CPC photoreactor were evaluated. The effect of the superficial liquid velocity on the overall rate constant was also studied. Both antipyrine degradation and E. coli photo-inactivation were found to be controlled by the catalyst surface reaction rate. Copyright © 2016 Elsevier Ltd. All rights reserved.

Fission Surface Power Systems (FSPS) Project Final Report for the Exploration Technology Development Program (ETDP): Fission Surface Power, Transition Face to Face

NASA Technical Reports Server (NTRS)

Palac, Donald T.

2011-01-01

The Fission Surface Power Systems Project became part of the ETDP on October 1, 2008. Its goal was to demonstrate fission power system technology readiness in an operationally relevant environment, while providing data on fission system characteristics pertinent to the use of a fission power system on planetary surfaces. During fiscal years 08 to 10, the FSPS project activities were dominated by hardware demonstrations of component technologies, to verify their readiness for inclusion in the fission surface power system. These Pathfinders demonstrated multi-kWe Stirling power conversion operating with heat delivered via liquid metal NaK, composite Ti/H2O heat pipe radiator panel operations at 400 K input water temperature, no-moving-part electromagnetic liquid metal pump operation with NaK at flight-like temperatures, and subscale performance of an electric resistance reactor simulator capable of reproducing characteristics of a nuclear reactor for the purpose of system-level testing, and a longer list of component technologies included in the attached report. Based on the successful conclusion of Pathfinder testing, work began in 2010 on design and development of the Technology Demonstration Unit (TDU), a full-scale 1/4 power system-level non-nuclear assembly of a reactor simulator, power conversion, heat rejection, instrumentation and controls, and power management and distribution. The TDU will be developed and fabricated during fiscal years 11 and 12, culminating in initial testing with water cooling replacing the heat rejection system in 2012, and complete testing of the full TDU by the end of 2014. Due to its importance for Mars exploration, potential applicability to missions preceding Mars missions, and readiness for an early system-level demonstration, the Enabling Technology Development and Demonstration program is currently planning to continue the project as the Fission Power Systems project, including emphasis on the TDU completion and testing.

Note: A dual temperature closed loop batch reactor for determining the partitioning of trace gases within CO2-water systems.

PubMed

Warr, Oliver; Rochelle, Christopher A; Masters, Andrew J; Ballentine, Christopher J

2016-01-01

An experimental approach is presented which can be used to determine partitioning of trace gases within CO2-water systems. The key advantages of this system are (1) The system can be isolated with no external exchange, making it ideal for experiments with conservative tracers. (2) Both phases can be sampled concurrently to give an accurate composition at each phase at any given time. (3) Use of a lower temperature flow loop outside of the reactor removes contamination and facilitates sampling. (4) Rapid equilibration at given pressure/temperature conditions is significantly aided by stirring and circulating the water phase using a magnetic stirrer and high-pressure liquid chromatography pump, respectively.

Fusion pumped laser

DOEpatents

Pappas, D.S.

1987-07-31

The apparatus of this invention may comprise a system for generating laser radiation from a high-energy neutron source. The neutron source is a tokamak fusion reactor generating a long pulse of high-energy neutrons and having a temperature and magnetic field effective to generate a neutron flux of at least 10/sup 15/ neutrons/cm/sup 2//center dot/s. Conversion means are provided adjacent the fusion reactor at a location operable for converting the high-energy neutrons to an energy source with an intensity and energy effective to excite a preselected lasing medium. A lasing medium is spaced about and responsive to the energy source to generate a population inversion effective to support laser oscillations for generating output radiation. 2 figs., 2 tabs.

Neutral Beam Development for the Lockheed Martin Compact Fusion Reactor

NASA Astrophysics Data System (ADS)

Ebersohn, Frans; Sullivan, Regina

2017-10-01

The Compact Fusion Reactor project at Lockheed Martin Skunk Works is developing a neutral beam injection system for plasma heating. The neutral beam plasma source consists of a high current lanthanum hexaboride (LaB6) hollow cathode which drives an azimuthal cusp discharge similar to gridded ion thrusters. The beam is extracted with a set of focusing grids and is then neutralized in a chamber pumped with Titanium gettering. The design, testing, and analyses of individual components are presented along with the most current full system results. The goal of this project is to advance in-house neutral beam expertise at Lockheed Martin to aid in operation, procurement, and development of neutral beam technology. ©2017 Lockheed Martin Corporation. All Rights Reserved.

Automotive Test Rig Final Design Report. Volume 2. Control System.

DTIC Science & Technology

1986-01-01

Pressure Switch Status P27 Low Brake Release Pressure Switch Status P26 Low Brake...Supply Pressure Switch Status P25 Low Port Charge Pump Pressure Switch Status P24 Low Starboard Charge Pump Pressure Switch Status P23 Hydraulic Filter By...Sensed Switch Status P31 Low Scavenge Pump Pressure Switch Status P30 P37 Signal Return for Computer J21 Not Used J22 P A +24 B Pwr Rtn C Ground C

Practical demonstration of heat pumps for utilization of animal-generated heat

NASA Astrophysics Data System (ADS)

Amberg, H. U.

1980-09-01

Airconditioning of pigpens to eliminate effects of temperature extremes is reported. A stall air conditioner was installed as heat pump in a pigpen for final fattening. The heat, recovered from the exhaust air, is supplied to the outside air so that heated fresh air is blown into the stall. The test was accomplished on a farm with intensive pig breeding with 120 preliminary fattening places and 240 final fattening places. The stall air conditioner offers the possibility to attenuate the extreme temperature variations during the year.

Optimization and Analysis of Centrifugal Pump considering Fluid-Structure Interaction

PubMed Central

Hu, Sanbao

2014-01-01

This paper presents the optimization of vibrations of centrifugal pump considering fluid-structure interaction (FSI). A set of centrifugal pumps with various blade shapes were studied using FSI method, in order to investigate the transient vibration performance. The Kriging model, based on the results of the FSI simulations, was established to approximate the relationship between the geometrical parameters of pump impeller and the root mean square (RMS) values of the displacement response at the pump bearing block. Hence, multi-island genetic algorithm (MIGA) has been implemented to minimize the RMS value of the impeller displacement. A prototype of centrifugal pump has been manufactured and an experimental validation of the optimization results has been carried out. The comparison among results of Kriging surrogate model, FSI simulation, and experimental test showed a good consistency of the three approaches. Finally, the transient mechanical behavior of pump impeller has been investigated using FSI method based on the optimized geometry parameters of pump impeller. PMID:25197690

Geothermal Heat Pumps Score High Marks in Schools.

ERIC Educational Resources Information Center

National Renewable Energy Lab (DOE).

Geothermal heat pumps (GHPs) are showing their value in providing lower operating and maintenance costs, energy efficiency, and superior classroom comfort. This document describes what GHPs are and the benefits a school can garner after installing a GHP system. Three case studies are provided that illustrate these benefits. Finally, the Department…

46 CFR 112.15-5 - Final emergency loads.

Code of Federal Regulations, 2013 CFR

2013-10-01

... or gas turbines that drive emergency generators; and (3) General alarm batteries. (d) One of the... of the subchapter under which the vessel is certificated. (f) Each sprinkler system, water spray extinguishing system, or foam system pump. (g) If necessary, the lube oil pump for each propulsion turbine and...

46 CFR 112.15-5 - Final emergency loads.

Code of Federal Regulations, 2014 CFR

2014-10-01

... or gas turbines that drive emergency generators; and (3) General alarm batteries. (d) One of the... of the subchapter under which the vessel is certificated. (f) Each sprinkler system, water spray extinguishing system, or foam system pump. (g) If necessary, the lube oil pump for each propulsion turbine and...

46 CFR 112.15-5 - Final emergency loads.

Code of Federal Regulations, 2012 CFR

2012-10-01

... or gas turbines that drive emergency generators; and (3) General alarm batteries. (d) One of the... of the subchapter under which the vessel is certificated. (f) Each sprinkler system, water spray extinguishing system, or foam system pump. (g) If necessary, the lube oil pump for each propulsion turbine and...

Stepwise pumping approach to improve free phase light hydrocarbon recovery from unconfined aquifers

NASA Astrophysics Data System (ADS)

Cooper, Grant S.; Peralta, Richard C.; Kaluarachchi, Jagath J.

1995-04-01

A stepwise, time-varying pumping approach is developed to improve free phase oil recovery of light non-aqueous phase liquids (LNAPL) from a homogeneous, unconfined aquifer. Stepwise pumping is used to contain the floating oil plume and obtain efficient free oil recovery. The graphical plots. The approach uses ARMOS ©, an areal two-dimensional multiphase flow, finite-element simulation model. Systematic simulations of free oil area changes to pumping rates are analyzed. Pumping rates are determined that achieve LNAPL plume containment at different times (i.e. 90, 180 and 360 days) for a planning period of 360 days. These pumping rates are used in reverse order as a stepwise (monotonically increasing) pumping strategy. This stepwise pumping strategy is analyzed further by performing additional simulations at different pumping rates for the last pumping period. The final stepwise pumping strategy is varied by factors of -25% and +30% to evaluate sensitivity in the free oil recovery process. Stepwise pumping is compared to steady pumping rates to determine the best free oil recovery strategy. Stepwise pumping is shown to improve oil recovery by increasing recoveredoil volume (11%) and decreasing residual oil (15%) when compared with traditional steady pumping strategies. The best stepwise pumping strategy recovers more free oil by reducing the amount of residual oil left in the system due to pumping drawdown. This stepwise pumping pproach can be used to enhance free oil recovery and provide for cost-effective design and management of LNAPL cleanup.

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE PAGES

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; ...

2016-12-21

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

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

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

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

Cirrito, A.J.

Combustion jet pumps ingest waste heat gases from power plant engines and boilers to boost their pressure for the ultimate low temperature utilization of the captured heat for heating homes, full-year hot houses, sterilization purposes, recreational hot water, absorption refrigeration and the like. Jet pump energy is sustained from the incineration of solids, liquids and gases and vapors or simply from burning fuels. This is the energy needed to transport the reaction products to the point of heat utilization and to optimize the heat transfer to that point. Sequent jet pumps raise and preserve energy levels. Crypto-steady and special jetmore » pumps increase pumping efficiency. The distribution conduit accepts fluidized solids, liquids, gases and vapors in multiphase flow. Temperature modulation and flow augmentation takes place by water injection. Macro solids such as dried sewage waste are removed by cyclone separation. Micro particles remain entrained and pass out with waste condensate just beyond each point of final heat utilization to recharge the water table. The non-condensible gases separated at this point are treated for pollution control. Further, jet pump reactions are controlled to yield fuel gas as necessary to power jet pumps or other use. In all these effects introduced sequentially, the available energy necessary to provide the flow energy, for the continuously distributed heating medium, is first extracted from fuel and fuel-like additions to the stream. As all energy, any way, finally converts to heat, which in this case is retained or recaptured in the flow, the captured heat is practically 90% available at the point of low temperature utilization. The jet pump for coal gasification is also disclosed as are examples of coal gasification and hydrogen production.« less

SNPSAM - Space Nuclear Power System Analysis Model

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Seo, Jong T.

The current version of SNPSAM is described, and the results of the integrated thermoeletric SP-100 system performance studies using SNPSAM are reported. The electric power output, conversion efficiency, coolant temperatures, and specific pumping power of the system are calculated as functions of the reactor thermal power and the liquid metal coolant type (Li or NaK-78) during steady state operation. The transient behavior of the system is also discussed.

Diesel-fired self-pumping water heater

NASA Astrophysics Data System (ADS)

Gertsmann, Joseph

1994-07-01

The object of this project was to study the feasibility of pumping and heating water by sustained oscillatory vaporization and condensation in a fired heat exchanger. Portable field liquid fueled water heaters would facilitate heating water for sanitation, personal hygiene, food service, laundry, equipment maintenance, and decontamination presently available only from larger, less portable, motorized pumping units. The technical tasks consisted of: development of an analytical model, operation of proof-of-principal prototypes, and determination of the thermal and mechanical relationships to evaluate operating range and control characteristics. Four successive pump models were analyzed and tested. The final analytical model gave reasonable agreement with the experimental results, indicating that the actual pumping effect was an order of magnitude lower than originally anticipated. It was concluded that a thermally-activated self pumping water heater based on the proposed principle is not feasible.

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

Time-gated real-time pump-probe imaging spectroscopy

NASA Astrophysics Data System (ADS)

Ferrari, Raffaele; D'Andrea, Cosimo; Bassi, Andrea; Valentini, Gianluca; Cubeddu, Rinaldo

2007-07-01

An experimental technique which allows one to perform pump-probe transient absorption spectroscopy in real-time is an important tool to study irreversible processes. This is particularly interesting in the case of biological samples which easily deteriorate upon exposure to light pulses, with the formation of permanent photoproducts and structural changes. In particular pump-probe spectroscopy can provide fundamental information for the design of optical chromophores. In this work a real-time pump-probe imaging spectroscopy system has been realized and we have explored the possibility to further reduce the number of laser pulses by using a time-gated camera. We believe that the use of a time-gated camera can provide an important step towards the final goal of pump-probe single shot spectroscopy.

Online Oxide Contamination Measurement and Purification Demonstration

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Godfroy, T. J.; Webster, K. L.; Garber, A. E.; Polzin, K. A.; Childers, D. J.

2011-01-01

Liquid metal sodium-potassium (NaK) has advantageous thermodynamic properties indicating its use as a fission reactor coolant for a surface (lunar, martian) power system. A major area of concern for fission reactor cooling systems is system corrosion due to oxygen contaminants at the high operating temperatures experienced. A small-scale, approximately 4-L capacity, simulated fission reactor cooling system employing NaK as a coolant was fabricated and tested with the goal of demonstrating a noninvasive oxygen detection and purification system. In order to generate prototypical conditions in the simulated cooling system, several system components were designed, fabricated, and tested. These major components were a fully-sealed, magnetically-coupled mechanical NaK pump, a graphite element heated reservoir, a plugging indicator system, and a cold trap. All system components were successfully demonstrated at a maximum system flow rate of approximately 150 cc/s at temperatures up to 550 C. Coolant purification was accomplished using a cold trap before and after plugging operations which showed a relative reduction in oxygen content.

NEUTRONIC REACTOR OPERATIONAL METHOD AND CORE SYSTEM

DOEpatents

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

1960-07-19

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

Critical analysis of submerged membrane sequencing batch reactor operating conditions.

PubMed

McAdam, Ewan; Judd, Simon J; Gildemeister, René; Drews, Anja; Kraume, Matthias

2005-10-01

To evaluate the Submerged Membrane Sequencing Batch Reactor process, several short-term studies were conducted to define critical flux, membrane aeration and intermittent filtration operation. Critical flux trials indicated that as mixed liquor suspended solids increased in concentration so would the propensity for membrane fouling. Consequently in order to characterise the impact of biomass concentration increase (that develops during permeate withdrawal) upon submerged microfiltration operation, two longer term studies were conducted, one with a falling hydraulic head and another with a continuous hydraulic head (as in membrane bio-reactors). Trans membrane pressure data was used to predict the maximum possible operating periods at 10 and 62 days for the falling hydraulic head and continuous hydraulic head respectively. Further analysis revealed that falling hydraulic head operation would require 21% more aeration to maintain a consistent crossflow velocity than continuous operation and would rely on pumping for full permeate withdrawal 80% earlier. This study concluded that further optimisation would be required to make this technology technically and economically viable.

RELAP5 Application to Accident Analysis of the NIST Research Reactor

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

Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.

Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accidentmore » and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.« less

Final Determination - signed August 31, 2008

EPA Pesticide Factsheets

Final Determination of the U.S. EPA’s Assistant Administrator for Water pursuant to Section 404(c) of the Clean Water Act concerning the proposed Yazoo Backwater area pumps project, Issaquena county, Mississippi.

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

Natesan, K.; Momozaki, Y.; Li, M.

This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory,more » the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.' Two reports were issued in 2009 (Natesan and Meimei Li 2009, Natesan et al. 2009) which examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design specifications for the ANL pumped loop for testing advanced structural materials. Available information was presented on solubility of several metallic and nonmetallic elements along with a discussion of the possible mechanisms for the accumulation of impurities in sodium. That report concluded that the solubility of many metals in sodium is low (<1 part per million) in the temperature range of interest in sodium reactors and such trace amounts would not impact the mechanical integrity of structural materials and components. The earlier report also analyzed the solubility and transport mechanisms of nonmetallic elements such as oxygen, nitrogen, carbon, and hydrogen in laboratory sodium loops and in reactor systems such as Experimental Breeder Reactor-II, Fast Flux Test Facility, and Clinch River Breeder Reactor. Among the nonmetallic elements discussed, oxygen is deemed controllable and its concentration in sodium can be maintained in sodium for long reactor life by using cold-trap method. It was concluded that among the cold-trap and getter-trap methods, the use of cold trap is sufficient to achieve oxygen concentration of the order of 1 part per million. Under these oxygen conditions in sodium, the corrosion performance of structural materials such as austenitic stainless steels and ferritic steels will be acceptable at a maximum core outlet sodium temperature of {approx}550 C. In the current sodium compatibility studies, the oxygen concentration in sodium will be controlled and maintained at {approx}1 ppm by controlling the cold trap temperature. The oxygen concentration in sodium in the forced convection sodium loop will be controlled and monitored by maintaining the cold trap temperature in the range of 120-150 C, which would result in oxygen concentration in the range of 1-2 ppm. Uniaxial tensile specimens are being exposed to flowing sodium and will be retrieved and analyzed for corrosion and post-exposure tensile properties. Advanced materials for sodium exposure include austenitic alloy HT-UPS and ferritic-martensitic steels modified 9Cr-1Mo and NF616. Among the nonmetallic elements in sodium, carbon was assessed to have the most influence on structural materials since carbon, as an impurity, is not amenable to control and maintenance by any of the simple purification methods. The dynamic equilibrium value for carbon in sodium systems is dependent on several factors, details of which were discussed in the earlier report. The current sodium compatibility studies will examine the role of carbon concentration in sodium on the carburization-decarburization of advanced structural materials at temperatures up to 650 C. Carbon will be added to the sodium by exposure of carbon-filled iron tubes, which over time will enable carbon to diffuse through iron and dissolve into sodium. The method enables addition of dissolved carbon (without carbon particulates) in sodium that is of interest for materials compatibility evaluation. The removal of carbon from the sodium will be accomplished by exposing carbon-gettering alloys such as refractory metals that have a high partitioning coefficient for carbon and also precipitate carbides, thereby decreasing the carbon concentration in sodium.« less

Current randomized control trials, observational studies and meta analysis in off-pump coronary surgery.

PubMed

Parissis, Haralabos; Lau, Man Chi; Parissis, Mondrian; Lampridis, Savvas; Graham, Victoria; Al-Saudi, Reza; Mhandu, Peter

2015-12-17

The off-pump literature is divided into three eras: the "early phase" with results favouring off-pump surgery supported with randomized control trials (RCTs) mainly from Bristol, UK; an "intermediate phase" dominated by the results of the ROOBY trial and finally a more "contemporary phase" whereby the off/on-pump argument is unsettled. Although the literature has failed to project an overall superiority of off-pump versus on-pump surgery, nevertheless, small randomized control trials and large meta-analysis studies are concluding that the incidence of a stroke is less than 1 % when an aortic off-pump techniques (especially the non-touch technique) are advocated in patients with diseased ascending aorta. Furthermore, off-pump combined with hybrid procedures may lead to a reduction of adverse outcome in the aged high-risk population with concomitant poor left ventricular function and co-morbidities.The current review attempts to bring an insight onto the last ten years knowledge on the on/off-pump debate, with an aim to draw some clear conclusions in order to allow practitioners to reflect on the subject.

A portable pressure pump for microfluidic lab-on-a-chip systems using a porous polydimethylsiloxane (PDMS) sponge.

PubMed

Cha, Kyoung Je; Kim, Dong Sung

2011-10-01

In this paper, we propose a novel portable and disposable pressure pump using a porous polydimethylsiloxane (PDMS) sponge and demonstrate its application to a microfluidic lab-on-a-chip. The porous PDMS sponge was simply fabricated by a sugar leaching technique based on capillary suction of pre-cured PDMS into lumps of sugar, thereby enabling us to achieve the porous PDMS sponge composed of interconnected micropores. To indicate the characteristics of the porous PDMS sponge and pump, we measured the average porosities of them whose values were 0.64 and 0.34, respectively. A stress-strain relationship of the fabricated portable pressure pump represented a linear behavior in the compressive strain range of 0 to 20%. Within this range, a pumping volume of the pressure pump could be linearly controlled by the compressed strain. Finally, the fabricated porous PDMS pump was successfully demonstrated as a portable pressure pump for a disposable microfluidic lab-on-a-chip for efficient detection of agglutination. The proposed portable pressure pump can be potentially applicable to various disposable microfluidic lab-on-a-chip systems.

New methods for the development of pneumatic displacement pumps for cardiac assist.

PubMed

Knierbein, B; Rosarius, N; Reul, H; Rau, G

1990-11-01

The primary goal of the presented project was to develop a pump family with stroke volumes of 20, 50, 70 and 90 ml, which could be produced at low cost but with sufficient quality. The housing parts of the pump were thermoformed from technical semifinished materials. All blood contacting surfaces of the pump were coated with biomaterials in a controlled dipping process. During the design and fabrication process a professional CAD-system was used. This facilitated spatial presentations of pump components for first evaluations at the initial draft stages. The CAD-design data were then transformed to CNC-controlled lathes and mill's for the fabrication of pump tools. The stresses and strains of the moving blood pump components, such as membranes and valves, were precalculated by means of Finite-Element-Analysis (FEM). After completion of the pump, the internal flow fields were investigated by flow-visualization techniques using non-Newtonian test fluids, and the pump characteristics (function curves) were investigated in appropriate circulatory mock loops. The paper covers all above aspects from first draft to final fabrication and testing.

Heat Pump Drying of Fruits and Vegetables: Principles and Potentials for Sub-Saharan Africa

PubMed Central

Fayose, Folasayo; Huan, Zhongjie

2016-01-01

Heat pump technology has been used for heating, ventilation, and air-conditioning in domestic and industrial sectors in most developed countries of the world including South Africa. However, heat pump drying (HPD) of fruits and vegetables has been largely unexploited in South Africa and by extension to the sub-Saharan African region. Although studies on heat pump drying started in South Africa several years ago, not much progress has been recorded to date. Many potential users view heat pump drying technology as fragile, slow, and high capital intensive when compared with conventional dryer. This paper tried to divulge the principles and potentials of heat pump drying technology and the conditions for its optimum use. Also, various methods of quantifying performances during heat pump drying as well as the quality of the dried products are highlighted. Necessary factors for maximizing the capacity and efficiency of a heat pump dryer were identified. Finally, the erroneous view that heat pump drying is not feasible economically in sub-Saharan Africa was clarified. PMID:26904668

A Bovine Hemoglobin-Based Oxygen Carrier as Pump Prime for Cardiopulmonary Bypass: Reduced Systemic Lactic Acidosis and Improved Cerebral Oxygen Metabolism During Low-flow in a Porcine Model

DTIC Science & Technology

2010-11-10

1 A bovine hemoglobin-based oxygen carrier as pump prime for cardiopulmonary bypass: reduced systemic lactic acidosis and improved cerebral...2010 2. REPORT TYPE Final Report 3. DATES COVERED (From - To) June 2007 - November 2010 4. TITLE AND SUBTITLE A bovine hemoglobin-based oxygen...carrier as pump prime for cardiopulmonary bypass: reduced systemic lactic acidosis and improved cerebral oxygen metabolism during low-flow in a

Investigation of Enersave series 500 pump. Final report

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

Smith, P.R.

A new type of pump to be used as a stripper pump for oil wells has been developed by Enersave Pumps, Incorporated of Roswell, New Mexico. The Enersave 500 pump has no moving mechanical parts between the down-hole pistons which lift the fluid and the driving unit at the surface. Rather, a pressure pulse created by the driving unit, usually called the pulser, is transmitted through the fluid in the well string to the down-hole unit and creates the pumping action. Object of the project was to optimize the configuration of the pump, that is, increase the production flow ratemore » while minimizing the energy consumption needed to obtain this flow rate. New Mexico State University's role in this project was to model the pump using computer techniques to provide guidelines for improvement in pump design, to supervise the performance of field and bench testing of the redesigned versions of the pump to validate the actual performance of the pump, and to provide a life cycle cost analysis of the pump. Experimental results at depths to as much as 1729 feet show that the redesigned pump will deliver 3 gpm with an average power input of about 1 hp. The energy requirements of the Enersave 500 pump are on the average 25% lower than the energy requirements of an equivalent pump-jack, the typical pump now used in the oil fields for stripper well operation. Further, a life cycle cost analysis of the Enersave 500 pump compared to an equivalent pump-jack shows the Enersave 500 pump to be more economical to purchase and operate.« less

A "place n play" modular pump for portable microfluidic applications.

PubMed

Li, Gang; Luo, Yahui; Chen, Qiang; Liao, Lingying; Zhao, Jianlong

2012-03-01

This paper presents an easy-to-use, power-free, and modular pump for portable microfluidic applications. The pump module is a degassed particle desorption polydimethylsiloxane (PDMS) slab with an integrated mesh-shaped chamber, which can be attached on the outlet port of microfluidic device to absorb the air in the microfluidic system and then to create a negative pressure for driving fluid. Different from the existing monolithic degassed PDMS pumps that are generally restricted to limited pumping capacity and are only compatible with PDMS-based microfluidic devices, this pump can offer various possible configures of pumping power by varying the geometries of the pump or by combining different pump modules and can also be employed in any material microfluidic devices. The key advantage of this pump is that its operation only requires the user to place the degassed PDMS slab on the outlet ports of microfluidic devices. To help design pumps with a suitable pumping performance, the effect of pump module geometry on its pumping capacity is also investigated. The results indicate that the performance of the degassed PDMS pump is strongly dependent on the surface area of the pump chamber, the exposure area and the volume of the PDMS pump slab. In addition, the initial volume of air in the closed microfluidic system and the cross-linking degree of PDMS also affect the performance of the degassed PDMS pump. Finally, we demonstrated the utility of this modular pumping method by applying it to a glass-based microfluidic device and a PDMS-based protein crystallization microfluidic device.

A “place n play” modular pump for portable microfluidic applications

PubMed Central

Li, Gang; Luo, Yahui; Chen, Qiang; Liao, Lingying; Zhao, Jianlong

2012-01-01

This paper presents an easy-to-use, power-free, and modular pump for portable microfluidic applications. The pump module is a degassed particle desorption polydimethylsiloxane (PDMS) slab with an integrated mesh-shaped chamber, which can be attached on the outlet port of microfluidic device to absorb the air in the microfluidic system and then to create a negative pressure for driving fluid. Different from the existing monolithic degassed PDMS pumps that are generally restricted to limited pumping capacity and are only compatible with PDMS-based microfluidic devices, this pump can offer various possible configures of pumping power by varying the geometries of the pump or by combining different pump modules and can also be employed in any material microfluidic devices. The key advantage of this pump is that its operation only requires the user to place the degassed PDMS slab on the outlet ports of microfluidic devices. To help design pumps with a suitable pumping performance, the effect of pump module geometry on its pumping capacity is also investigated. The results indicate that the performance of the degassed PDMS pump is strongly dependent on the surface area of the pump chamber, the exposure area and the volume of the PDMS pump slab. In addition, the initial volume of air in the closed microfluidic system and the cross-linking degree of PDMS also affect the performance of the degassed PDMS pump. Finally, we demonstrated the utility of this modular pumping method by applying it to a glass-based microfluidic device and a PDMS-based protein crystallization microfluidic device. PMID:22685507

Core Design Characteristics of the Fluoride Salt-Cooled High Temperature Demonstration Reactor

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

Brown, Nicholas R; Qualls, A L; Betzler, Benjamin R

2016-01-01

Fluoride salt-cooled high temperature reactors (FHRs) are a promising reactor technology option with significant knowledge gaps to implementation. One potential approach to address those technology gaps is via a small-scale demonstration reactor with the goal of increasing the technology readiness level (TRL) of the overall system for the longer term. The objective of this paper is to outline a notional concept for such a system, and to address how the proposed concept would advance the TRL of FHR concepts. Development of the proposed FHR Demonstration Reactor (DR) will enable commercial FHR deployment through disruptive and rapid technology development and demonstration.more » The FHR DR will close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. Important capabilities that will be demonstrated by building and operating the FHR DR include core design methodologies; fabrication and operation of high temperature reactors; salt procurement, handling, maintenance, and ultimate disposal; salt chemistry control to maximize vessel life; tritium management; heat exchanger performance; pump performance; and reactivity control. The FHR DR is considered part of a broader set of FHR technology development and demonstration efforts, some of which are already underway. Nonreactor test efforts (e.g., heated salt loops or loops using simulant fluids) can demonstrate many technologies necessary for commercial deployment of FHRs. The FHR DR, however, fulfills a crucial role in FHR technology development by advancing the technical maturity and readiness level of the system as a whole.« 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

Advances in Valveless Piezoelectric Pump with Cone-shaped Tubes

NASA Astrophysics Data System (ADS)

Zhang, Jian-Hui; Wang, Ying; Huang, Jun

2017-07-01

This paper reviews the development of valveless piezoelectric pump with cone-shaped tube chronologically, which have widely potential application in biomedicine and micro-electro-mechanical systems because of its novel principles and deduces the research direction in the future. Firstly, the history of valveless piezoelectric pumps with cone-shaped tubes is reviewed and these pumps are classified into the following types: single pump with solid structure or plane structure, and combined pump with parallel structure or series structure. Furthermore, the function of each type of cone-shaped tubes and pump structures are analyzed, and new directions of potential expansion of valveless piezoelectric pumps with cone-shaped tubes are summarized and deduced. The historical argument, which is provided by the literatures, that for a valveless piezoelectric pump with cone-shaped tubes, cone angle determines the flow resistance and the flow resistance determines the flow direction. The argument is discussed in the reviewed pumps one by one, and proved to be convincing. Finally, it is deduced that bionics is pivotal in the development of valveless piezoelectric pump with cone-shaped tubes from the perspective of evolution of biological structure. This paper summarizes the current valveless piezoelectric pumps with cone-shaped tubes and points out the future development, which may provide guidance for the research of piezoelectric actuators.

Overview of NASA Lewis Research Center free-piston Stirling engine technology activities applicable to space power systems

NASA Technical Reports Server (NTRS)

Slaby, Jack G.

1987-01-01

A brief overview is presented of the development and technological activities of the free-piston Stirling engine. The engine started as a small scale fractional horsepower engine which demonstrated basic engine operating principles and the advantages of being hermetically sealed, highly efficient, and simple. It eventually developed into the free piston Stirling engine driven heat pump, and then into the SP-100 Space Reactor Power Program from which came the Space Power Demonstrator Engine (SPDE). The SPDE successfully operated for over 300 hr and delivered 20 kW of PV power to an alternator plunger. The SPDE demonstrated that a dynamic power conversion system can, with proper design, be balanced; and the engine performed well with externally pumped hydrostatic gas bearings.

Lithium Circuit Test Section Design and Fabrication

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Garber, Anne

2006-01-01

The Early Flight Fission - Test Facilities (EFF-TF) team has designed and built an actively pumped lithium flow circuit. Modifications were made to a circuit originally designed for NaK to enable the use of lithium that included application specific instrumentation and hardware. Component scale freeze/thaw tests were conducted to both gain experience with handling and behavior of lithium in solid and liquid form and to supply anchor data for a Generalized Fluid System Simulation Program (GFSSP) model that was modified to include the physics for freeze/thaw transitions. Void formation was investigated. The basic circuit components include: reactor segment, lithium to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and trace heaters. This paper will discuss the overall system design and build and the component testing findings.

Lithium Circuit Test Section Design and Fabrication

NASA Astrophysics Data System (ADS)

Godfroy, Thomas; Garber, Anne; Martin, James

2006-01-01

The Early Flight Fission - Test Facilities (EFF-TF) team has designed and built an actively pumped lithium flow circuit. Modifications were made to a circuit originally designed for NaK to enable the use of lithium that included application specific instrumentation and hardware. Component scale freeze/thaw tests were conducted to both gain experience with handling and behavior of lithium in solid and liquid form and to supply anchor data for a Generalized Fluid System Simulation Program (GFSSP) model that was modified to include the physics for freeze/thaw transitions. Void formation was investigated. The basic circuit components include: reactor segment, lithium to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and trace heaters. This paper discusses the overall system design and build and the component testing findings.

RAMONA-3B application to Browns Ferry ATWS

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

Slovik, G.C.; Neymotin, L.; Cazzoli, E.

1984-01-01

This paper discusses two preliminary MSIV clsoure ATWS calculations done using the RAMONA-3B code and the work being done to create the necessary cross section sets for the Browns Ferry Unit 1 reactor. The RAMONA-3B code employs a three-dimensional neutron kinetics model coupled with one-dimensional, four equation, nonhomogeneous, nonequilibrium thermal hydraulics. To be compatible with 3-D neutron kinetics, the code uses parallel coolant channels in the core. It also includes a boron transport model and all necessary BWR components such as jet pump, recirculation pump, steam separator, steamline with safety and relief valves, main steam isolation valve, turbine stop valve,more » and turbine bypass valve. A summary of RAMONA-3B neutron kinetics and thermal hydraulics models is presented in the Appendix.« less

ANALYSIS OF BORON DILUTION TRANSIENTS IN PWRS.

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

DIAMOND,D.J.BROMLEY,B.P.ARONSON,A.L.

2004-02-04

A study has been carried out with PARCS/RELAP5 to understand the consequences of hypothetical boron dilution events in pressurized water reactors. The scenarios of concern start with a small-break loss-of-coolant accident. If the event leads to boiling in the core and then the loss of natural circulation, a boron-free condensate can accumulate in the cold leg. The dilution event happens when natural circulation is re-established or a reactor coolant pump (RCP) is restarted in violation of operating procedures. This event is of particular concern in B&W reactors with a lowered-loop design and is a Generic Safety Issue for the U.S.more » Nuclear Regulatory Commission. The results of calculations with the reestablishment of natural circulation show that there is no unacceptable fuel damage. This is determined by calculating the maximum fuel pellet enthalpy, based on the three-dimensional model, and comparing it with the criterion for damage. The calculation is based on a model of a B&W reactor at beginning of the fuel cycle. If an RCP is restarted, unacceptable fuel damage may be possible in plants with sufficiently large volumes of boron-free condensate in the cold leg.« less

Heated-Pressure-Ball Monopropellant Rocket Engine

NASA Technical Reports Server (NTRS)

Greene, William D.

2005-01-01

A recent technology disclosure presents a concept for a monopropellant thermal spacecraft thruster that would feature both the simplicity of a typical prior pressure-fed propellant supply system and the smaller mass and relative compactness of a typical prior pump-fed system. The source of heat for this thruster would likely be a nuclear- fission reactor. The propellant would be a cryogenic fluid (a liquefied low-molecular-weight gas) stored in a tank at a low pressure. The propellant would flow from the tank, through a feedline, into three thick-walled spherical tanks, denoted pressure balls, that would be thermally connected to the reactor. Valves upstream and downstream of the pressure balls would be operated in a three-phase cycle in which propellant would flow into one pressure ball while the fluid underwent pressurization through heating in another ball and pressurized propellant was discharged from the remaining ball into the reactor. After flowing through the reactor, wherein it would be further heated, the propellant would be discharged through an exhaust nozzle to generate thrust. A fraction of the pressurized gas from the pressure balls would be diverted to maintain the desired pressure in the tank.

Full reactor coolant system chemical decontamination qualification programs

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

Miller, P.E.

1995-03-01

Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systemsmore » leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.« less

The Munich accelerator for fission fragments MAFF

NASA Astrophysics Data System (ADS)

Habs, D.; Groß, M.; Assmann, W.; Ames, F.; Bongers, H.; Emhofer, S.; Heinz, S.; Henry, S.; Kester, O.; Neumayr, J.; Ospald, F.; Reiter, P.; Sieber, T.; Szerypo, J.; Thirolf, P. G.; Varentsov, V.; Wilfart, T.; Faestermann, T.; Krücken, R.; Maier-Komor, P.

2003-05-01

The Munich Accelerator for Fission Fragments MAFF has been designed for the new Munich research reactor FRM-II. It will deliver several intense beams (˜3×10 11 s -1) of very neutron-rich fission fragments with a final energy of 30 keV (low-energy beam) or energies between 3.7 and 5.9 MeV· A (high-energy beam). Such beams are of interest for the creation of super-heavy elements by fusion reactions, nuclear spectroscopy of exotic nuclei, but they also have a potential for applications, e.g. in medicine. Presently the Munich research reactor FRM-II is ready for operation, but authorities delay the final permission to turn the reactor critical probably till the end of 2002. Only after this final permission the financing of the major parts of MAFF can start. On the other hand all major components have been designed and special components have been tested in separate setups.

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.

Embedded Sensors and Controls to Improve Component Performance and Reliability -- Loop-scale Testbed Design Report

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

Melin, Alexander M.; Kisner, Roger A.

2016-09-01

Embedded instrumentation and control systems that can operate in extreme environments are challenging to design and operate. Extreme environments limit the options for sensors and actuators and degrade their performance. Because sensors and actuators are necessary for feedback control, these limitations mean that designing embedded instrumentation and control systems for the challenging environments of nuclear reactors requires advanced technical solutions that are not available commercially. This report details the development of testbed that will be used for cross-cutting embedded instrumentation and control research for nuclear power applications. This research is funded by the Department of Energy's Nuclear Energy Enabling Technologymore » program's Advanced Sensors and Instrumentation topic. The design goal of the loop-scale testbed is to build a low temperature pump that utilizes magnetic bearing that will be incorporated into a water loop to test control system performance and self-sensing techniques. Specifically, this testbed will be used to analyze control system performance in response to nonlinear and cross-coupling fluid effects between the shaft axes of motion, rotordynamics and gyroscopic effects, and impeller disturbances. This testbed will also be used to characterize the performance losses when using self-sensing position measurement techniques. Active magnetic bearings are a technology that can reduce failures and maintenance costs in nuclear power plants. They are particularly relevant to liquid salt reactors that operate at high temperatures (700 C). Pumps used in the extreme environment of liquid salt reactors provide many engineering challenges that can be overcome with magnetic bearings and their associated embedded instrumentation and control. This report will give details of the mechanical design and electromagnetic design of the loop-scale embedded instrumentation and control testbed.« less

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

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

Fanning, T. H.; Brunett, A. J.; Sumner, T.

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

Bulk-Flow Analysis, part A

NASA Technical Reports Server (NTRS)

Childs, Dara W.

1993-01-01

The bulk-flow analysis results for this contract are incorporated in the following publications: 'Fluid-Structure Interaction Forces at Pump-Impeller Shroud Surfaces for Axial Vibration Analysis'; 'Centrifugal Acceleration Modes for Incompressible Fluid in the Leakage Annulus Between a Shrouded Pump Impeller and Its Housing'; 'Influence of Impeller Shroud Forces on Pump Rotordynamics'; 'Pressure Oscillation in the Leakage Annulus Between a Shrouded Impeller and Its Housing Due to Impeller-Discharge-Pressure Disturbances'; and 'Compressibility Effects on Rotor Forces in the Leakage Path Between a Shrouded Pump Impeller and Its Housing'. These publications are summarized and included in this final report. Computational Fluid Mechanics (CFD) results developed by Dr. Erian Baskharone are reported separately.

Analytical model for thermal lensing and spherical aberration in diode side-pumped Nd:YAG laser rod having Gaussian pump profile

NASA Astrophysics Data System (ADS)

M, H. Moghtader Dindarlu; M Kavosh, Tehrani; H, Saghafifar; A, Maleki

2015-12-01

In this paper, according to the temperature and strain distribution obtained by considering the Gaussian pump profile and dependence of physical properties on temperature, we derive an analytical model for refractive index variations of the diode side-pumped Nd:YAG laser rod. Then we evaluate this model by numerical solution and our maximum relative errors are 5% and 10% for variations caused by thermo-optical and thermo-mechanical effects; respectively. Finally, we present an analytical model for calculating the focal length of the thermal lens and spherical aberration. This model is evaluated by experimental results.

Dynamic modeling of temperature change in outdoor operated tubular photobioreactors.

PubMed

Androga, Dominic Deo; Uyar, Basar; Koku, Harun; Eroglu, Inci

2017-07-01

In this study, a one-dimensional transient model was developed to analyze the temperature variation of tubular photobioreactors operated outdoors and the validity of the model was tested by comparing the predictions of the model with the experimental data. The model included the effects of convection and radiative heat exchange on the reactor temperature throughout the day. The temperatures in the reactors increased with increasing solar radiation and air temperatures, and the predicted reactor temperatures corresponded well to the measured experimental values. The heat transferred to the reactor was mainly through radiation: the radiative heat absorbed by the reactor medium, ground radiation, air radiation, and solar (direct and diffuse) radiation, while heat loss was mainly through the heat transfer to the cooling water and forced convection. The amount of heat transferred by reflected radiation and metabolic activities of the bacteria and pump work was negligible. Counter-current cooling was more effective in controlling reactor temperature than co-current cooling. The model developed identifies major heat transfer mechanisms in outdoor operated tubular photobioreactors, and accurately predicts temperature changes in these systems. This is useful in determining cooling duty under transient conditions and scaling up photobioreactors. The photobioreactor design and the thermal modeling were carried out and experimental results obtained for the case study of photofermentative hydrogen production by Rhodobacter capsulatus, but the approach is applicable to photobiological systems that are to be operated under outdoor conditions with significant cooling demands.

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Hoover, Mark D.

1991-07-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects. (For individual items see A93-13752 to A93-13937)

SNAP (Space Nuclear Auxiliary Power) reactor overview. Final report, June 1982-December 1983

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

Voss, S.S.

1984-08-01

The SNAP reactor programs are outlined in this report. A summary of the program is included along with a technical outline of the SER, S2DR, SNAP 10A/SNAPSHOT, S8ER, and S8DR reactor systems. Specifications of the designs, the design logic and a conclusion outlining some of the program weaknesses are given.

Research and development of a heat-pump water heater. Volume 2. R and D task reports

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

Dunning, R.L.; Amthor, F.R.; Doyle, E.J.

1978-08-01

The heat pump water heater is a device that works much like a window air conditioner except that heat from the home is pumped into a water tank rather than to the outdoors. The objective established for the device is to operate with a Coefficient of Performance (COP) of 3 or, an input of one unit of electric energy would create three units of heat energy in the form of hot water. With such a COP, the device would use only one-third the energy and at one-third the cost of a standard resistance water heater. This Volume 2 contains themore » final reports of the three major tasks performed in Phase I. In Task 2, a market study identifies the future market and selects an initial target market and channel of distribution, all based on an analysis of the parameters affecting feasibility of the device and the factors that will affect its market acceptance. In the Task 3 report, the results of a design and test program to arrive at final designs of heat pumps for both new water heaters and for retrofitting existing water heaters are presented. In the Task 4 report, a plan for an extensive field demonstration involving use in actual homes is presented. Volume 1 contains a final summary report of the information in Volume 2.« less

Sensor-based atomic layer deposition for rapid process learning and enhanced manufacturability

NASA Astrophysics Data System (ADS)

Lei, Wei

In the search for sensor based atomic layer deposition (ALD) process to accelerate process learning and enhance manufacturability, we have explored new reactor designs and applied in-situ process sensing to W and HfO 2 ALD processes. A novel wafer scale ALD reactor, which features fast gas switching, good process sensing compatibility and significant similarity to the real manufacturing environment, is constructed. The reactor has a unique movable reactor cap design that allows two possible operation modes: (1) steady-state flow with alternating gas species; or (2) fill-and-pump-out cycling of each gas, accelerating the pump-out by lifting the cap to employ the large chamber volume as ballast. Downstream quadrupole mass spectrometry (QMS) sampling is applied for in-situ process sensing of tungsten ALD process. The QMS reveals essential surface reaction dynamics through real-time signals associated with byproduct generation as well as precursor introduction and depletion for each ALD half cycle, which are then used for process learning and optimization. More subtle interactions such as imperfect surface saturation and reactant dose interaction are also directly observed by QMS, indicating that ALD process is more complicated than the suggested layer-by-layer growth. By integrating in real-time the byproduct QMS signals over each exposure and plotting it against process cycle number, the deposition kinetics on the wafer is directly measured. For continuous ALD runs, the total integrated byproduct QMS signal in each ALD run is also linear to ALD film thickness, and therefore can be used for ALD film thickness metrology. The in-situ process sensing is also applied to HfO2 ALD process that is carried out in a furnace type ALD reactor. Precursor dose end-point control is applied to precisely control the precursor dose in each half cycle. Multiple process sensors, including quartz crystal microbalance (QCM) and QMS are used to provide real time process information. The sensing results confirm the proposed surface reaction path and once again reveal the complexity of ALD processes. The impact of this work includes: (1) It explores new ALD reactor designs which enable the implementation of in-situ process sensors for rapid process learning and enhanced manufacturability; (2) It demonstrates in the first time that in-situ QMS can reveal detailed process dynamics and film growth kinetics in wafer-scale ALD process, and thus can be used for ALD film thickness metrology. (3) Based on results from two different processes carried out in two different reactors, it is clear that ALD is a more complicated process than normally believed or advertised, but real-time observation of the operational chemistries in ALD by in-situ sensors provides critical insight to the process and the basis for more effective process control for ALD applications.

Concrete volute pumps: technology review and improvement

NASA Astrophysics Data System (ADS)

Prunières, R.; Longatte, F.; Catelan, F. X.; Philippot, J. M.

2012-11-01

When pumps need to deliver large water flow rates (typically more than 5 m3.s-1), concrete volute pumps (CVP) offer an interesting alternative to standard vertical wet-pit pumps. One of the major advantages of CVP is its simplicity in terms of design, manufacturability and maintainability. In addition, CVP geometrical arrangement allows to reach high performances in terms of hydraulic and mechanical behaviour. These advantages can be specifically appreciated when such pumps are used in the energy field for Power Plants which need high flow rate and reliability, and can lead to important financial savings over the Plant lifetime compared to vertical wet-pit pumps. Finally, as CVP was for a long time limited to total head rise lower than 30 mWC, it was established through CFD analysis that the addition of guide vanes between the impeller and the volute allows to achieve higher head rise without risk.

Experimental study on hydraulic characteristic around trash rack of a pumping station

NASA Astrophysics Data System (ADS)

Zhou, MinZhe; Li, TongChun; Lin, XiangYang; Liu, XiaoQing; Ding, Yuan; Liu, GuangYuan

2017-11-01

This paper focuses on flow pattern around trash rack of intake of a pumping station project. This pumping station undertake the task of supplying up to 3,500,000 m3 water per day for a megacity. Considering the large flow rate, high lift, multi-pipe supply and long-time operation in this water conveyance pumping station, we built a physical model test to measure the flow velocity and observe the flow pattern to verify the reasonability of preliminary design. In this test, we set 3 layers of current meters around each trash rack of intake in reservoir to collect the flow velocity. Furthermore, we design 2 operating conditions of 9 pumps to observe the change of flow pattern. Finally, we found the velocity data were in a normal range under 2 different operating conditions of the 9 pump units.

Simulation and Particle-Tracking Analysis of Selected Ground-Water Pumping Scenarios at Vogtle Electric Generation Plant, Burke County, Georgia

USGS Publications Warehouse

Cherry, Gregory S.; Clarke, John S.

2007-01-01

The source of ground water to production wells at Vogtle Electric Generation Plant (VEGP), a nuclear power plant in Burke County, Georgia, was simulated under existing (2002) and potential future pumping conditions using an existing U.S. Geological Survey (USGS) MODFLOW ground-water flow model of a 4,455-square-mile area in the Coastal Plain of Georgia and South Carolina. Simulation results for three steady-state pumping scenarios were compared to each other and to a 2002 Base Case condition. The pumping scenarios focused on pumping increases at VEGP resulting from projected future demands and the addition of two electrical-generating reactor units. Scenarios simulated pumping increases at VEGP ranging from 1.09 to 3.42 million gallons per day (Mgal/d), with one of the scenarios simulating the elimination of 5.3 Mgal/d of pumping at the Savannah River Site (SRS), a U.S. Department of Energy facility located across the Savannah River from VEGP. The largest simulated water-level changes at VEGP were for the scenario whereby pumping at the facility was more than tripled, resulting in drawdown exceeding 4-8 feet (ft) in the aquifers screened in the production wells. For the scenario that eliminated pumping at SRS, water-level rises of as much as 4-8 ft were simulated in the same aquifers at SRS. Results of MODFLOW simulations were analyzed using the USGS particle-tracking code MODPATH to determine the source of water and associated time of travel to VEGP production wells. For each of the scenarios, most of the recharge to VEGP wells originated in an upland area near the county line between Burke and Jefferson Counties, Georgia, with none of the recharge originating on SRS or elsewhere in South Carolina. An exception occurs for the scenario whereby pumping at VEGP was more than tripled. For this scenario, some of the recharge originates in an upland area in eastern Barnwell County, South Carolina. Simulated mean time of travel from recharge areas to VEGP wells for the Base Case and the three other pumping scenarios was between about 2,700 and 3,800 years, with some variation related to changes in head gradients because of pumping changes.

Study of application rates of aerosol and pump hair sprays. Final report, July 1986-November 1987

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

Boggs, R.R.; Belmont, B.

Application rates of three hair spray dispensing systems, aerosol, pump, and Exxel packaging were determined through a six-week user panel of approximately 300 people. In addition, photochemically reactive organic compounds (PROC) application rates were determined through chemical analysis of the products. The user panel was stratified on the basis of sex, dispenser (pump/aerosol), and age (adult/teen). Weighted-application rates and weighted PROC application rates are included. A Mann-Whitney evaluation was made to evaluate differences between data sets. Product-usage data for both male and female adult groups support the conclusion that increased use of either pumps or Exxel packaging for hair spraymore » would reduce PROC emissions in California. Data from adult groups also indicate that use of Exxel packaging in place of pumps would not reduce PROC. Consumer preference was also sampled. Adult pump users were not very willing to switch to aerosols, but on the order of half of aerosol users were willing to switch to pumps.« less

Plasma Reforming of Liquid Hydrocarbon Fuels in Non-Thermal Plasma-Liquid Systems

DTIC Science & Technology

2010-04-30

microporous liquid which has a very large ratio of the plasma-liquid contact surface to the plasma volume. As is known the ultrasonic (US) cavitation is a very...effective method for creating micropores in liquid [17]. Therefore, the DGCLW with additional US pumping is also very interesting for research and...electrodes. Another PLS reactor was prepared with the DGCLW working with the air flow in the liquid under the induced microporous

High Power Mid Wave Infrared Semiconductor Lasers

DTIC Science & Technology

2006-06-15

resonance and the gain spectrum. The devices were grown using solid source molecular beam epitaxy (MBE) in a V80 reactor. Two side polished, undoped...verify the inherent low activation energy. N-type and P-type AISb, and various compositions of InxAl 1xSb, were grown by solid-source molecular beam ...level monitoring. Advances in epitaxial growth of semiconductor materials have allowed the development of Arsenic- free optically-pumped MWIR lasers on

Final Report for the Erosion-Corrosion Anaysis of Tank 241-AW-02E Feed Pump Pit Jumpers B-2 and 1-4 Removed from Service in 2013

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

Page, Jason S.

2014-04-07

This document is the final report summarizing the results in the examination of two pipe sections (jumpers) from the tank 241-AW-02E feed pump pit in the 241-AW tank farm. These pipe section samples consisted of jumper AW02E-WT-J-[B – 2] and jumper AW02E-WT-J-[1 – 4]. For the remainder of this report, these jumpers will be referred to as B – 2 and 1 – 4.

Modelling of the anti-neutrino production and spectra from a Magnox reactor

NASA Astrophysics Data System (ADS)

Mills, Robert W.; Mountford, David J.; Coleman, Jonathon P.; Metelko, Carl; Murdoch, Matthew; Schnellbach, Yan-Jie

2018-01-01

The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor operation was calculated. The results of the preliminary calculations are shown and limitations in the methods and data discussed.

Note: A dual temperature closed loop batch reactor for determining the partitioning of trace gases within CO{sub 2}-water systems

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

Warr, Oliver, E-mail: oliver.warr@earth.ox.ac.uk; Ballentine, Christopher J.; Rochelle, Christopher A.

An experimental approach is presented which can be used to determine partitioning of trace gases within CO{sub 2}-water systems. The key advantages of this system are (1) The system can be isolated with no external exchange, making it ideal for experiments with conservative tracers. (2) Both phases can be sampled concurrently to give an accurate composition at each phase at any given time. (3) Use of a lower temperature flow loop outside of the reactor removes contamination and facilitates sampling. (4) Rapid equilibration at given pressure/temperature conditions is significantly aided by stirring and circulating the water phase using a magneticmore » stirrer and high-pressure liquid chromatography pump, respectively.« less

Pilot Study for UVA-LED Disinfection of Escherichia coli in Water for Space and Earth Applications

NASA Technical Reports Server (NTRS)

Ragolta, Carolina

2010-01-01

To test the efficacy of UVA-LED disinfection, a solution of Escherichia coli was pumped through a modified drip flow reactor at a flow rate of 1 ml/min. The experiment was conducted in a controlled environment chamber to ensure that temperature did not cause disinfection. The reactor featured three wells with different treatments: UVA-LED irradiation, UVA-LEDs with Ti02, and UVA-LEDs with nanosilver. Samples from each well were taken throughout a 340 hour period, inactivated, assayed, and analyzed for E. coli disinfection. Results of the duplicate experiments indicated longer exposure times are needed for UVA-LED disinfection of E. coli in water. Further research would consider a longer sampling period and different test conditions, such as increased contact area and various flow rates.

The slightly-enriched spectral shift control reactor

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

Martin, W.R.; Lee, J.C.; Larsen, E.W.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

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

Liu, Xiaobing; Zheng, O'Neill; Niu, Fuxin

Most commercial ground source heat pump systems (GSHP) in the United States are in a distributed configuration. These systems circulate water or an anti-freeze solution through multiple heat pump units via a central pumping system, which usually uses variable speed pump(s). Variable speed pumps have potential to significantly reduce pumping energy use; however, the energy savings in reality could be far away from its potential due to improper pumping system design and controls. In this paper, a simplified hydronic pumping system was simulated with the dynamic Modelica models to evaluate three different pumping control strategies. This includes two conventional controlmore » strategies, which are to maintain a constant differential pressure across either the supply and return mains, or at the most hydraulically remote heat pump; and an innovative control strategy, which adjusts system flow rate based on the demand of each heat pump. The simulation results indicate that a significant overflow occurs at part load conditions when the variable speed pump is controlled to main a constant differential pressure across the supply and return mains of the piping system. On the other hand, an underflow occurs at part load conditions when the variable speed pump is controlled to maintain a constant differential pressure across the furthest heat pump. The flow-demand-based control can provide needed flow rate to each heat pump at any given time, and with less pumping energy use than the two conventional controls. Finally, a typical distributed GSHP system was studied to evaluate the energy saving potential of applying the flow-demand-based pumping control strategy. This case study shows that the annual pumping energy consumption can be reduced by 62% using the flow-demand-based control compared with that using the conventional pressure-based control to maintain a constant differential pressure a cross the supply and return mains.« less

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

Reactor Simulator Testing Overview

NASA Technical Reports Server (NTRS)

Schoenfeld, Michael P.

2013-01-01

OBJECTIVE: Integrated testing of the TDU components TESTING SUMMARY: a) Verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. b) Thermal test heat regeneration design aspect of a cold trap purification filter. c) Pump performance test at pump voltages up to 150 V (targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V). TESTING HIGHLIGHTS: a) Gas and vacuum ground support test equipment performed effectively for NaK fill, loop pressurization, and NaK drain operations. b) Instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

Advanced shield development for a fission surface power system for the lunar surface

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

A. E. Craft; I. J. Silver; C. M. Clark

A nuclear reactor power system such as the affordable fission surface power system enables a potential outpostonthemoon.Aradiation shieldmustbe included in the reactor system to reduce the otherwise excessive dose to the astronauts and other vital system components. The radiation shield is typically the most massive component of a space reactor system, and thus must be optimized to reduce mass asmuchas possible while still providing the required protection.Various shield options for an on-lander reactor system are examined for outpost distances of 400m and 1 kmfromthe reactor. Also investigated is the resulting mass savings from the use of a high performance cermetmore » fuel. A thermal analysis is performed to determine the thermal behaviours of radiation shields using borated water. For an outpost located 1000m from the core, a tetramethylammonium borohydride shield is the lightest (5148.4 kg), followed by a trilayer shield (boron carbide–tungsten–borated water; 5832.3 kg), and finally a borated water shield (6020.7 kg). In all of the final design cases, the temperature of the borated water remains below 400 K.« less

An economic evaluation comparison of solar water pumping system with engine pumping system for rice cultivation

NASA Astrophysics Data System (ADS)

Treephak, Kasem; Thongpron, Jutturit; Somsak, Dhirasak; Saelao, Jeerawan; Patcharaprakiti, Nopporn

2015-08-01

In this paper we propose the design and economic evaluation of the water pumping systems for rice cultivation using solar energy, gasoline fuel and compare both systems. The design of the water and gasoline engine pumping system were evaluated. The gasoline fuel cost used in rice cultivation in an area of 1.6 acres. Under same conditions of water pumping system is replaced by the photovoltaic system which is composed of a solar panel, a converter and an electric motor pump which is compose of a direct current (DC) motor or an alternating current (AC) motor with an inverter. In addition, the battery is installed to increase the efficiency and productivity of rice cultivation. In order to verify, the simulation and economic evaluation of the storage energy battery system with batteries and without batteries are carried out. Finally the cost of four solar pumping systems was evaluated and compared with that of the gasoline pump. The results showed that the solar pumping system can be used to replace the gasoline water pumping system and DC solar pump has a payback less than 10 years. The systems that can payback the fastest is the DC solar pumping system without batteries storage system. The system the can payback the slowest is AC solar pumping system with batteries storage system. However, VAC motor pump of 220 V can be more easily maintained than the motor pump of 24 VDC and batteries back up system can supply a more stable power to the pump system.

Biological nutrient removal by internal circulation upflow sludge blanket reactor after landfill leachate pretreatment.

PubMed

Abood, Alkhafaji R; Bao, Jianguo; Abudi, Zaidun N

2013-10-01

The removal of biological nutrient from mature landfill leachate with a high nitrogen load by an internal circulation upflow sludge blanket (ICUSB) reactor was studied. The reactor is a set of anaerobic-anoxic-aerobic (A2/O) bioreactors, developed on the basis of an expended granular sludge blanket (EGSB), granular sequencing batch reactor (GSBR) and intermittent cycle extended aeration system (ICEAS). Leachate was subjected to stripping by agitation process and poly ferric sulfate coagulation as a pretreatment process, in order to reduce both ammonia toxicity to microorganisms and the organic contents. The reactor was operated under three different operating systems, consisting of recycling sludge with air (A2/O), recycling sludge without air (low oxygen) and a combination of both (A2/O and low oxygen). The lowest effluent nutrient levels were realised by the combined system of A2/O and low oxygen, which resulted in effluent of chemical oxygen demand (COD), NH3-N and biological oxygen demand (BOD5) concentrations of 98.20, 13.50 and 22.50 mg/L. The optimal operating conditions for the efficient removal of biological nutrient using the ICUSB reactor were examined to evaluate the influence of the parameters on its performance. The results showed that average removal efficiencies of COD and NH3-N of 96.49% and 99.39%, respectively were achieved under the condition of a hydraulic retention time of 12 hr, including 4 hr of pumping air into the reactor, with dissolved oxygen at an rate of 4 mg/L and an upflow velocity 2 m/hr. These combined processes were successfully employed and effectively decreased pollutant loading.

Vacuum system transient simulator and its application to TFTR

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

Sredniawski, J.

The vacuum system transient simulator (VSTS) models transient gas transport throughout complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. VSTS is capable of treating gas models of up to 10 species, for all flow regimes from pure molecular to continuum. Viscous interactions between species are considered as well as non-uniform temperature of a system. Although this program was specifically developed for use on the Tokamak Fusion Test Reactor (TFTR) project at Princeton, it is a generalized tool capable of handling a broad range of vacuum system problems. During the TFTR engineering design phase, VSTSmore » has been used in many applications. Two applications selected for presentation are: torus vacuum pumping system performance between 400 Ci tritium pulses and tritium backstreaming to neutral beams during pulses.« less

Assembly and Thermal Hydraulic Test of a Stainless Steel Sodium-Potassium Circuit

NASA Technical Reports Server (NTRS)

Garber, A.; Godfroy, T.; Webster, K.

2007-01-01

Early Flight Fission Test Facilities (EFF-TF) team has been tasked by the NASA Marshall Space Flight Center Nuclear Systems Office to design, fabricate, and test an actively pumped alkali metal flow circuit. The system was originally built for use with lithium, but due to a shift in focus, it was redesigned for use with a eutectic mixture of sodium potassium (NaK). Basic circuit components include: reactor segment, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and a spill reservoir. A 37-pin partial-array core (pin and flow path dimensions are the same as those in a full design) was selected for fabrication and test. This paper summarizes the first fill and checkout testing of the Stainless Steel NaK-Cooled Circuit (SNaKC).

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.

Effects of high source flow and high pumping speed on gas source molecular beam epitaxy / chemical beam epitaxy

NASA Astrophysics Data System (ADS)

McCollum, M. J.; Jackson, S. L.; Szafranek, I.; Stillman, G. E.

1990-10-01

We report the growth of GaAs by molecular beam epitaxy (MBE), gas source molecular beam epitaxy (GSMBE), and chemical beam epitaxy (CBE) in an epitaxial III-V reactor which features high pumping speed. The system is comprised of a modified Perkin-Elmer 430P molecular beam epitaxy system and a custom gas source panel from Emcore. The growth chamber is pumped with a 7000 1/s (He) diffusion pump (Varian VHS-10 with Monsanto Santovac 5 oil). The gas source panel includes pressure based flow controllers (MKS 1150) allowing triethylaluminum (TEA), triethylgallium (TEG), and trimethylindium (TMI) to be supplied without the use of hydrogen. All source lines, including arsine and phosphine, are maintained below atmospheric pressure. The high pumping speed allows total system flow rates as high as 100 SCCM and V/III ratios as high as 100. The purity of GaAs grown by MBE in this system increases with pumping speed. GaAs layers grown by GSMBE with arsine flows of 10 and 20 SCCM have electron concentrations of 1 × 10 15 cm -3 (μ 77=48,000 cm 2/V·) and 2 × 10 14 cm -3 (μ 77=78,000 cm 2/V·s) respectively. El ectron concentration varies with hydride injector temperature such that the minimum in electron concentration occurs for less than complete cracking. The effect of V/III ratio and the use of a metal eutectic bubbler on residual carrier concentration in GaAs grown by CBE is presented. Intentional Si and Be doping of CBE grown GaAs is demonstrated at a high growth rate of 5.4 μm/h.

Cost Reductions for Wastewater Treatment Utilizing Water Management at Holston Army Ammunition Plant

DTIC Science & Technology

1976-05-01

says that the granular carbon used is made from bituminous coal. As the waste stream pass through a bed of carbon granules, com- pounds are adsorbed to...findings of laboratory-scale reactor studies conducted at Purdue University for * Clark, Dietz and Associates. The original recommendations and cost...Pretreatment Denitrification by Submerged Anaerbbic I ilters I ~ Trickling Filters S F ,2al Clarification "•’i Pump - ~ Sludge ,Treatment Dual Media Filh:ration

Accident analysis of heavy water cooled thorium breeder reactor

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

Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

2015-04-16

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k,more » and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.« less

Accident analysis of heavy water cooled thorium breeder reactor

NASA Astrophysics Data System (ADS)

Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

2015-04-01

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

Vacuum system design and tritium inventory for the charge exchange diagnostic on the Tokamak Fusion Test Reactor

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

Medley, S.S.

The application of charge exchange analyzers for the measurement of ion temperature in fusion plasma experiments requires a direct connection between the diagnostic and plasma-discharge vacuum chambers. Differential pumping of the gas load from the diagnostic stripping cell operated at > or approx. = 10/sup -3/ Torr is required to maintain the analyzer chamber at a pressure of < or approx. = 10/sup -6/ Torr. The migration of gases between the diagnostic and plasma vacuum chambers must be minimized. In particular, introduction of the analyzer stripping cell gas into the plasma chamber having a base pressure of < or approx.more » = 10/sup -8/ Torr must be suppressed. The charge exchange diagnostic for the Tokamak Fusion Test Reactor (TFTR) is comprised of two analyzer systems designed to contain a total of 18 independent mass/energy analyzers and one diagnostic neutral beam rated at 80 keV, 15 A. The associated arrays of multiple, interconnected vacuum systems were analyzed using the Vacuum System Transient Simulator (Vsts) computer program which models the transient transport of multigas species through complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. In addition to providing improved design performance at reduced costs, the analysis yields estimates for the exchange of tritium from the torus to the diagnostic components and of the diagnostic working gases to the torus.« less

Completion Summary for Well NRF-16 near the Naval Reactors Facility, Idaho National Laboratory, Idaho

USGS Publications Warehouse

Twining, Brian V.; Fisher, Jason C.; Bartholomay, Roy C.

2010-01-01

In 2009, the U.S. Geological Survey in cooperation with the U.S. Department of Energy's Naval Reactors Laboratory Field Office, Idaho Branch Office cored and completed well NRF-16 for monitoring the eastern Snake River Plain (SRP) aquifer. The borehole was initially cored to a depth of 425 feet below land surface and water samples and geophysical data were collected and analyzed to determine if well NRF-16 would meet criteria requested by Naval Reactors Facility (NRF) for a new upgradient well. Final construction continued after initial water samples and geophysical data indicated that NRF-16 would produce chemical concentrations representative of upgradient aquifer water not influenced by NRF facility disposal, and that the well was capable of producing sustainable discharge for ongoing monitoring. The borehole was reamed and constructed as a Comprehensive Environmental Response Compensation and Liability Act monitoring well complete with screen and dedicated pump. Geophysical and borehole video logs were collected after coring and final completion of the monitoring well. Geophysical logs were examined in conjunction with the borehole core to identify primary flow paths for groundwater, which are believed to occur in the intervals of fractured and vesicular basalt and to describe borehole lithology in detail. Geophysical data also were examined to look for evidence of perched water and the extent of the annular seal after cement grouting the casing in place. Borehole videos were collected to confirm that no perched water was present and to examine the borehole before and after setting the screen in well NRF-16. Two consecutive single-well aquifer tests to define hydraulic characteristics for well NRF-16 were conducted in the eastern SRP aquifer. Transmissivity and hydraulic conductivity averaged from the aquifer tests were 4.8 x 103 ft2/d and 9.9 ft/d, respectively. The transmissivity for well NRF-16 was within the range of values determined from past aquifer tests in other wells near NRF of 4.4 x 102 to 5.1 x 105 ft2/d. Water samples were analyzed for metals, nutrients, total organic carbon, volatile organic compounds, semi-volatile organic compounds, herbicides, pesticides, polychlorinated biphenols, and radionuclides. All chloride, nitrate, and sulfate concentrations were less than background concentrations for the eastern SRP aquifer north of the NRF. Concentrations in water samples for most of the organic compounds and radionuclides were less than the reporting limits and reporting levels.

The effect of mixing ratio variation of sludge and organic solid waste on biodrying process

NASA Astrophysics Data System (ADS)

Nasution, A. C.; Kristanto, G. A.

2018-01-01

In this study, organic waste was co-biodried with sludge cake to determine which mixing ratio gave the best result. The organic waste was consisted of dried leaves and green leaves, while the sludge cake was obtained from a waste water treatment plant in Bekasi. The experiment was performed on 3 lab-scale reactors with same specifications. After 21 days of experiment, it was found that the reactor with the lowest mixing fraction of sludge (5:1) has the best temperature profile and highest moisture content depletion compared with others. Initial moisture content and initial volatile solid content of this reactor’s feedstock was 52.25% and 82.4% respectively. The airflow rate was 10 lpm. After biodrying was done, the final moisture content of the feedstock from Reactor C was 22.0% and the final volatile solid content was 75.9%.The final calorific value after biodrying process was 3179,28kcal/kg.

A simulation-based study on different control strategies for variable speed pump in distributed ground source heat pump systems

DOE PAGES

Liu, Xiaobing; Zheng, O'Neill; Niu, Fuxin

2016-01-01

Most commercial ground source heat pump systems (GSHP) in the United States are in a distributed configuration. These systems circulate water or an anti-freeze solution through multiple heat pump units via a central pumping system, which usually uses variable speed pump(s). Variable speed pumps have potential to significantly reduce pumping energy use; however, the energy savings in reality could be far away from its potential due to improper pumping system design and controls. In this paper, a simplified hydronic pumping system was simulated with the dynamic Modelica models to evaluate three different pumping control strategies. This includes two conventional controlmore » strategies, which are to maintain a constant differential pressure across either the supply and return mains, or at the most hydraulically remote heat pump; and an innovative control strategy, which adjusts system flow rate based on the demand of each heat pump. The simulation results indicate that a significant overflow occurs at part load conditions when the variable speed pump is controlled to main a constant differential pressure across the supply and return mains of the piping system. On the other hand, an underflow occurs at part load conditions when the variable speed pump is controlled to maintain a constant differential pressure across the furthest heat pump. The flow-demand-based control can provide needed flow rate to each heat pump at any given time, and with less pumping energy use than the two conventional controls. Finally, a typical distributed GSHP system was studied to evaluate the energy saving potential of applying the flow-demand-based pumping control strategy. This case study shows that the annual pumping energy consumption can be reduced by 62% using the flow-demand-based control compared with that using the conventional pressure-based control to maintain a constant differential pressure a cross the supply and return mains.« less

Alcohol synthesis in a high-temperature slurry reactor

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

Roberts, G.W.; Marquez, M.A.; McCutchen, M.S.

1995-12-31

The overall objective of this contract is to develop improved process and catalyst technology for producing higher alcohols from synthesis gas or its derivatives. Recent research has been focused on developing a slurry reactor that can operate at temperatures up to about 400{degrees}C and on evaluating the so-called {open_quotes}high pressure{close_quotes} methanol synthesis catalyst using this reactor. A laboratory stirred autoclave reactor has been developed that is capable of operating at temperatures up to 400{degrees}C and pressures of at least 170 atm. The overhead system on the reactor is designed so that the temperature of the gas leaving the system canmore » be closely controlled. An external liquid-level detector is installed on the gas/liquid separator and a pump is used to return condensed slurry liquid from the separator to the reactor. In order to ensure that gas/liquid mass transfer does not influence the observed reaction rate, it was necessary to feed the synthesis gas below the level of the agitator. The performance of a commercial {open_quotes}high pressure {close_quotes} methanol synthesis catalyst, the so-called {open_quotes}zinc chromite{close_quotes} catalyst, has been characterized over a range of temperature from 275 to 400{degrees}C, a range of pressure from 70 to 170 atm., a range of H{sub 2}/CO ratios from 0.5 to 2.0 and a range of space velocities from 2500 to 10,000 sL/kg.(catalyst),hr. Towards the lower end of the temperature range, methanol was the only significant product.« less

Numerical Simulation of Tubular Pumping Systems with Different Regulation Methods

NASA Astrophysics Data System (ADS)

Zhu, Honggeng; Zhang, Rentian; Deng, Dongsheng; Feng, Xusong; Yao, Linbi

2010-06-01

Since the flow in tubular pumping systems is basically along axial direction and passes symmetrically through the impeller, most satisfying the basic hypotheses in the design of impeller and having higher pumping system efficiency in comparison with vertical pumping system, they are being widely applied to low-head pumping engineering. In a pumping station, the fluctuation of water levels in the sump and discharge pool is most common and at most time the pumping system runs under off-design conditions. Hence, the operation of pump has to be flexibly regulated to meet the needs of flow rates, and the selection of regulation method is as important as that of pump to reduce operation cost and achieve economic operation. In this paper, the three dimensional time-averaged Navier-Stokes equations are closed by RNG κ-ɛ turbulent model, and two tubular pumping systems with different regulation methods, equipped with the same pump model but with different designed system structures, are numerically simulated respectively to predict the pumping system performances and analyze the influence of regulation device and help designers make final decision in the selection of design schemes. The computed results indicate that the pumping system with blade-adjusting device needs longer suction box, and the increased hydraulic loss will lower the pumping system efficiency in the order of 1.5%. The pumping system with permanent magnet motor, by means of variable speed regulation, obtains higher system efficiency partly for shorter suction box and partly for different structure design. Nowadays, the varied speed regulation is realized by varied frequency device, the energy consumption of which is about 3˜4% of output power of the motor. Hence, when the efficiency of variable frequency device is considered, the total pumping system efficiency will probably be lower.

Continuous microwave flow synthesis of mesoporous hydroxyapatite.

PubMed

Akram, Muhammad; Alshemary, Ammar Z; Goh, Yi-Fan; Wan Ibrahim, Wan Aini; Lintang, Hendrik O; Hussain, Rafaqat

2015-11-01

We have successfully used continuous microwave flow synthesis (CMFS) technique for the template free synthesis of mesoporous hydroxyapatite. The continuous microwave flow reactor consisted of a modified 2.45GHz household microwave, peristaltic pumps and a Teflon coil. This cost effective and efficient system was exploited to produce semi-crystalline phase pure nano-sized hydroxyapatite. Effect of microwave power, retention time and the concentration of reactants on the phase purity, degree of crystallinity and surface area of the final product was studied in detail. X-ray diffraction (XRD) and Fourier transform infrared spectroscopy (FTIR) were used to study the phase purity and composition of the product, while transmission electron microscopy (TEM) was used to study the effect of process parameters on the morphology of hydroxyapatite. The TEM analysis confirmed the formation of spherical particles at low microwave power; however the morphology of the particles changed to mesoporous needle and rod-like structure upon exposing the reaction mixture to higher microwave power and longer retention time inside the microwave. The in-vitro ion dissolution behavior of the as synthesized hydroxyapatite was studied by determining the amount of Ca(2+) ion released in SBF solution. Copyright © 2015 Elsevier B.V. All rights reserved.

Reactor-Scale Cultivation of the Hyperthermophilic Methanarchaeon Methanococcus jannaschii to High Cell Densities

PubMed Central

Mukhopadhyay, Biswarup; Johnson, Eric F.; Wolfe, Ralph S.

1999-01-01

For the hyperthermophilic and barophilic methanarchaeon Methanococcus jannaschii, we have developed a medium and protocols for reactor-scale cultivation that improved the final cell yield per liter from ∼0.5 to ∼7.5 g of packed wet cells (∼1.8 g dry cell mass) under autotrophic growth conditions and to ∼8.5 g of packed wet cells (∼2 g dry cell mass) with yeast extract (2 g liter−1) and tryptone (2 g liter−1) as medium supplements. For growth in a sealed bottle it was necessary to add Se to the medium, and a level of 2 μM for added Se gave the highest final cell yield. In a reactor M. jannaschii grew without added Se in the medium; it is plausible that the cells received Se as a contaminant from the reactor vessel and the H2S supply. But, for the optimal performance of a reactor culture, an addition of Se to a final concentration of 50 to 100 μM was needed. Also, cell growth in a reactor culture was inhibited at much higher Se concentrations. These observations and the data from previous work with methanogen cell extracts (B. C. McBride and R. S. Wolfe, Biochemistry 10:4312–4317, 1971) suggested that from a continuously sparged reactor culture Se was lost in the exhaust gas as volatile selenides, and this loss raised the apparent required level of and tolerance for Se. In spite of having a proteinaceous cell wall, M. jannaschii withstood an impeller tip speed of 235.5 cms−1, which was optimal for achieving high cell density and also was the higher limit for the tolerated shear rate. The organism secreted one or more acidic compounds, which lowered pH in cultures without pH control; this secretion continued even after cessation of growth. PMID:10543823

pH-oscillations in the bromate-sulfite reaction in semibatch and in gel-fed batch reactors

NASA Astrophysics Data System (ADS)

Poros, Eszter; Kurin-Csörgei, Krisztina; Szalai, István; Rábai, Gyula; Orbán, Miklós

2015-06-01

The simplest bromate oxidation based pH-oscillator, the two component BrO3--SO32- flow system was transformed to operate under semibatch and closed arrangements. The experimental preconditions of the pH-oscillations in semibatch configuration were predicted by model calculations. Using this information as guideline large amplitude (ΔpH˜3), long lasting (11-24 h) pH-oscillations accompanied with only a 20% increase of the volume in the reactor were measured when a mixture of Na2SO3 and H2SO4 was pumped into the solution of BrO3- with a very low rate. Batch-like pH-oscillations, similar in amplitude and period time appeared when the sulfite supply was substituted by its dissolution from a gel layer prepared previously in the reactor in presence of high concentration of Na2SO3. The dissolution vs time curve and the pH-oscillations in the semibatch and closed systems were successfully simulated. Due to the simplicity in composition and in experimental technique, the semibatch and batch-like BrO3--SO32- pH-oscillators may become superior to their CSTR (continuous flow stirred tank reactor) version in some present and future applications.

[Effect of gas-lift device on the morphology and performance of ANAMMOX sludge].

PubMed

Li, Xiang; Huang, Yong; Yuan, Yi; Zhou, Cheng; Chen, Zong-Heng; Zhang, Da-Lin

2014-12-01

The upflow reactor with gas-lift device was started up by inoculating ANAMMOX sludge granules of less than 0.9 mm. The effects of gas lift device system on the morphology and performance of ANAMMOX sludge were studied by using the nitrogen gas produced in ANAMMOX to drive the effluent circulation in the reactor. The results showed that, the airlift circulation function was not clear in the startup stage of the reactor, because the nitrogen gas production was very low. At the same time, the ANAMMOX granular sludge was easy to condensate. When the load rate of nitrogen removal reached 3.4 kg x (m3 x d)(-1), the function of gas lift was significant, resulting in gradually increased effluent self-circulation, and the granules were dispersed and grew gradually. After 183d of operation, the granular sludge was dominated by the granules with sizes of 1.6-2.5 mm, which accounted for 53.2% of the total sludge volume. The MLVSS content increased with the increase of sludge particle size. The gas lift device had the same function as the external reflux pump, and was helpful for sludge granulation in the ANAMMOX reactor, while reducing power consumption and the cost of the equipment.

Investigation of a submerged membrane reactor for continuous biomass hydrolysis

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

Malmali, Mohammadmahdi; Stickel, Jonathan; Wickramasinghe, S. Ranil

Enzymatic hydrolysis of cellulose is one of the most costly steps in the bioconversion of lignocellulosic biomass. Use of a submerged membrane reactor has been investigated for continuous enzymatic hydrolysis of cellulose thus allowing for greater use of the enzyme compared to a batch process. Moreover, the submerged 0.65 μm polyethersulfone microfiltration membrane avoids the need to pump a cellulose slurry through an external loop. Permeate containing glucose is withdrawn at pressures slightly below atmospheric pressure. The membrane rejects cellulose particles and cellulase enzyme bound to cellulose. Our proof-of-concept experiments have been conducted using a modified, commercially available membrane filtrationmore » cell under low fluxes around 75 L/(m2 h). The operating flux is determined by the rate of glucose production. Maximizing the rate of glucose production involves optimizing mixing, reactor holding time, and the time the feed is held in the reactor prior to commencement of membrane filtration and continuous operation. When we maximize glucose production rates it will require that we operate it at low glucose concentration in order to minimize the adverse effects of product inhibition. Consequently practical submerged membrane systems will require a combined sugar concentration step in order to concentrate the product sugar stream prior to fermentation.« less

Improving online risk assessment with equipment prognostics and health monitoring

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

Coble, Jamie B.; Liu, Xiaotong; Briere, Chris

The current approach to evaluating the risk of nuclear power plant (NPP) operation relies on static probabilities of component failure, which are based on industry experience with the existing fleet of nominally similar light water reactors (LWRs). As the nuclear industry looks to advanced reactor designs that feature non-light water coolants (e.g., liquid metal, high temperature gas, molten salt), this operating history is not available. Many advanced reactor designs use advanced components, such as electromagnetic pumps, that have not been used in the US commercial nuclear fleet. Given the lack of rich operating experience, we cannot accurately estimate the evolvingmore » probability of failure for basic components to populate the fault trees and event trees that typically comprise probabilistic risk assessment (PRA) models. Online equipment prognostics and health management (PHM) technologies can bridge this gap to estimate the failure probabilities for components under operation. The enhanced risk monitor (ERM) incorporates equipment condition assessment into the existing PRA and risk monitor framework to provide accurate and timely estimates of operational risk.« less

Core cooling under accident conditions at the high-flux beam reactor

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

Tichler, P.; Cheng, L.; Fauske, H.

The High-Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is cooled and moderated by heavy water and contains {sup 235}U in the form of narrow-channel, parallel-plate-type fuel elements. During normal operation, the flow direction is downward through the core. This flow direction is maintained at a reduced flow rate during routine shutdown and on loss of commercial power by means of redundant pumps and power supplies. However, in certain accident scenarios, e.g. loss-of-coolant accidents (LOCAs), all forced-flow cooling is lost. Although there was experimental evidence during the reactor design period (1958-1963) that the heat removal capacity in the fullymore » developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. Accordingly, a test program was initiated using an electrically heated section to simulate the fuel channel and a cooling loop to simulate the balance of the primary cooling system.« less

Graphite distortion ``C`` Reactor. Final report

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

Wood, N.H.

1962-02-08

This report covers the efforts of the Laboratory in an investigation of the graphite distortion in the ``C`` reactor at Hanford. The particular aspects of the problem to be covered by the Laboratory were possible ``fixes`` to the control rod sticking problem caused by VSR channel distortion.

Decommissioning of the Northrop TRIGA reactor

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

Cozens, George B.; Woo, Harry; Benveniste, Jack

1986-07-01

An overview of the administrative and operational aspects of decommissioning and dismantling the Northrop Mark F TRIGA Reactor, including: planning and preparation, personnel requirements, government interfacing, costs, contractor negotiations, fuel shipments, demolition, disposal of low level waste, final survey and disposition of the concrete biological shielding. (author)

Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

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

Fromm, Bradley; Hauch, Benjamin; Sridharan, Kumar

2016-12-01

A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulationmore » tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.« less

Synthesis of carbohydrates in a continuous flow reactor by immobilized phosphatase and aldolase.

PubMed

Babich, Lara; Hartog, Aloysius F; van Hemert, Lieke J C; Rutjes, Floris P J T; Wever, Ron

2012-12-01

Herein, we report a new flow process with immobilized enzymes to synthesize complex chiral carbohydrate analogues from achiral inexpensive building blocks in a three-step cascade reaction. The first reactor contained immobilized acid phosphatase, which phosphorylated dihydroxyacetone to dihydroxyacetone phosphate using pyrophosphate as the phosphate donor. The second flow reactor contained fructose-1,6-diphosphate aldolase (RAMA, rabbit muscle aldolase) or rhamnulose-1-phosphate aldolase (RhuA from Thermotoga maritima) and acid phosphatase. The immobilized aldolases coupled the formed dihydroxyacetone phosphate to aldehydes, resulting in phosphorylated carbohydrates. A final reactor containing acid phosphatase that dephosphorylated the phosphorylated product yielded the final product. Different aldehydes were used to synthesize carbohydrates on a gram scale. To demonstrate the feasibility of the flow systems, we synthesized 0.6 g of the D-fagomine precursor. By using immobilized aldolase RhuA we were also able to obtain other stereoisomers of the D-fagomine precursor. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Process and apparatus for coal hydrogenation

DOEpatents

Ruether, John A.

1988-01-01

In a coal liquefaction process an aqueous slurry of coal is prepared containing a dissolved liquefaction catalyst. A small quantity of oil is added to the slurry and then coal-oil agglomerates are prepared by agitation of the slurry at atmospheric pressure. The resulting mixture of agglomerates, excess water, dissolved catalyst, and unagglomerated solids is pumped to reaction pressure and then passed through a drainage device where all but a small amount of surface water is removed from the agglomerates. Sufficient catalyst for the reaction is contained in surface water remaining on the agglomerates. The agglomerates fall into the liquefaction reactor countercurrently to a stream of hot gas which is utilized to dry and preheat the agglomerates as well as deposit catalyst on the agglomerates before they enter the reactor where they are converted to primarily liquid products under hydrogen pressure.

Method and apparatus to produce and maintain a thick, flowing, liquid lithium first wall for toroidal magnetic confinement DT fusion reactors

DOEpatents

Woolley, Robert D.

2002-01-01

A system for forming a thick flowing liquid metal, in this case lithium, layer on the inside wall of a toroid containing the plasma of a deuterium-tritium fusion reactor. The presence of the liquid metal layer or first wall serves to prevent neutron damage to the walls of the toroid. A poloidal current in the liquid metal layer is oriented so that it flows in the same direction as the current in a series of external magnets used to confine the plasma. This current alignment results in the liquid metal being forced against the wall of the toroid. After the liquid metal exits the toroid it is pumped to a heat extraction and power conversion device prior to being reentering the toroid.

Design of Test Loops for Forced Convection Heat Transfer Studies at Supercritical State

NASA Astrophysics Data System (ADS)

Balouch, Masih N.

Worldwide research is being conducted to improve the efficiency of nuclear power plants by using supercritical water (SCW) as the working fluid. One such SCW reactor considered for future development is the CANDU-Supercritical Water Reactor (CANDU-SCWR). For safe and accurate design of the CANDU-SCWR, a detailed knowledge of forced-convection heat transfer in SCW is required. For this purpose, two supercritical fluid loops, i.e. a SCW loop and an R-134a loop are developed at Carleton University. The SCW loop is designed to operate at pressures as high as 28 MPa, temperatures up to 600 °C and mass fluxes of up to 3000 kg/m2s. The R-134a loop is designed to operate at pressures as high as 6 MPa, temperatures up to 140 °C and mass fluxes in the range of 500-6000 kg/m2s. The test loops designs allow for up to 300 kW of heating power to be imparted to the fluid. Both test loops are of the closed-loop design, where flow circulation is achieved by a centrifugal pump in the SCW loop and three parallel-connected gear pumps in the R-134a loop, respectively. The test loops are pressurized using a high-pressure nitrogen cylinder and accumulator assembly, which allows independent control of the pressure, while simultaneously dampening pump induced pressure fluctuations. Heat exchangers located upstream of the pumps control the fluid temperature in the test loops. Strategically located measuring instrumentation provides information on the flow rate, pressure and temperature in the test loops. The test loops have been designed to accommodate a variety of test-section geometries, ranging from a straight circular tube to a seven-rod bundle, achieving heat fluxes up to 2.5 MW/m2 depending on the test-section geometry. The design of both test loops allows for easy reconfiguration of the test-section orientation relative to the gravitational direction. All the test sections are of the directly-heated design, where electric current passing through the pressure retaining walls of the test sections provides the Joule heating required to heat up the fluid to supercritical conditions. A high-temperature dielectric gasket isolates the current carrying parts of the test section from the rest of the assembly. Temperature and pressure drop data are collected at the inlet and outlet, and along the heated length of the test section. The test loops and test sections are designed according to American Society of Mechanical Engineers (ASME) Pressure Piping B31.1, and Boiler and Pressure Vessel Code, Section VIII-Division 1 rules. The final test loops and test sections assemblies are certified by Technical Standards and Safety Authority (TSSA). Every attempt is made to use off-the-shelf components where possible in order to streamline the design process and reduce costs. Following a rigorous selection process, stainless steel Types 316 and 316H are selected as the construction materials for the test loops, and Inconel 625 is selected as the construction material for the test sections. This thesis describes the design of the SCW and R-134a loops along with the three test-section geometries (i.e., tubular, annular and bundle designs).

Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

PubMed

Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

2011-01-01

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. Copyright © 2010 Elsevier Ltd. All rights reserved.

Computation of water hammer protection of modernized pumping station

NASA Astrophysics Data System (ADS)

Himr, Daniel

2014-03-01

Pumping station supplies water for irrigation. Maximal capacity 2 × 1.2m3·s-1 became insufficient, thus it was upgraded to 2 × 2m3·s-1. Paper is focused on design of protection against water hammer in case of sudden pumps trip. Numerical simulation of the most dangerous case (when pumps are giving the maximal flow rate) showed that existing air vessels were not able to protect the system and it would be necessary to add new vessels. Special care was paid to influence of their connection to the main pipeline, because the resistance of the connection has a significant impact on the scale of pressure pulsations. Finally, the pump trip was performed to verify if the system worked correctly. The test showed that pressure pulsations are lower (better) than computation predicted. This discrepancy was further analysed.

Determinants of respiratory pump function in patients with cystic fibrosis.

PubMed

Dassios, Theodore

2015-01-01

Respiratory failure constitutes the major cause of morbidity and mortality in patients with Cystic Fibrosis (CF). Respiratory failure could either be due to lung parenchyma damage or to insufficiency of the respiratory pump which consists of the respiratory muscles, the rib cage and the neuromuscular transmission pathways. Airway obstruction, hyperinflation and malnutrition have been historically recognised as the major determinants of respiratory pump dysfunction in CF. Recent research has identified chronic infection, genetic predisposition, dietary and pharmaceutical interventions as possible additional determinants of this impairment. Furthermore, new methodological approaches in assessing respiratory pump function have led to a better understanding of the pathogenesis of respiratory pump failure in CF. Finally, respiratory muscle function could be partially preserved in CF patients with structured interventions such as aerobic exercise, inspiratory muscle training and non-invasive ventilation and CF patients could consequently be relatively protected from respiratory fatigue and respiratory failure. Copyright © 2014 Elsevier Ltd. All rights reserved.

TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

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

Ryan, B.C.

1997-05-01

This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely themore » total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.« less

10 CFR 52.171 - Finality of manufacturing licenses; information requests.

Code of Federal Regulations, 2010 CFR

2010-01-01

... required for issuance of a construction permit, operating license, combined license, in any hearing under... manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases for not incorporating severe accident mitigation design alternatives into the design of the reactor to...

Safety approach to the selection of design criteria for the CRBRP reactor refueling system

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

Meisl, C J; Berg, G E; Sharkey, N F

1979-01-01

The selection of safety design criteria for Liquid Metal Fast Breeder Reactor (LMFBR) refueling systems required the extrapolation of regulations and guidelines intended for Light Water Reactor refueling systems and was encumbered by the lack of benefit from a commercially licensed predecessor other than Fermi. The overall approach and underlying logic are described for developing safety design criteria for the reactor refueling system (RRS) of the Clinch River Breeder Reactor Plant (CRBRP). The complete selection process used to establish the criteria is presented, from the definition of safety functions to the finalization of safety design criteria in the appropriate documents.more » The process steps are illustrated by examples.« less

Wireless Sensor for Measuring Pump Efficiency: Small Business Voucher Project with KCF Technologies

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

Fugate, David L.; Liu, Xiaobing; Gehl, Anthony C.

This document is to fulfill the final report requirements for the Small Business Voucher (SBV) CRADA project with ORNL and KCF Technologies (CRADA/NFE-16-06133). Pumping systems account for nearly 20% of the world’s electrical energy demand and range from 25-50% of the energy usage in many industrial and building power plants. The energy cost is the largest element in the total cost of owning a pump (~40%). In response to a recent DOE mandate for improved pump efficiency pump manufacturers are preparing for the changes that the impending regulations will bring, including design improvements. This mandate also establishes a need formore » new low cost pump efficiency measurement systems. The standard industry definition of pump efficiency is the mechanical water horsepower delivered divided by the electrical horsepower input to the motor. KCF Technologies has developed a new sensor measurement technique to estimate fluid pump efficiency using a thermodynamic approach. KCF Technologies applied for a SBV grant with ORNL as the research partner. KCF needed a research partner with the proper facilities to demonstrate the efficacy of its wireless sensor unit for measuring pump efficiency. The ORNL Building Technologies Research and Integration Center (BTRIC) test resources were used to test and demonstrate the successful measurement of pump efficiency with the KCF sensor technology. KCF is now working on next steps to commercialize this sensing technology.« less

Assessing the impact of electrolyte conductivity and viscosity on the reactor cost and pressure drop of redox-active polymer flow batteries

NASA Astrophysics Data System (ADS)

Iyer, Vinay A.; Schuh, Jonathon K.; Montoto, Elena C.; Pavan Nemani, V.; Qian, Shaoyi; Nagarjuna, Gavvalapalli; Rodríguez-López, Joaquín; Ewoldt, Randy H.; Smith, Kyle C.

2017-09-01

Redox-active small molecules, used traditionally in redox flow batteries (RFBs), are susceptible to crossover and require expensive ion exchange membranes (IEMs) to achieve long lifetimes. Redox-active polymer (RAP) solutions show promise as candidate electrolytes to mitigate crossover through size-exclusion, enabling the use of porous separators instead of IEMs. Here, poly(vinylbenzyl ethyl viologen) is studied as a surrogate RAP for RFBs. For oxidized RAPs, ionic conductivity varies weakly between 1.6 and 2.1 S m-1 for RAP concentrations of 0.13-1.27 mol kg-1 (monomeric repeat unit per kg solvent) and 0.32 mol kg-1 LiBF4 with a minor increase upon reduction. In contrast, viscosity varies between 1.8 and 184.0 mPa s over the same concentration range with weakly shear-thinning rheology independent of oxidation state. Techno-economic analysis is used to quantify reactor cost as a function of electrolyte transport properties for RAP concentrations of 0.13-1.27 mol kg-1, assuming a hypothetical 3V cell and facile kinetics. Among these concentrations, reactor cost is minimized over a current density range of 600-1000 A m-2 with minimum reactor cost between 11-17 per kWh, and pumping pressures below 10 kPa. The predicted low reactor cost of RAP RFBs is enabled by sustained ionic mobility in spite of the high viscosity of concentrated RAP solutions.

Reactor-Produced Medical Radionuclides

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

Mirzadeh, Saed; Mausner, Leonard; Garland, Marc A

2011-01-01

The therapeutic use of radionuclides in nuclear medicine, oncology and cardiology is the most rapidly growing use of medical radionuclides. Since most therapeutic radionuclides are neutron rich and decay by beta emission, they are reactor-produced. This chapter deals mainly with production approaches with neutrons. Neutron interactions with matter, neutron transmission and activation rates, and neutron spectra of nuclear reactors are discussed in some detail. Further, a short discussion of the neutron-energy dependence of cross sections, reaction rates in thermal reactors, cross section measurements and flux monitoring, and general equations governing the reactor production of radionuclides are presented. Finally, the chaptermore » is concluded by providing a number of examples encompassing the various possible reaction routes for production of a number of medical radionuclides in a reactor.« less

Energy Distribution of Electrons in Radiation Induced-Helium Plasmas. Ph.D. Thesis

NASA Technical Reports Server (NTRS)

Lo, R. H.

1972-01-01

Energy distribution of high energy electrons as they slow down and thermalize in a gaseous medium is studied. The energy distribution in the entire energy range from source energies down is studied analytically. A helium medium in which primary electrons are created by the passage of heavy-charged particles from nuclear reactions is emphasized. A radiation-induced plasma is of interest in a variety of applications, such as radiation pumped lasers and gaseous core nuclear reactors.

SP-100 Program: space reactor system and subsystem investigations

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

Harty, R.B.

1983-09-30

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

SP-100 program: Space reactor system and subsystem investigations

NASA Astrophysics Data System (ADS)

Harty, R. B.

1983-09-01

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. The nuclear safety review/approval process that is required for a space reactor system is summarized. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that is expected and to provide information that could be usable in future programs.

Integrated microfluidic systems for cell lysis, mixing/pumping and DNA amplification

NASA Astrophysics Data System (ADS)

Lee, Chia-Yen; Lee, Gwo-Bin; Lin, Jr-Lung; Huang, Fu-Chun; Liao, Chia-Sheng

2005-06-01

The present paper reports a fully automated microfluidic system for the DNA amplification process by integrating an electroosmotic pump, an active micromixer and an on-chip temperature control system. In this DNA amplification process, the cell lysis is initially performed in a micro cell lysis reactor. Extracted DNA samples, primers and reagents are then driven electroosmotically into a mixing region where they are mixed by the active micromixer. The homogeneous mixture is then thermally cycled in a micro-PCR (polymerase chain reaction) chamber to perform DNA amplification. Experimental results show that the proposed device can successfully automate the sample pretreatment operation for DNA amplification, thereby delivering significant time and effort savings. The new microfluidic system, which facilitates cell lysis, sample driving/mixing and DNA amplification, could provide a significant contribution to ongoing efforts to miniaturize bio-analysis systems by utilizing a simple fabrication process and cheap materials.

10 CFR 50.82 - Termination of license.

Code of Federal Regulations, 2011 CFR

2011-01-01

... to the NRC, consistent with the requirements of § 50.4(b)(8); (ii) Once fuel has been permanently... fuel from the reactor vessel, or when a final legally effective order to permanently cease operations... emplacement or retention of fuel into the reactor vessel. (3) Decommissioning will be completed within 60...

Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 2, Final report

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

Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

1992-03-01

This report contains a review of design studies for Inertial Confinement reactor. This second of three volumes discussions is some detail the following: Objectives, requirements, and assumptions; rationale for design option selection; key technical issues and R&D requirements; and conceptual design selection and description.

Bioaugmentation of activated sludge towards 3-chloroaniline removal with a mixed bacterial population carrying a degradative plasmid.

PubMed

Bathe, Stephan; Schwarzenbeck, Norbert; Hausner, Martina

2009-06-01

A bioaugmentation approach combining several strategies was applied to achieve degradation of 3-chloroaniline (3CA) in semicontinuous activated sludge reactors. In a first step, a 3CA-degrading Comamonas testosteroni strain carrying the degradative plasmid pNB2 was added to a biofilm reactor, and complete 3CA degradation together with spread of the plasmid within the indigenous biofilm population was achieved. A second set of reactors was then bioaugmented with either a suspension of biofilm cells removed from the carrier material or with biofilm-containing carrier material. 3CA degradation was established rapidly in all bioaugmented reactors, followed by a slow adaptation of the non-bioaugmented control reactors. In response to variations in 3CA concentration, all reactors exhibited temporary performance breakdowns. Whereas duplicates of the control reactors deviated in their behaviour, the bioaugmented reactors appeared more reproducible in their performance and population dynamics. Finally, the carrier-bioaugmented reactors showed an improved performance in the presence of high 3CA influent concentrations over the suspension-bioaugmented reactors. In contrast, degradation in one control reactor failed completely, but was rapidly established in the remaining control reactor.

Design, Construction and Testing of an In-Pile Loop for PWR (Pressurized Water Reactor) Simulation.

DTIC Science & Technology

1987-06-01

computer modeling remains at best semiempirical (C-i), this large variation in scaling factor makes extrapolation of data impossible. The DIDO Water...in a full scale PWR are not practical. The reactor plant is not controlled to tolerances necessary for research, and utilities are reluctant to vary...MIT Reactor Safeguards Committee, in revision 1 to the PCCL Safety Evaluation Report (SER), for final approval to begin in-pile testing and

Characterization of biofilm in 200W fluidized bed reactors

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

Lee, Michelle H.; Saurey, Sabrina D.; Lee, Brady D.

2014-09-29

Contaminated groundwater beneath the 200 West Area at the Hanford Site in Southeast Washington is currently being treated using a pump and treat system to remove organics, inorganics, radionuclides, and metals. A granular activated carbon-based fluidized bed reactor (FBR) has been added to remove nitrate, hexavalent chromium and carbon tetrachloride. Initial analytical results indicated the microorganisms effectively reduced many of the contaminants to less than cleanup levels. However shortly thereafter operational upsets of the FBR include carbon carry over, over production of microbial extracellular polymeric substance (biofilm) materials, and over production of hydrogen sulfide. As a result detailed investigations weremore » undertaken to understand the functional diversity and activity of the microbial community present in the FBR over time. Molecular analyses including terminal restriction fragment length polymorphism analysis, quantitative polymerase chain reaction and fluorescent in situ hybridization analyses were performed on the microbial community extracted from the biofilm within the bed and from the inoculum, to determine functional dynamics of the FBR bed over time and following operational changes. Findings from these analyses indicated: 1) the microbial community within the bed was completely different than community used for inoculation, and was likely from the groundwater; 2) analyses early in the testing showed an FBR community dominated by a few Curvibacter and Flavobacterium species; 3) the final sample taken indicated that the microbial community in the FBR bed had become more diverse; and 4) qPCR analyses indicated that bacteria involved in nitrogen cycling, including denitrifiers and anaerobic ammonia oxidizing bacteria, were dominant in the bed. These results indicate that molecular tools can be powerful for determining functional diversity within FBR type reactors. Coupled with micronutrient, influent and effluent chemistry evaluations, a more complete understanding of the balance between system additions (nutrients, groundwater) and biology can be achieved, thus increasing long-term predictions of performance. These analyses uniquely provide information that can be used in optimizing the overall performance, efficiency, and stability of the system both in real time as well as over the long-term, as the system design is altered or improved and/or new streams are added.« less

Pump-probe studies of radiation induced defects and formation of warm dense matter with pulsed ion beams

NASA Astrophysics Data System (ADS)

Schenkel, T.; Persaud, A.; Gua, H.; Seidl, P. A.; Waldron, W. L.; Gilson, E. P.; Kaganovich, I. D.; Davidson, R. C.; Friedman, A.; Barnard, J. J.; Minior, A. M.

2014-10-01

We report results from the 2nd generation Neutralized Drift Compression Experiment at Berkeley Lab. NDCX-II is a pulsed, linear induction accelerator designed to drive thin foils to warm dense matter (WDM) states with peak temperatures of ~ 1 eV using intense, short pulses of 1.2 MeV lithium ions. Tunability of the ion beam enables pump-probe studies of radiation effects in solids as a function of excitation density, from isolated collision cascades to the onset of phase-transitions and WDM. Ion channeling is an in situ diagnostic of damage evolution during ion pulses with a sensitivity of <0.1% displacements per atom. We will report results from damage evolution studies in thin silicon crystals with Li + and K + beams. Detection of channeled ions tracks lattice disorder evolution with a resolution of ~ 1 ns using fast current measurements. We will discuss pump-probe experiments with pulsed ion beams and the development of diagnostics for WDM and multi-scale (ms to fs) access to the materials physics of collision cascades e.g. in fusion reactor materials. Work performed under auspices of the US DOE under Contract No. DE-AC02-05CH11231.

Organizational, Cultural, and Psychological Determinants of Smart Infusion Pump Work Arounds: A Study of 3 U.S. Health Systems.

PubMed

Dunford, Benjamin B; Perrigino, Matthew; Tucker, Sharon J; Gaston, Cynthia L; Young, Jim; Vermace, Beverly J; Walroth, Todd A; Buening, Natalie R; Skillman, Katherine L; Berndt, Dawn

2017-09-01

We investigated nurse perceptions of smart infusion medication pumps to provide evidence-based insights on how to help reduce work around and improve compliance with patient safety policies. Specifically, we investigated the following 3 research questions: (1) What are nurses' current attitudes about smart infusion pumps? (2) What do nurses think are the causes of smart infusion pump work arounds? and (3) To whom do nurses turn for smart infusion pump training and troubleshooting? We surveyed a large number of nurses (N = 818) in 3 U.S.-based health care systems to address the research questions above. We assessed nurses' opinions about smart infusion pumps, organizational perceptions, and the reasons for work arounds using a voluntary and anonymous Web-based survey. Using qualitative research methods, we coded open-ended responses to questions about the reasons for work arounds to organize responses into useful categories. The nurses reported widespread satisfaction with smart infusion pumps. However, they reported numerous organizational, cultural, and psychological causes of smart pump work arounds. Of 1029 open-ended responses to the question "why do smart pump work arounds occur?" approximately 44% of the causes were technology related, 47% were organization related, and 9% were related to individual factors. Finally, an overwhelming majority of nurses reported seeking solutions to smart pump problems from coworkers and being trained primarily on the job. Hospitals may significantly improve adherence to smart pump safety features by addressing the nontechnical causes of work arounds and by providing more leadership and formalized training for resolving smart pump-related problems.

Operational restoration of the Pen Branch bottomland hardwood and swamp wetlands - the research setting

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

Nelson, E.A.

2000-01-05

The Savannah River Swamp is a 3020 Ha forested wetland on the floodplain of the Savannah River and is located on the Department of Energy's Savannah River Site (SRS) near Aiken, SC. Historically the swamp consisted of approximately 50 percent bald cypress-water tupelo stands, 40 percent mixed bottomland hardwood stands, and 10 percent shrub, marsh, and open water. Creek corridors were typical of Southeastern bottomland hardwood forests. The hydrology was controlled by flooding of the Savannah River and by flow from four creeks that drain into the swamp prior to flow into the Savannah River. Upstream dams have caused somemore » alteration of the water levels and timing of flooding within the floodplain. Major impacts to the swamp hydrology occurred with the completion of the production reactors and one coal-fired powerhouse at the SRS in the early 1950's. Water was pumped from the Savannah River, through secondary heat exchangers of the reactors, and discharged into three of the tributary streams that flow into the swamp. Flow in one of the tributaries, Pen Branch, was typically 0.3 m3 s-1 (10-20) cfs prior to reactor pumping and 11.0 m3 s-1 (400 cfs) during pumping. This continued from 1954 to 1988 at various levels. The sustained increases in water volume resulted in overflow of the original stream banks and the creation of additional floodplains. Accompanying this was considerable erosion of the original stream corridor and deposition of a deep silt layer on the newly formed delta. Heated water was discharged directly into Pen Branch and water temperature in the stream often exceeded 65 degrees C. The nearly continuous flooding of the swamp, the thermal load of the water, and the heavy silting resulted in complete mortality of the original vegetation in large areas of the floodplain. In the years since pumping was reduced, early succession has begun in some affected areas. Most of this has been herbs, grasses, and shrubs. Areas that have seedlings are generally willow thickets that support a lower diversity of wildlife. No volunteer seedlings of heavy-seeded hardwoods or cypress have been found in the corridor areas. Research was conducted to determine methods to reintroduce tree species characteristic of more mature forested wetlands. Three restoration strategies were formulated to deal with the differing conditions of the Upper Corridor, the Lower Corridor, and the Delta regions of the impacted area. Site preparation and planting of each area with mixtures of tree species were carried out to speed the restoration of the ecosystem. Species composition and selection were altered based on the current and expected hydrological regimes that the reforestation areas will be experiencing. Because of the operational design of the restoration project, a research program naturally followed to document the success. Many of those efforts are detailed here.« less

Fiber Attachment Module Experiment (FAME): Using a Multiplexed Miniature Hollow Fiber Membrane Bioreactor Solution for Rapid Process Testing

NASA Astrophysics Data System (ADS)

Lunn, Griffin; Wheeler, Raymond; Hummerick, Mary; Birmele, Michele; Richards, Jeffrey; Coutts, Janelle; Koss, Lawrence; Spencer, Lashelle.; Johnsey, Marissa; Ellis, Ronald

Bioreactor research, even today, is mostly limited to continuous stirred-tank reactors (CSTRs). These are not an option for microgravity applications due to the lack of a gravity gradient to drive aeration as described by the Archimedes principle. This has led to testing of Hollow Fiber Membrane Bioreactors (HFMBs) for microgravity applications, including possible use for wastewater treatment systems for the International Space Station (ISS). Bioreactors and filtration systems for treating wastewater could avoid the need for harsh pretreatment chemicals and improve overall water recovery. However, the construction of these reactors is difficult and commercial off-the-shelf (COTS) versions do not exist in small sizes. We have used 1-L modular HFMBs in the past, but the need to perform rapid testing has led us to consider even smaller systems. To address this, we designed and built 125-mL, rectangular reactors, which we have called the Fiber Attachment Module Experiment (FAME) system. A polycarbonate rack of four square modules was developed with each module containing removable hollow fibers. Each FAME reactor is self-contained and can be easily plumbed with peristaltic and syringe pumps for continuous recycling of fluids and feeding, as well as fitted with sensors for monitoring pH, dissolved oxygen, and gas measurements similar to their larger counterparts. The first application tested in the FAME racks allowed analysis of over a dozen fiber surface treatments and three inoculation sources to achieve rapid reactor startup and biofilm attachment (based on carbon oxidation and nitrification of wastewater). With these miniature FAME reactors, data for this multi-factorial test were collected in duplicate over a six-month period; this greatly compressed time period required for gathering data needed to study and improve bioreactor performance.

Using Additive Manufacturing to Optimize FLiBe Coolant Blanket in Fusion Reactors

NASA Astrophysics Data System (ADS)

Fry, Vincent Michael

Fusion reactors have often been hailed as the holy grail of clean energy generation, though a power-generating reactor has never been built due to a multitude of limiting factors. One such factor is the immense 12-15 MW/m2 heat fluxes experienced by the inner wall of the reactor. Multiple groups have proposed the use of tungsten swirl tubes to withstand the heat generated within the reactor core. The primary focus of this investigation is to parameterize this 'first wall' interior structure to determine the highest achievable heat transfer coefficient given the many tungsten configurations enabled via additive manufacturing. Two general tube structures were considered: an orthogonal three-dimensional mesh of various diameters and spacings, as well as a swirl tube geometry with varying 'tape' thicknesses. The coolant liquid proposed is FLiBe (2LiF-BeF2) due to its high specific heat capacity as well as its ability to breed tritium, the fuel for the reactor. This was accomplished using theoretical calculations; computational fluid dynamics and conjugate heat transfer simulations in ANSYS Workbench; as well as an experimental setup to confirm tube pressure drop along the pipe. It was determined that heat transfer coefficients between upwards of 60,000 W/m 2K were readily achievable, keeping the first wall temperature around 1300 K. A multitude of designs proved to be feasible given the pumping power restrictions, though the suggested design going forward is a swirl tube with 2 mm 'tape' thickness and 3 m/s inlet velocity. Simulated pressure drop with water was accurate to within 30% of experimentally measured values, giving confidence in the credibility of the results.

Modeling the transport of nitrogen in an NPP-2006 reactor circuit

NASA Astrophysics Data System (ADS)

Stepanov, O. E.; Galkin, I. Yu.; Sledkov, R. M.; Melekh, S. S.; Strebnev, N. A.

2016-07-01

Efficient radiation protection of the public and personnel requires detecting an accident-initiating event quickly. Specifically, if a heat-exchange tube in a steam generator is ruptured, the 16N radioactive nitrogen isotope, which contributes to a sharp increase in the steam activity before the turbine, may serve as the signaling component. This isotope is produced in the core coolant and is transported along the circulation circuit. The aim of the present study was to model the transport of 16N in the primary and the secondary circuits of a VVER-1000 reactor facility (RF) under nominal operation conditions. KORSAR/GP and RELAP5/Mod.3.2 codes were used to perform the calculations. Computational models incorporating the major components of the primary and the secondary circuits of an NPP-2006 RF were constructed. These computational models were subjected to cross-verification, and the calculation results were compared to the experimental data on the distribution of the void fraction over the steam generator height. The models were proven to be valid. It was found that the time of nitrogen transport from the core to the heat-exchange tube leak was no longer than 1 s under RF operation at a power level of 100% N nom with all primary circuit pumps activated. The time of nitrogen transport from the leak to the γ-radiation detection unit under the same operating conditions was no longer than 9 s, and the nitrogen concentration in steam was no less than 1.4% (by mass) of its concentration at the reactor outlet. These values were obtained using conservative approaches to estimating the leak flow and the transport time, but the radioactive decay of nitrogen was not taken into account. Further research concerned with the calculation of thermohydraulic processes should be focused on modeling the transport of nitrogen under RF operation with some primary circuit pumps deactivated.

Heat pump assisted geothermal heating system for Felix Spa, Romania

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

Rosca, Marcel; Maghiar, Teodor

1996-01-24

The paper presents a pre-feasibility type study of a proposed heat pump assisted geothermal heating system for an average hotel in Felix Spa, Romania. After a brief presentation of the geothermal reservoir, the paper gives the methodology and the results of the technical and economical calculations. The technical and economical viability of the proposed system is discussed in detail in the final part of the paper.

Heat pump assisted geothermal heating system for Felix Spa, Romania

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

Rosca, M.; Maghiar, T.

1996-12-31

The paper presents a pre-feasibility type study of a proposed heat pump assisted geothermal heating system for an average hotel in Felix Spa, Romania. After a brief presentation of the geothermal reservoir, the paper gives the methodology and the results of the technical and economical calculations. The technical and economical viability of the proposed system is discussed in detail in the final part of the paper.

Next Generation Electromagnetic Pump Analysis Tools (PLM DOC-0005-2188). Final Report

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

Stregy, Seth; Dasilva, Ana; Yilmaz, Serkan

2015-10-29

This report provides the broad historical review of EM Pump development and details of MATRIX development under this project. This report summarizes the efforts made to modernize the legacy performance models used in previous EM Pump designs and the improvements made to the analysis tools. This report provides information on Tasks 1, 3, and 4 of the entire project. The research for Task 4 builds upon Task 1: Update EM Pump Databank and Task 3: Modernize the Existing EM Pump Analysis Model, which are summarized within this report. Where research for Task 2: Insulation Materials Development and Evaluation identified parametersmore » applicable to the analysis model with Task 4, the analysis code was updated, and analyses were made for additional materials. The important design variables for the manufacture and operation of an EM Pump that the model improvement can evaluate are: space constraints; voltage capability of insulation system; maximum flux density through iron; flow rate and outlet pressure; efficiency and manufacturability. The development of the next-generation EM Pump analysis tools during this two-year program provides information in three broad areas: Status of analysis model development; Improvements made to older simulations; and Comparison to experimental data.« less

River Gardens Intermediate-Care Facility water-to-air heating and air-conditioning demonstration project. Final report

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

Brown, R.C.

An integrated system of heat pumps is used to reject heat into or extract heat from circulating water from a shallow well adjacent to the river to demonstrate the efficiency and fuel cost savings of water-to-air heat pumps, without the expense of drilling a deep well. Water is returned unpolluted to the Guadalupe River and is circulated through a five-building complex at River Gardens Intermediate Care Facility for the Mentally Retarded in New Braunfels, Texas. The water is used as a heat source or sink for 122 heat pumps providing space heating and cooling, and for refrigeration and freezer units.more » The system was not installed as designed, which resulted in water pumping loads being higher than the original design. Electrical consumption for pumping water represented 36 to 37% of system electrical consumption. Without the water pumping load, the water-to-air system was an average of 25% more efficient in heating than a comparable air-to-air unit with resistance heating. With water pumping load included, the installed system averaged 17% less efficient in cooling and 19% more efficient in heating than the comparable unit.« less

A method for gear fatigue life prediction considering the internal flow field of the gear pump

NASA Astrophysics Data System (ADS)

Shen, Haidong; Li, Zhiqiang; Qi, Lele; Qiao, Liang

2018-01-01

Gear pump is the most widely used volume type hydraulic pump, and it is the main power source of the hydraulic system. Its performance is influenced by many factors, such as working environment, maintenance, fluid pressure and so on. It is different from the gear transmission system, the internal flow field of gear pump has a greater impact on the gear life, therefore it needs to consider the internal hydraulic system when predicting the gear fatigue life. In this paper, a certain aircraft gear pump as the research object, aim at the typical failure forms, gear contact fatigue, of gear pump, proposing the prediction method based on the virtual simulation. The method use CFD (Computational fluid dynamics) software to analyze pressure distribution of internal flow field of the gear pump, and constructed the unidirectional flow-solid coupling model of gear to acquire the contact stress of tooth surface on Ansys workbench software. Finally, employing nominal stress method and Miner cumulative damage theory to calculated the gear contact fatigue life based on modified material P-S-N curve. Engineering practice show that the method is feasible and efficient.

DE-NE0008277_PROTEUS final technical report 2018

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

Enqvist, Andreas

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce suchmore » a comprehensive suite of experimental data to support any future GCFR endeavors.« less

76 FR 37407 - Energy Conservation Program: Energy Conservation Standards for Residential Furnaces and...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-27

...The Energy Policy and Conservation Act of 1975 (EPCA), as amended, prescribes energy conservation standards for various consumer products and certain commercial and industrial equipment, including residential furnaces and residential central air conditioners and heat pumps. EPCA also requires the U.S. Department of Energy (DOE) to determine whether more-stringent, amended standards for these products would be technologically feasible and economically justified, and would save a significant amount of energy. In this direct final rule, DOE adopts amended energy conservation standards for residential furnaces and for residential central air conditioners and heat pumps. A notice of proposed rulemaking that proposes identical energy efficiency standards is published elsewhere in this issue of the Federal Register. If DOE receives adverse comment and determines that such comment may provide a reasonable basis for withdrawing the direct final rule, this final rule will be withdrawn, and DOE will proceed with the proposed rule.

76 FR 37549 - Energy Conservation Program: Energy Conservation Standards for Residential Furnaces and...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-27

...The Energy Policy and Conservation Act of 1975 (EPCA), as amended, prescribes energy conservation standards for various consumer products and certain commercial and industrial equipment, including residential furnaces and residential central air conditioners and heat pumps. EPCA also requires the U.S. Department of Energy (DOE) to determine whether more-stringent, amended standards for these products would be technologically feasible and economically justified, and would save a significant amount of energy. In this notice, DOE proposes energy conservation standards for residential furnaces and for residential central air conditioners and heat pumps identical to those set forth in a direct final rule published elsewhere in today's Federal Register. If DOE receives adverse comment and determines that such comment may provide a reasonable basis for withdrawing the direct final rule, DOE will publish a notice withdrawing the direct final rule and will proceed with this proposed rule.

Calculation to experiment comparison of SPND signals in various nuclear reactor environments

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

Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien

2015-07-01

In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first partmore » of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)« less

Final environmental statement, Liquid Metal Fast Breeder Reactor Program. Volume 1

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

None

1975-12-01

Information is presented under the following section headings: LMFBR program options and their compatibility with the major issues affecting commercial development, Proposed Final Environmental Statement for the LMFBR program, December 1974, WASH-1535, supplemental material, and material relating to Proposed Final Environmental Statement review. (DG)

Design and development of a prototype wet oxidation system for the reclamation of water and the disposition of waste residues onboard space vehicles

NASA Technical Reports Server (NTRS)

Jagow, R. B.

1972-01-01

Laboratory investigations to define optimum process conditions for oxidation of fecal/urine slurries were conducted in a one-liter batch reactor. The results of these tests formed the basis for the design, fabrication, and testing of an initial prototype system, including a 100-hour design verification test. Areas of further development were identified during this test. Development of a high pressure slurry pump, materials corrosion studies, oxygen supply trade studies, comparison of salt removal water recovery devices, ammonia removal investigation, development of a solids grinder, reactor design studies and bearing life tests, and development of shutoff valves and a back pressure regulator were undertaken. The development work has progressed to the point where a prototype system suitable for manned chamber testing can be fabricated and tested with a high degree of confidence of success.

Design consideration for a nuclear electric propulsion system

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Pawlik, E. V.

1978-01-01

A study is currently underway to design a nuclear electric propulsion vehicle capable of performing detailed exploration of the outer-planets. Primary emphasis is on the power subsystem. Secondary emphasis includes integration into a spacecraft, and integration with the thrust subsystem and science package or payload. The results of several design iterations indicate an all-heat-pipe system offers greater reliability, elimination of many technology development areas and a specific weight of under 20 kg/kWe at the 400 kWe power level. The system is compatible with a single Shuttle launch and provides greater safety than could be obtained with designs using pumped liquid metal cooling. Two configurations, one with the reactor and power conversion forward on the spacecraft with the ion engines aft and the other with reactor, power conversion and ion engines aft were selected as dual baseline designs based on minimum weight, minimum required technology development and maximum growth potential and flexibility.

Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Briggs, Maxwell H.; Sanzi, James L.; Brace, Michael H.

2013-01-01

Design concepts for small Fission Power Systems (FPS) have shown that heat pipe cooled reactors provide a passive, redundant, and lower mass option to transfer heat from the fuel to the power conversion system, as opposed to pumped loop designs typically associated with larger FPS. Although many systems have been conceptually designed and a few making it to electrically heated testing, none have been coupled to a real nuclear reactor. A demonstration test named DUFF Demonstration Using Flattop Fission, was planned by the Los Alamos National Lab (LANL) to use an existing criticality experiment named Flattop to provide the nuclear heat source. A team from the NASA Glenn Research Center designed, built, and tested a heat pipe and power conversion system to couple to Flattop with the end goal of making electrical power. This paper will focus on the design and testing performed in preparation for the DUFF test.

Dispersed bubble reactor for enhanced gas-liquid-solids contact and mass transfer

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

Vimalchand, Pannalal; Liu, Guohai; Peng, WanWang

An apparatus to promote gas-liquid contact and facilitate enhanced mass transfer. The dispersed bubble reactor (DBR) operates in the dispersed bubble flow regime to selectively absorb gas phase constituents into the liquid phase. The dispersion is achieved by shearing the large inlet gas bubbles into fine bubbles with circulating liquid and additional pumped liquid solvent when necessary. The DBR is capable of handling precipitates that may form during absorption or fine catalysts that may be necessary to promote liquid phase reactions. The DBR can be configured with multistage counter current flow sections by inserting concentric cylindrical sections into the risermore » to facilitate annular flow. While the DBR can absorb CO.sub.2 in liquid solvents that may lead to precipitates at high loadings, it is equally capable of handling many different types of chemical processes involving solids (precipitates/catalysts) along with gas and liquid phases.« less

Materials considerations in the design of a metal-hydride heat pump for an advanced extravehicular mobility unit

NASA Technical Reports Server (NTRS)

Liebert, B. E.

1986-01-01

A metal-hydride heat pump (HHP) has been proposed to provide an advanced regenerable nonventing thermal sink for the liquid-cooled garment worn during an extravehicular activity (EVA). The conceptual design indicates that there is a potential for significant advantages over the one presently being used by shuttle crew personnel as well as those that have been proposed for future use with the space station. Compared to other heat pump designs, a HHP offers the potential for extended use with no electrical power requirements during the EVA. In addition, a reliable, compact design is possible due to the absence of moving parts other than high-reliability check valves. Because there are many subtleties in the properties of metal hydrides for heat pump applications, it is essential that a prototype hydride heat pump be constructed with the selected materials before a committment is made for the final design. Particular care must be given to the evaporator heat exchanger worn by the astronaut since the performance of hydride heat pumps is generally heat transfer limited.

[Research on magnetic coupling centrifugal blood pump control based on a self-tuning fuzzy PI algorithm].

PubMed

Yang, Lei; Yang, Ming; Xu, Zihao; Zhuang, Xiaoqi; Wang, Wei; Zhang, Haibo; Han, Lu; Xu, Liang

2014-10-01

The purpose of this paper is to report the research and design of control system of magnetic coupling centrifugal blood pump in our laboratory, and to briefly describe the structure of the magnetic coupling centrifugal blood pump and principles of the body circulation model. The performance of blood pump is not only related to materials and structure, but also depends on the control algorithm. We studied the algorithm about motor current double-loop control for brushless DC motor. In order to make the algorithm adjust parameter change in different situations, we used the self-tuning fuzzy PI control algorithm and gave the details about how to design fuzzy rules. We mainly used Matlab Simulink to simulate the motor control system to test the performance of algorithm, and briefly introduced how to implement these algorithms in hardware system. Finally, by building the platform and conducting experiments, we proved that self-tuning fuzzy PI control algorithm could greatly improve both dynamic and static performance of blood pump and make the motor speed and the blood pump flow stable and adjustable.

[Coronary revascularization in patients with preoperative electrical storm].

PubMed

Kawashima, Toshiya; Naraoka, S

2007-03-01

We report 5 cases who underwent surgical coronary revascularization for subacute myocardial ischemia with preoperative electrical storm. All patients showed severe left ventricular dysfunction. Mean ejection fraction was 24.4 +/- 7.6%. Three patients had already had implantable cardioverter-defibrillator (ICD) therapy. Procedures were on-pump coronary artery bypass grafting (CABG) and mitral valvuloplasty (MVP) [case 1], on-pump CABG, MVP, left ventricular restoration (LVR) and cryoablation (case 2), and off-pump CABG (case 3-5). Case 5 necessitated conversion to on-pump for electrical storm during left circumflex artery (LCx) anastomosis. Case 3 suddenly died on the 2nd postoperative day due to electrical storm. Case 1 had recurrent attack of electrical storm postoperatively, treated by ICD, overdrive pacing, repeated intraaortic balloon pumping (IABP), deep sedation with endotracheal intubation, and finally catheter ablation. Four patients have survived 2 years (mean) postoperatively without any arrhythmia, and are all in good condition [New York Heart Association (NYHA) I] now. It was concluded that off-pump procedure was not suitable for subacute myocardial ischemia with electrical storm and that LVR with surgical cryoablation would be effective if indicated.

Posttest analysis of MIST Test 320201 using TRAC-PF1/MOD1

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

Siebe, D.A.; Steiner, J.L.; Boyack, B.E.

A posttest calculation and analysis of Multi-Loop Integral System Test 320201, a small-break loss-of-coolant accident (SBLOCA) test with a scaled 50-cm{sup 2} cold-leg pump discharge leak, has been completed and is reported herein. It was one in a series of tests, with leak size varied parametrically. Scaled leak sizes included 5, 10, (the nominal, Test 3109AA), and 50 cm{sub 2}. The test exhibited the major post-SBLOCA phenomena, as expected, including depressurization to saturation, interruption of loop flow, boiler-condenser mode cooling, refill, and postrefill cooldown. Full high-pressure injection and auxiliary feedwater were available, reactor coolant pumps were not available, and reactor-vesselmore » vent valves and guard heaters were automatically controlled. Constant level control in the steam-generator (SG) secondaries was used after SG-secondary refill; and symmetric SG pressure control was also used. The sequence of events seen in this test was similar to the sequence of events for much of the nominal test except that events occurred in a shorter time frame as the system inventory was reduced and the system depressurized at a faster rate. The calculation was performed using TRAC-PFL/MOD 1. Agreement between test data and the calculation was generally reasonable. All major trends and phenomena were correctly predicted. We believe that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications.« less

Posttest analysis of MIST Test 320201 using TRAC-PF1/MOD1

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

Siebe, D.A.; Steiner, J.L.; Boyack, B.E.

A posttest calculation and analysis of Multi-Loop Integral System Test 320201, a small-break loss-of-coolant accident (SBLOCA) test with a scaled 50-cm[sup 2] cold-leg pump discharge leak, has been completed and is reported herein. It was one in a series of tests, with leak size varied parametrically. Scaled leak sizes included 5, 10, (the nominal, Test 3109AA), and 50 cm[sub 2]. The test exhibited the major post-SBLOCA phenomena, as expected, including depressurization to saturation, interruption of loop flow, boiler-condenser mode cooling, refill, and postrefill cooldown. Full high-pressure injection and auxiliary feedwater were available, reactor coolant pumps were not available, and reactor-vesselmore » vent valves and guard heaters were automatically controlled. Constant level control in the steam-generator (SG) secondaries was used after SG-secondary refill; and symmetric SG pressure control was also used. The sequence of events seen in this test was similar to the sequence of events for much of the nominal test except that events occurred in a shorter time frame as the system inventory was reduced and the system depressurized at a faster rate. The calculation was performed using TRAC-PFL/MOD 1. Agreement between test data and the calculation was generally reasonable. All major trends and phenomena were correctly predicted. We believe that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications.« less

Taming The Next Set of Strategic Weapons Threats

DTIC Science & Technology

2006-06-01

Reactors Victor Gilinsky . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 6. Coping with Biological Threats after...Regime (MTCR) is not yet optimized to cope with these challenges. Finally, nuclear technologies have become much more difficult to control. New...resistance of the most popular type of power reactor concludes that the current international nuclear safeguards system needs to be modified to cope

High Efficiency Water Heating Technology Development Final Report, Part II: CO 2 and Absorption-Based Residential Heat Pump Water Heater Development

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

Gluesenkamp, Kyle R.; Abdelaziz, Omar; Patel, Viral K.

2017-05-01

The two objectives of this project were to 1.demonstrate an affordable path to an ENERGY STAR qualified electric heat pump water heater (HPWH) based on low-global warming potential (GWP) CO 2 refrigerant, and 2.demonstrate an affordable path to a gas-fired absorption-based heat pump water heater with a gas energy factor (EF) greater than 1.0. The first objective has been met, and the project has identified a promising low-cost option capable of meeting the second objective. This report documents the process followed and results obtained in addressing these objectives.

77 FR 5505 - Eagle Crest Energy Company; Notice of Availability of the Final Environmental Impact Statement...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-03

... Energy Projects has reviewed the application for license for the Eagle Mountain Pumped Storage... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Project No. 13123-002--California] Eagle Crest Energy Company; Notice of Availability of the Final Environmental Impact Statement for the...

Chemical vapor deposition of epitaxial silicon

DOEpatents

Berkman, Samuel

1984-01-01

A single chamber continuous chemical vapor deposition (CVD) reactor is described for depositing continuously on flat substrates, for example, epitaxial layers of semiconductor materials. The single chamber reactor is formed into three separate zones by baffles or tubes carrying chemical source material and a carrier gas in one gas stream and hydrogen gas in the other stream without interaction while the wafers are heated to deposition temperature. Diffusion of the two gas streams on heated wafers effects the epitaxial deposition in the intermediate zone and the wafers are cooled in the final zone by coolant gases. A CVD reactor for batch processing is also described embodying the deposition principles of the continuous reactor.

Process control, energy recovery and cost savings in acetic acid wastewater treatment.

PubMed

Vaiopoulou, E; Melidis, P; Aivasidis, A

2011-02-28

An anaerobic fixed bed loop (AFBL) reactor was applied for treatment of acetic acid (HAc) wastewater. Two pH process control concepts were investigated; auxostatic and chemostatic control. In the auxostatic pH control, feed pump is interrupted when pH falls below a certain pH value in the bioreactor, which results in reactor operation at maximum load. Chemostatic control assures alkaline conditions by setting a certain pH value in the influent, preventing initial reactor acidification. The AFBL reactor treated HAc wastewater at low hydraulic residence time (HRT) (10-12 h), performed at high space time loads (40-45 kg COD/m(3) d) and high space time yield (30-35 kg COD/m(3) d) to achieve high COD (Chemical Oxygen Demand) removal (80%). Material and cost savings were accomplished by utilizing the microbial potential for wastewater neutralization during anaerobic treatment along with application of favourable pH-auxostatic control. NaOH requirement for neutralization was reduced by 75% and HRT was increased up to 20 h. Energy was recovered by applying costless CO(2) contained in the biogas for neutralization of alkaline wastewater. Biogas was enriched in methane by 4 times. This actually brings in more energy profits, since biogas extra heating for CO(2) content during biogas combustion is minimized and usage of other acidifying agents is omitted. Copyright © 2010 Elsevier B.V. All rights reserved.

A simple device using magnetic transportation for droplet-based PCR.

PubMed

Ohashi, Tetsuo; Kuyama, Hiroki; Hanafusa, Nobuhiro; Togawa, Yoshiyuki

2007-10-01

The Polymerase chain reaction (PCR) was successfully and rapidly performed in a simple reaction device devoid of channels, pumps, valves, or other control elements used in conventional lab-on-a-chip technology. The basic concept of this device is the transportation of aqueous droplets containing hydrophilic magnetic beads in a flat-bottomed, tray-type reactor filled with silicone oil. The whole droplets sink to the bottom of the reactor because their specific gravity is greater than that of the silicone oil used here. The droplets follow the movement of a magnet located underneath the reactor. The notable advantage of the droplet-based PCR is the ability to switch rapidly the proposed reaction temperature by moving the droplets to the required temperature zones in the temperature gradient. The droplet-based reciprocative thermal cycling was performed by moving the droplets composed of PCR reaction mixture to the designated temperature zones on a linear temperature gradient from 50 degrees C to 94 degrees C generated on the flat bottom plate of the tray reactor. Using human-derived DNA containing the mitochondria genes as the amplification targets, the droplet-based PCR with magnetic reciprocative thermal cycling successfully provided the five PCR products ranging from 126 to 1,219 bp in 11 min with 30 cycles. More remarkably, the human genomic gene amplification targeting glyceraldehyde-3-phosphate dehydrogenase (GAPDH) gene was accomplished rapidly in 3.6 min with 40 cycles.

Treating domestic effluent wastewater treatment by aerobic biofilter with bioballs medium

NASA Astrophysics Data System (ADS)

Permatasari, R.; Rinanti, A.; Ratnaningsih, R.

2018-01-01

This laboratory scale research aimed to treat wastewater effluent with advanced treatment utilizing aerobic biofilter with bio-balls medium to obtain effluent quality in accordance with DKI Jakarta Governor Regulation No. 122 of 2005. The seeding and acclimatization were conducted in 4 weeks. The effluent were accommodated in a 150 L water barrel supported by a submersible pump. The effluent were treated in two boxes shaped reactors made of glasses with 36 L of each capacity. These reactors were equipped with aquarium aerators, sampling tap is 10 cm from the base of reactors, and bio-balls with 3 cm diameter are made of PVC. Reactors operated continuously with variations of retention time of 4 hours, 8 hours, 12 hours, 18 hours, and 24 hours and also variations of Carbon: Nitrogen: Phosphor = C: N: P ratio were, 100:5:1, 100:8:1, 100:10:1, 100:12:1, 100:15:1. The results showed that the optimum variance of retention time was 24 hours and the ratio of C:N:P was 100:10:1 yielded the largest removal efficiency for 83,33% of COD, 87,33% of BOD, 82,5% of Ammonia, 79,1% of Nitrate, 92% of Nitrite, 84,82% of Oil and Grease. The concentration parameter resulted from outlet biofilter has met the domestic wastewater quality standard of DKI Jakarta.

IN-PILE LOOP IRRADIATION OF AQUEOUS THORIA-URANIA SLURRY AT ELEVATED TEMPERATURE. DESIGN AND IN-PILE OPERATION OF LOOP L-2-27S

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

Savage, H.C.; Compere, E.L.; Baker, J.M.

1962-02-14

An in-pile pump loop, designed to fit within horizontal beam hole HB-2 of the Low-Intensity Test Reactor (LlTR), was used to circulate an aqueous thoria- urania slurry while exposed to reactor irradiation. The total loop volume was about 1600 ml, including pump and pressurizer, but the slurry was confined to the 900-ml volume of the main loop stream by means of a sintered stainless steel filter. The filter was an important feature of the loop design in that it provided a thoria-free filtrate as a purge stream to the pressurizer and pump bearings to prevent entry and accumulation of thoriamore » in these two regions. Corrosion-test specimens of Zircaloy-2, titanium, and type 347 stainless steel were placed in the loop at three different locations for exposure to three different levels of irradiation. Duplicate sets of specimens in each position were exposed to flow velocities of 8 and 22 fps, respectively. For the in-pile irradiation, thorium oxide containing 0.43 wt of enriched U, based on Th, was used. This thoria-urania was produced by air calcination at l225 deg C of coprecipit.ited oxalates and had a me.in particle size of l.7 mu . A Pd catalyst w-as dispersed in the slurry for liquid-phase recombination of the radiolytic gas. The loop was operated in beam-holc HB-2 of the LlTR from July 19 to October l9, l960. During this period slurry was continuously eireulated at 280 deg C for 2220 hr without incident; 1839 hr were at full reactor power (3 Mwt), at which the estimated average thermal flux over the 300volume core section was 5 x l0/sup 12/ neutrons/cm/sup 2/- see. At the start of in-pile operation the loop was charged to a concei tration of 979 g of Th and 3.83 g of fully enriched U per li (at 280 deg C) which was reduced by sampling to 748 g of Th idnd 2.74 g of U per liter at the end of the irradiiition period bascd on the assunmption that no losses had occurrcd. Siinmples of slurry were withdrawn at intervals for analyses to determine the effects of radiation on the thoria-urania slurry. (auth)« less

Hydraulic development of high specific-speed pump-turbines by means of an inverse design method, numerical flow-simulation (CFD) and model testing

NASA Astrophysics Data System (ADS)

Kerschberger, P.; Gehrer, A.

2010-08-01

In recent years an increased interest in pump-turbines has been recognized in the market. The rapid availability of pumped storage schemes and the benefits to the power system by peak lopping, providing reserve and rapid response for frequency control are becoming of growing advantage. In that context it is requested to develop pump-turbines that reliably stand dynamic operation modes, fast changes of the discharge rate by adjusting the variable diffuser vanes as well as fast changes from pump to turbine operation. Within the present study various flow patterns linked to the operation of a pump-turbine system are discussed. In that context pump and turbine mode are presented separately and different load cases at both operation modes are shown. In order to achieve modern, competitive pump-turbine designs it is further explained which design challenges should be considered during the geometry definition of a pump-turbine impeller. Within the present study a runner-blade profile for a low head pump-turbine has been developed. For the initial hydraulic runner-blade design, an inverse design method has been applied. Within this design procedure, a first blade geometry is generated by imposing the pressure loading-distribution and by means of an inverse 3D potential-flow-solution. The hydraulic behavior of both, pump-mode and turbine-mode is then evaluated by solving the full 3D Navier-Stokes equations in combination with a robust turbulence model. Based on this initial design the blade profile has been further optimized and redesigned considering various hydraulic pump-turbine requirements. Finally, the progress in hydraulic design is demonstrated by model test results which show a significant improvement in hydraulic performance compared to an existing reference design.

Simulation analysis and experimental verification of spiral-tube-type valveless piezoelectric pump with gyroscopic effect

NASA Astrophysics Data System (ADS)

Leng, Xuefei; Zhang, Jianhui; Jiang, Yan; Wang, Shouyin; Zhao, Chunsheng

2014-07-01

The current research of the valveless piezoelectric pump focuses on increasing the flow rate and pressure differential. Compared with the valve piezoelectric pump, the valveless one has excellent performances in simple structure, low cost, and easy miniaturization. So, their important development trend is the mitigation of their weakness, and the multi-function integration. The flow in a spiral tube element is sensitive to the element attitude caused by the Coriolis force, and that a valveless piezoelectric pump is designed by applying this phenomenon. The pump has gyroscopic effect, and has both the actuator function of fluid transfer and the sensor function, which can obtain the angular velocity when its attitude changes. First, the present paper analyzes the flow characteristics in the tube, obtains the calculation formula for the pump flow, and identifies the relationship between pump attitude and flow, which clarifies the impact of flow and driving voltage, frequency, spiral line type and element attitude, and verifies the gyroscopic effect of the pump. Then, the finite element simulation is used to verify the theory. Finally, a pump is fabricated for experimental testing of the relationship between pump attitude and pressure differential. Experimental results show that when Archimedes spiral θ=4π is selected for the tube design, and the rotation speed of the plate is 70 r/min, the pressure differential is 88.2 Pa, which is 1.5 times that of 0 r/min rotation speed. The spiral-tube-type valveless piezoelectric pump proposed can turn the element attitude into a form of pressure output, which is important for the multi-function integration of the valveless piezoelectric pump and for the development of civil gyroscope in the future.

Research concerning the evaluation of the connection forces in the joints of the sucker rod pumping units mechanism

NASA Astrophysics Data System (ADS)

Bădoiu, D.; Petrescu, M. G.; Antonescu, N. N.; Toma, G.

2018-01-01

At present, the sucker rod pumping installations are the most used in the case of the wells in production, when an eruptive exploitation is not possible. The practice has demonstrated that an important role in increasing safety in the operation of the pumping units has the design of the various component bearings because of the extremely high values of the connection forces to which they are loaded. That is why it is necessary to establish as accurately as possible the values of these connecting forces. In the paper is analyzed the dynamics of a conventional pumping unit mechanism. The dynamic study which allows establishing the values of the connecting forces in the joints is performed within the Assur structural groups. The dynamic analysis was implemented into a computer program using Maple programming environment and finally it has been presented some simulation results in the case of a C-320D-256-100 pumping unit.

Research on solar pumped liquid lasers. Final Report

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

Cox, J.D.; Kurzweg, U.H.; Weinstein, N.H.

1985-04-01

A solar pumped liquid laser that can be scaled up to high power (10 mW CW) for space applications was developed. Liquid lasers have the advantage over gases in that they provide much higher lasant densities and thus high-power densities. Liquids also have advantages over solids in that they have much higher damage thresholds and are much cheaper to produce for large scale applications. Among the liquid laser media that are potential candidates for solar pumping, the POC13: Nd sup 3+:ZrCl4 liquid was chosen for its high intrinsic efficiency and its relatively good stability against decomposition due to protic contamination.more » The development of a manufacturing procedure and performance testing of the laser liquid and the development of an inexpensive large solar concentrator to pump the laser are examined.« less

Research on Mechanism and Model of Centralized Bidding for Pumped Storage Power in Shanghai

NASA Astrophysics Data System (ADS)

Hua, Zhong; Ying, Zhiwei; Lv, Zhengyu; Jianlin, Yang; Huang, Yupeng; Li, Dong

2017-05-01

China is now in the transition stage toward power market and in some specific area, market approach has already been adopted to improve the overall efficiency. In this paper, Bidding and trading modes of pumped storage energy in various regions of China are analysed. Based on the constraints of bidding price and electricity, as well as the system power flow, the trading model is established to collect the capacity cost of pumped storage energy in Shanghai. With the trading model proposed, that the generators who actively undertake the capacity cost of pumped storage energy and bid enough electricity with lower price can be rewarded, while those attempts to conspire and manipulate the market will be penalized. Finally, using seven generators in Shanghai as examples to simulate the market operation, the effectiveness of the proposed model is verified.

Entrainment sampling at the Savannah River Site (SRS) Savannah River water intakes (1991)

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

Paller, M.

1990-11-01

Cooling water for the Westinghouse Savannah River Company (WSRC) L-Reactor, K-Reactor, and makeup water for Par Pond is pumped from the Savannah River at the 1G, 3G, and 5G pumphouses. Ichthyoplankton (drifting fish larvae and eggs) from the river are entrained into the reactor cooling systems with the river water. They are passed through the reactor heat exchangers where temperatures may reach 70{degree}C during full power operation. Ichthyoplankton mortality under such conditions is presumably 100%. Apart from a small pilot study conducted in 1989, ichthyoplankton samples have not been collected from the vicinity of the SRS intake canals since 1985.more » The Department of Energy (DOE) has requested that the Environmental Sciences Section (ESS) of the Savannah River Laboratory (SRL) resume ichthyoplankton sampling for the purpose of assessing entrainment at the SRS Savannah River intakes. This request is due to the anticipated restart of several SRS reactors and the growing concern surrounding striped bass and American shad stocks in the Savannah River. The following scope of work presents a sampling plan that will collect information on the spatial and temporal distribution of fish eggs and larvae near the SRS intake canal mouths. This data will be combined with information on water movement patterns near the canal mouths in order to determine the percentage of ichthyoplankton that are removed from the Savannah River by the SRS intakes. The following sampling plan incorporates improvements in experimental design that resulted from the findings of the 1989 pilot study. 1 fig.« less

Final Environmental Impact Statement for Oil Refinery, Georgetown, South Carolina. Volume 2.

DTIC Science & Technology

1984-09-01

at gauging and pumping stations. _adio and telephone communication exists between gaugers and pumping stations or vessels. 2. Describe secondary...thickness gauging . 4. Internal heatingO coil leakage is controlled by one or more of the following control factors: (a) M onitoring the steam return...behavior of sediment and oil emulsions. Journal of Sedimentary Petrology 47(2):671-677. Beam, H.W., South Carolina Coastal Council. [Letter to J.E. Jenkins

Performance and Costs of Ductless Heat Pumps in Marine-Climate High-Performance Homes -- Habitat for Humanity The Woods

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

Michael Lubliner; Howard, Luke; Hales, David

2016-02-23

This final Building America Partnership report focuses on the results of field testing, modeling, and monitoring of ductless mini-split heat pump hybrid heating systems in seven homes built and first occupied at various times between September 2013 and October 2014. The report also provides WSU documentation of high-performance home observations, lessons learned, and stakeholder recommendations for builders of affordable high-performance housing.

Pressurized-water reactor internals aging degradation study. Phase 1

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

Luk, K.H.

1993-09-01

This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pinsmore » and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.« less

On The Stability Of Model Flows For Chemical Vapour Deposition

NASA Astrophysics Data System (ADS)

Miller, Robert

2016-11-01

The flow in a chemical vapour deposition (CVD) reactor is assessed. The reactor is modelled as a flow over an infinite-radius rotating disk, where the mean flow and convective instability of the disk boundary layer are measured. Temperature-dependent viscosity and enforced axial flow are used to model the steep temperature gradients present in CVD reactors and the pumping of the gas towards the disk, respectively. Increasing the temperature-dependence parameter of the fluid viscosity (ɛ) results in an overall narrowing of the fluid boundary layer. Increasing the axial flow strength parameter (Ts) accelerates the fluid both radially and axially, while also narrowing the thermal boundary layer. It is seen that when both effects are imposed, the effects of axial flow generally dominate those of the viscosity temperature dependence. A local stability analysis is performed and the linearized stability equations are solved using a Galerkin projection in terms of Chebyshev polynomials. The neutral stability curves are then plotted for a range of ɛ and Ts values. Preliminary results suggest that increasing Ts has a stabilising effect on both type I and type II stationary instabilities, while small increases in ɛ results in a significant reduction to the critical Reynolds number.

Modelling biological Cr(VI) reduction in aquifer microcosm column systems.

PubMed

Molokwane, Pulane E; Chirwa, Evans M N

2013-01-01

Several chrome processing facilities in South Africa release hexavalent chromium (Cr(VI)) into groundwater resources. Pump-and-treat remediation processes have been implemented at some of the sites but have not been successful in reducing contamination levels. The current study is aimed at developing an environmentally friendly, cost-effective and self-sustained biological method to curb the spread of chromium at the contaminated sites. An indigenous Cr(VI)-reducing mixed culture of bacteria was demonstrated to reduce high levels of Cr(VI) in laboratory samples. The effect of Cr(VI) on the removal rate was evaluated at concentrations up to 400 mg/L. Following the detailed evaluation of fundamental processes for biological Cr(VI) reduction, a predictive model for Cr(VI) breakthrough through aquifer microcosm reactors was developed. The reaction rate in batch followed non-competitive rate kinetics with a Cr(VI) inhibition threshold concentration of approximately 99 mg/L. This study evaluates the application of the kinetic parameters determined in the batch reactors to the continuous flow process. The model developed from advection-reaction rate kinetics in a porous media fitted best the effluent Cr(VI) concentration. The model was also used to elucidate the logistic nature of biomass growth in the reactor systems.

Changing concepts of geologic structure and the problem of siting nuclear reactors: Examples from Washington State

NASA Astrophysics Data System (ADS)

Tabor, R. W.

1986-09-01

The conflict between regulation and healthy evolution of geological science has contributed to the difficulties of siting nuclear reactors. On the Columbia Plateau in Washington, but for conservative design of the Hanford reactor facility, the recognition of the little-understood Olympic-Wallowa lineament as a major, possibly still active structural alinement might have jeopardized the acceptability of the site for nuclear reactors. On the Olympic Peninsula, evolving concepts of compressive structures and their possible recent activity and the current recognition of a subducting Juan de Fuca plate and its potential for generating great earthquakes—both concepts little-considered during initial site selection—may delay final acceptance of the Satsop site. Conflicts of this sort are inevitable but can be accommodated if they are anticipated in the reactor-licensing process. More important, society should be increasing its store of geologic knowledge now, during the current recess in nuclear reactor siting.

Development of a Model and Computer Code to Describe Solar Grade Silicon Production Processes

NASA Technical Reports Server (NTRS)

Srivastava, R.; Gould, R. K.

1979-01-01

Mathematical models and computer codes based on these models, which allow prediction of the product distribution in chemical reactors for converting gaseous silicon compounds to condensed-phase silicon were developed. The following tasks were accomplished: (1) formulation of a model for silicon vapor separation/collection from the developing turbulent flow stream within reactors of the Westinghouse (2) modification of an available general parabolic code to achieve solutions to the governing partial differential equations (boundary layer type) which describe migration of the vapor to the reactor walls, (3) a parametric study using the boundary layer code to optimize the performance characteristics of the Westinghouse reactor, (4) calculations relating to the collection efficiency of the new AeroChem reactor, and (5) final testing of the modified LAPP code for use as a method of predicting Si(1) droplet sizes in these reactors.

Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

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

Feltus, M.A.

1989-11-01

The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

Development of a reactor with carbon catalysts for modular-scale, low-cost electrochemical generation of H 2O 2

DOE PAGES

Chen, Zhihua; Chen, Shucheng; Siahrostami, Samira; ...

2017-03-01

The development of small-scale, decentralized reactors for H 2O 2 production that can couple to renewable energy sources would be of great benefit, particularly for water purification in the developing world. Herein, we describe our efforts to develop electrochemical reactors for H 2O 2 generation with high Faradaic efficiencies of >90%, requiring cell voltages of only ~1.6 V. The reactor employs a carbon-based catalyst that demonstrates excellent performance for H 2O 2 production under alkaline conditions, as demonstrated by fundamental studies involving rotating-ring disk electrode methods. Finally, the low-cost, membrane-free reactor design represents a step towards a continuous, modular-scale, de-centralizedmore » production of H 2O 2.« less

Long lifetime fast spectrum reactor for lunar surface power system

NASA Astrophysics Data System (ADS)

Kambe, Mitsuru

1993-01-01

In the framework of innovative reactor research activities, a conceptual design study of fast spectrum reactor and primary system for 800 kWe lunar surface power system to be combined with potassium Rankine cycle power conversion has been conducted to meet the power requirements of the lunar base activities in the next century. The reactor subsystem is characterized by RAPID (Refueling by All Pins Integrated Design) concept to enhance inherent safety and to enable quick and simplifed refueling in every 10 years. RAPID concept affords power plant design lifetime of up to 30 years. Integrity of the reactor structure and replacement of failed primary circuits are also discussed. Substantial reduction in per-kWh cost on considering launch, emplacement, and final disposition can be expected by a long system lifetime.

Microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1984-01-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety systemmore » is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.« less

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

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

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« less

78 FR 3470 - DTE Electric Company (Formerly the Detroit Edison Company), Notice of Availability of Final...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-16

...--specific Web page at http://www.nrc.gov/reactors/new-reactors/col/fermi.html . The Ellis Library and... possesses and are publicly-available, using any of the following methods: Federal Rulemaking Web site: Go to... Documents Access and Management System (ADAMS): You may access publicly-available documents online in the...

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2012 CFR

2012-01-01

... complete and acceptable for docketing under § 2.101(a)(3), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, shall determine whether the... a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION RULES OF...

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2011 CFR

2011-01-01

... complete and acceptable for docketing under § 2.101(a)(3), the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, shall determine whether the... a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION RULES OF...

Completion summary for borehole USGS 136 near the Advanced Test Reactor Complex, Idaho National Laboratory, Idaho

USGS Publications Warehouse

Twining, Brian V.; Bartholomay, Roy C.; Hodges, Mary K.V.

2012-01-01

In 2011, the U.S. Geological Survey, in cooperation with the U.S. Department of Energy, cored and completed borehole USGS 136 for stratigraphic framework analyses and long-term groundwater monitoring of the eastern Snake River Plain aquifer at the Idaho National Laboratory. The borehole was initially cored to a depth of 1,048 feet (ft) below land surface (BLS) to collect core, open-borehole water samples, and geophysical data. After these data were collected, borehole USGS 136 was cemented and backfilled between 560 and 1,048 ft BLS. The final construction of borehole USGS 136 required that the borehole be reamed to allow for installation of 6-inch (in.) diameter carbon-steel casing and 5-in. diameter stainless-steel screen; the screened monitoring interval was completed between 500 and 551 ft BLS. A dedicated pump and water-level access line were placed to allow for aquifer testing, for collecting periodic water samples, and for measuring water levels.Geophysical and borehole video logs were collected after coring and after the completion of the monitor well. Geophysical logs were examined in conjunction with the borehole core to describe borehole lithology and to identify primary flow paths for groundwater, which occur in intervals of fractured and vesicular basalt.A single-well aquifer test was used to define hydraulic characteristics for borehole USGS 136 in the eastern Snake River Plain aquifer. Specific-capacity, transmissivity, and hydraulic conductivity from the aquifer test were at least 975 gallons per minute per foot, 1.4 × 105 feet squared per day (ft2/d), and 254 feet per day, respectively. The amount of measureable drawdown during the aquifer test was about 0.02 ft. The transmissivity for borehole USGS 136 was in the range of values determined from previous aquifer tests conducted in other wells near the Advanced Test Reactor Complex: 9.5 × 103 to 1.9 × 105 ft2/d.Water samples were analyzed for cations, anions, metals, nutrients, total organic carbon, volatile organic compounds, stable isotopes, and radionuclides. Water samples from borehole USGS 136 indicated that concentrations of tritium, sulfate, and chromium were affected by wastewater disposal practices at the Advanced Test Reactor Complex. Depth-discrete groundwater samples were collected in the open borehole USGS 136 near 965, 710, and 573 ft BLS using a thief sampler; on the basis of selected constituents, deeper groundwater samples showed no influence from wastewater disposal at the Advanced Test Reactor Complex.

Temperature Dependences for the Reactions of O2- and O- with N and O Atoms in a Selected-Ion Flow Tube Instrument

DTIC Science & Technology

2013-10-07

quadrupole mass filter, mass selected, and injected into the flow reactor via a Venturi - type inlet. Ions undergo ∼105 collisions with helium buffer... gas at pressures of 0.4 to 0.8 Torr resulting in complete or near-complete thermalization.10 The higher pressure was used when studying the high...butterfly gate valve resulting in lower pumping speeds and thus longer reaction times. Neutrals were injected 49 cm before the end of the flow tube and

Integrated Systems Performance Assessment for the Evaluation of Space Nuclear Reactor Design Concepts (Phase 1: Demonstration of the Methodology).

DTIC Science & Technology

1992-11-01

Incorporated. Each design is characterized by a moderated core, a NaK pumped loop primary coolant system, and a potassium heat pipe radiator as the...1 1 10 1 RelHX 1 2 10 2 nRel HX 3 3 RelSS nRelSS Irr 4 3 7 8 9 io 2 + 2 + 2 + 2 nRel Pwr nRel NaK nRel RC nRel HX 1 1 11 1 RelSS 1 2 11 2 nRel SS 3 3