Small reactor power system for space application

NASA Technical Reports Server (NTRS)

Shirbacheh, M.

1987-01-01

A development history and comparative performance capability evaluation is presented for spacecraft nuclear powerplant Small Reactor Power System alternatives. The choice of power conversion technology depends on the reactor's operating temperature; thermionic, thermoelectric, organic Rankine, and Alkali metal thermoelectric conversion are the primary power conversion subsystem technology alternatives. A tabulation is presented for such spacecraft nuclear reactor test histories as those of SNAP-10A, SP-100, and NERVA.

Current Development of Nuclear Thermal Propulsion technologies at the Center for Space Nuclear Research

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

Robert C. O'Brien; Steven K. Cook; Nathan D. Jerred

Nuclear power and propulsion has been considered for space applications since the 1950s. Between 1955 and 1972 the US built and tested over twenty nuclear reactors / rocket engines in the Rover/NERVA programs1. The Aerojet Corporation was the prime contractor for the NERVA program. Modern changes in environmental laws present challenges for the redevelopment of the nuclear rocket. Recent advances in fuel fabrication and testing options indicate that a nuclear rocket with a fuel composition that is significantly different from those of the NERVA project can be engineered; this may be needed to ensure public support and compliance with safetymore » requirements. The Center for Space Nuclear Research (CSNR) is pursuing a number of technologies, modeling and testing processes to further the development of safe, practical and affordable nuclear thermal propulsion systems.« less

L-Area STS MTR/NRU/NRX Grapple Assembly Closure Mechanics Review

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

Huizenga, D. J.

2016-06-08

A review of the closure mechanics associated with the Shielded Transfer System (STS) MTR/NRU/NRX grapple assembly utilized at the Savannah River Site (SRS) was performed. This review was prompted by an operational event which occurred at the Canadian Nuclear Laboratories (CNL) utilizing a DTS-XL grapple assembly which is essentially identical to the STS MTR/NRU/NRX grapple assembly used at the SRS. The CNL operational event occurred when a NRU/NRX fuel basket containing spent nuclear fuel assemblies was inadvertently released by the DTS-XL grapple assembly during a transfer. The SM review of the STS MTR/NRU/NRX grapple assembly will examine the operational aspectsmore » of the STS and the engineered features of the STS which prevent such an event at the SRS. The design requirements for the STS NRU/NRX modifications and the overall layout of the STS are provided in other documents.« less

Nuclear Propulsion for Space, Understanding the Atom Series.

ERIC Educational Resources Information Center

Corliss, William R.; Schwenk, Francis C.

The operation of nuclear rockets with respect both to rocket theory and to various fuels is described. The development of nuclear reactors for use in nuclear rocket systems is provided, with the Kiwi and NERVA programs highlighted. The theory of fuel element and reactor construction and operation is explained with particular reference to rocket…

Nuclear power for space based systems

NASA Astrophysics Data System (ADS)

Livingston, J. M.; Ivanenok, Joseph F., III

1991-09-01

A 100 kWe closed Brayton cycle power conversion system utilizing a recuperator coupled to a NERVA derivative reactor for a lunar power plant is presented. Power plant mass versus recuperator effectiveness, compressor inlet temperature, and turbine pressure ratio are described.

NERVA-Derived Nuclear Thermal Propulsion Dual Mode Operation

NASA Astrophysics Data System (ADS)

Zweig, Herbert R.; Hundal, Rolv

1994-07-01

Generation of electrical power using the nuclear heat source of a NERVA-derived nuclear thermal rocket engine is presented. A 111,200 N thrust engine defined in a study for NASA-LeRC in FY92 is the reference engine for a three-engine vehicle for which a 50 kWe capacity is required. Processes are described for energy extraction from the reactor and for converting the energy to electricity. The tie tubes which support the reactor fuel elements are the source of thermal energy. The study focuses on process systems using Stirling cycle energy conversion operating at 980 K and an alternate potassium-Rankine system operating at 1,140 K. Considerations are given of the effect of the power production on turbopump operation, ZrH moderator dissociation, creep strain in the tie tubes, hydrogen permeation through the containment materials, requirements for a backup battery system, and the effects of potential design changes on reactor size and criticality. Nuclear considerations include changing tie tube materials to TZM, changing the moderator to low vapor-pressure yttrium hydride, and changing the fuel form from graphite matrix to a carbon-carbide composite.

Raising Nuclear Thermal Propulsion (NTP) Technology Readiness Above 3

NASA Technical Reports Server (NTRS)

Gerrish, Harold P., Jr.

2014-01-01

NTP development is currently supported by the NASA program office "Advanced Exploration Systems". The concept is a main propulsion option being considered for human missions to Mars in the 2030's. Major NTP development took place in the 1960's and 1970's under the Rover/NERVA program. The technology had matured to TRL 6 and was preparing to go to TRL 7 with a prototype flight engine before the program was cancelled. Over the last 40 years, a variety of continuations started, but only lasted a few years each. The Rover/NERVA infrastructure is almost all gone. The only remains are a few pieces of hardware, final reports and a few who worked the Rover/NERVA. Two types of nuclear fuel are being investigated to meet the current engine design specific impulse of 900 seconds compared to approximately 850 seconds demonstrated during Rover/NERVA. One is a continuation of composite fuel with new coatings to better control mid-band corrosion. The other type is a CERMET fuel made of Tungsten and UO2. Both fuels are being made from Rover/NERVA lessons learned, but with slightly different recipes to increase fuel endurance at higher operating temperatures. The technology readiness level (TRL) of these current modified reactor fuels is approximately TRL 3. To keep the development cost low and help mature the TRL level past 4 quickly, a few special non-nuclear test facilities have been made to test surrogate fuel, with depleted uranium, as coupons and full length elements. Both facilities utilize inductive heating and are licensed to handle depleted uranium. TRL 5 requires exposing the fuel to a nuclear environment and TRL 6 requires a prototype ground or flight engine system test. Currently, three different NTP ground test facility options are being investigated: exhaust scrubber, bore hole, and total exhaust containment. In parallel, a prototype flight demonstration test is also being studied. The first human mission to Mars in the 2030's is currently 2033. For an advanced propulsion concept to be seriously considered for use, the engine development plans need to show it is feasible and affordable to reach TRL 8 by 2027 and can be qualified for human mission use.

Rover/NERVA-derived near-term nuclear propulsion

NASA Technical Reports Server (NTRS)

1993-01-01

FY-92 accomplishments centered on conceptual design and analyses for 25, 50, and 75 K engines with emphasis on the 50 K engine. During the first period of performance, flow and energy balances were prepared for each of these configurations and thrust-to-weight values were estimated. A review of fuel technology and key data from the Rover/NERVA program established a baseline for proven reactor performance and areas of enhancement to meet near-term goals. Studies were performed of the criticality and temperature profiles for probable fuel and moderator loadings for the three engine sizes, with a more detailed analysis of the 50 K size. During the second period of performance, analyses of the 50 K engine continued. A chamber/nozzle contour was selected and heat transfer and fatigue analyses were performed for likely construction materials. Reactor analyses were performed to determine component radiation heating rates, reactor radiation fields, water immersion poisoning requirements, temperature limits for restartability, and a tie-tube thermal analysis. Finally, a brief assessment of key enabling technologies was made, with a view toward identifying development issues and identification of the critical path toward achieving engine qualification within 10 years.

The lateral mobility of cell adhesion molecules is highly restricted at septate junctions in Drosophila.

PubMed

Laval, Monique; Bel, Christophe; Faivre-Sarrailh, Catherine

2008-07-18

A complex of three cell adhesion molecules (CAMs) Neurexin IV(Nrx IV), Contactin (Cont) and Neuroglian (Nrg) is implicated in the formation of septate junctions between epithelial cells in Drosophila. These CAMs are interdependent for their localization at septate junctions and e.g. null mutation of nrx IV or cont induces the mislocalization of Nrg to the baso-lateral membrane. These mutations also result in ultrastructural alteration of the strands of septate junctions and breakdown of the paracellular barrier. Varicose (Vari) and Coracle (Cora), that both interact with the cytoplasmic tail of Nrx IV, are scaffolding molecules required for the formation of septate junctions. We conducted photobleaching experiments on whole living Drosophila embryos to analyze the membrane mobility of CAMs at septate junctions between epithelial cells. We show that GFP-tagged Nrg and Nrx IV molecules exhibit very stable association with septate junctions in wild-type embryos. Nrg-GFP is mislocalized to the baso-lateral membrane in nrx IV or cont null mutant embryos, and displays increased mobile fraction. Similarly, Nrx IV-GFP becomes distributed to the baso-lateral membrane in null mutants of vari and cora, and its mobile fraction is strongly increased. The loss of Vari, a MAGUK protein that interacts with the cytoplasmic tail of Nrx IV, has a stronger effect than the null mutation of nrx IV on the lateral mobility of Nrg-GFP. The strands of septate junctions display a stable behavior in vivo that may be correlated with their role of paracellular barrier. The membrane mobility of CAMs is strongly limited when they take part to the multimolecular complex forming septate junctions. This restricted lateral diffusion of CAMs depends on both adhesive interactions and clustering by scaffolding molecules. The lateral mobility of CAMs is strongly increased in embryos presenting alteration of septate junctions. The stronger effect of vari by comparison with nrx IV null mutation supports the hypothesis that this scaffolding molecule may cross-link different types of CAMs and play a crucial role in stabilizing the strands of septate junctions.

Engine management during NTRE start up

NASA Technical Reports Server (NTRS)

Bulman, Mel; Saltzman, Dave

1993-01-01

The topics are presented in viewgraph form and include the following: total engine system management critical to successful nuclear thermal rocket engine (NTRE) start up; NERVA type engine start windows; reactor power control; heterogeneous reactor cooling; propellant feed system dynamics; integrated NTRE start sequence; moderator cooling loop and efficient NTRE starting; analytical simulation and low risk engine development; accurate simulation through dynamic coupling of physical processes; and integrated NTRE and mission performance.

Scoping Calculations of Power Sources for Nuclear Electric Propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1994-01-01

This technical memorandum describes models and calculational procedures to fully characterize the nuclear island of power sources for nuclear electric propulsion. Two computer codes were written: one for the gas-cooled NERVA derivative reactor and the other for liquid metal-cooled fuel pin reactors. These codes are going to be interfaced by NASA with the balance of plant in order to make scoping calculations for mission analysis.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas M.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas L.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic- metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Review of Nuclear Thermal Propulsion Ground Test Options

NASA Technical Reports Server (NTRS)

Coote, David J.; Power, Kevin P.; Gerrish, Harold P.; Doughty, Glen

2015-01-01

High efficiency rocket propulsion systems are essential for humanity to venture beyond the moon. Nuclear Thermal Propulsion (NTP) is a promising alternative to conventional chemical rockets with relatively high thrust and twice the efficiency of highest performing chemical propellant engines. NTP utilizes the coolant of a nuclear reactor to produce propulsive thrust. An NTP engine produces thrust by flowing hydrogen through a nuclear reactor to cool the reactor, heating the hydrogen and expelling it through a rocket nozzle. The hot gaseous hydrogen is nominally expected to be free of radioactive byproducts from the nuclear reactor; however, it has the potential to be contaminated due to off-nominal engine reactor performance. NTP ground testing is more difficult than chemical engine testing since current environmental regulations do not allow/permit open air testing of NTP as was done in the 1960's and 1970's for the Rover/NERVA program. A new and innovative approach to rocket engine ground test is required to mitigate the unique health and safety risks associated with the potential entrainment of radioactive waste from the NTP engine reactor core into the engine exhaust. Several studies have been conducted since the ROVER/NERVA program in the 1970's investigating NTP engine ground test options to understand the technical feasibility, identify technical challenges and associated risks and provide rough order of magnitude cost estimates for facility development and test operations. The options can be divided into two distinct schemes; (1) real-time filtering of the engine exhaust and its release to the environment or (2) capture and storage of engine exhaust for subsequent processing.

The Nuclear Cryogenic Propulsion Stage

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Kim, Tony; Emrich, William J.; Hickman, Robert R.; Broadway, Jeramie W.; Gerrish, Harold P.; Belvin, Anthony D.; Borowski, Stanley K.; Scott, John H.

2014-01-01

Nuclear Thermal Propulsion (NTP) development efforts in the United States have demonstrated the technical viability and performance potential of NTP systems. For example, Project Rover (1955 - 1973) completed 22 high power rocket reactor tests. Peak performances included operating at an average hydrogen exhaust temperature of 2550 K and a peak fuel power density of 5200 MW/m3 (Pewee test), operating at a thrust of 930 kN (Phoebus-2A test), and operating for 62.7 minutes in a single burn (NRX-A6 test). Results from Project Rover indicated that an NTP system with a high thrust-to-weight ratio and a specific impulse greater than 900 s would be feasible. Excellent results were also obtained by the former Soviet Union. Although historical programs had promising results, many factors would affect the development of a 21st century nuclear thermal rocket (NTR). Test facilities built in the US during Project Rover no longer exist. However, advances in analytical techniques, the ability to utilize or adapt existing facilities and infrastructure, and the ability to develop a limited number of new test facilities may enable affordable development, qualification, and utilization of a Nuclear Cryogenic Propulsion Stage (NCPS). Bead-loaded graphite fuel was utilized throughout the Rover/NERVA program, and coated graphite composite fuel (tested in the Nuclear Furnace) and cermet fuel both show potential for even higher performance than that demonstrated in the Rover/NERVA engine tests.. NASA's NCPS project was initiated in October, 2011, with the goal of assessing the affordability and viability of an NCPS. FY 2014 activities are focused on fabrication and test (non-nuclear) of both coated graphite composite fuel elements and cermet fuel elements. Additional activities include developing a pre-conceptual design of the NCPS stage and evaluating affordable strategies for NCPS development, qualification, and utilization. NCPS stage designs are focused on supporting human Mars missions. The NCPS is being designed to readily integrate with the Space Launch System (SLS). A wide range of strategies for enabling affordable NCPS development, qualification, and utilization should be considered. These include multiple test and demonstration strategies (both ground and in-space), multiple potential test sites, and multiple engine designs. Two potential NCPS fuels are currently under consideration - coated graphite composite fuel and tungsten cermet fuel. During 2014 a representative, partial length (approximately 16") coated graphite composite fuel element with prototypic depleted uranium loading is being fabricated at Oak Ridge National Laboratory (ORNL). In addition, a representative, partial length (approximately 16") cermet fuel element with prototypic depleted uranium loading is being fabricated at Marshall Space Flight Center (MSFC). During the development process small samples (approximately 3" length) will be tested in the Compact Fuel Element Environmental Tester (CFEET) at high temperature (approximately 2800 K) in a hydrogen environment to help ensure that basic fuel design and manufacturing process are adequate and have been performed correctly. Once designs and processes have been developed, longer fuel element segments will be fabricated and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREE) at high temperature (approximately 2800 K) and in flowing hydrogen.

Satellite nuclear power station: An engineering analysis

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Kirby, K. D.; Yang, Y. Y.

1973-01-01

A nuclear-MHD power plant system which uses a compact non-breeder reactor to produce power in the multimegawatt range is analyzed. It is shown that, operated in synchronous orbit, the plant would transmit power safely to the ground by a microwave beam. Fuel reprocessing would take place in space, and no radioactive material would be returned to earth. Even the effect of a disastrous accident would have negligible effect on earth. A hydrogen moderated gas core reactor, or a colloid-core, or NERVA type reactor could also be used. The system is shown to approach closely the ideal of economical power without pollution.

Mars mission safety

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

Buden, D.

1989-06-01

Precautions that need to be taken to assure safety on a manned Mars mission with nuclear thermal propulsion are briefly considered. What has been learned from the 1955 SNAP-10A operation of a nuclear reactor in space and from the Rover/NERVA project is reviewed. The ways that radiation hazards can be dealt with at various stages of a Mars mission are examined.

A mini-cavity probe reactor.

NASA Technical Reports Server (NTRS)

Hyland, R. E.

1971-01-01

The mini-cavity reactor is a rocket engine concept which combines the high specific impulse from a central gaseous fueled cavity (0.6 m diam) and NERVA type fuel elements in a driver region that is external to a moderator-reflector zone to produce a compact light weight reactor. The overall dimension including a pressure vessel that is located outside of the spherical reactor is approximately 1.21 m in diameter. Specific impulses up to 2000 sec are obtainable for 220 to 890 N of thrust with pressures less than 1000 atm. Powerplant weights including a radiator for disposing of the power in the driver region are between 4600 and 32,000 kg - less than payloads of the shuttle. This reactor could also be used as a test reactor for gas-core, MHD, breeding and materials research.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W. Y.

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc.; a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR); NERVA-derivative; and other concepts are discussed. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggest that full-scale PBR elements could be tested at an average energy deposition of approximately 60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of approximately 100 MW/L may be achievable.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

Application of the Enabler to nuclear electric propulsion

NASA Astrophysics Data System (ADS)

Pierce, Bill L.

This paper describes a power system concept that provides the electric power for a baseline electric propulsion system for a piloted mission to Mars. A 10-MWe space power system is formed by coupling an Enabler reactor with a simple non-recuperated closed Brayton cycle. The Enabler reactor is a gas-cooled reactor based on proven reactor technology developed under the NERVA/Rover programs. The selected power cycle, which uses a helium-xenon mixture at 1920 K at the turbine inlet, is diagramed and described. The specific mass of the power system over the power range from 5 to 70 MWe is given. The impact of operating life on the specific mass of a 10-MWe system is also shown.

Near Earth Asteroid Human Mission Possibilities Using Nuclear Thermal Rocket (NTR) Propulsion

NASA Technical Reports Server (NTRS)

Borowski, Stanley; McCurdy, David R.; Packard, Thomas W.

2012-01-01

The NTR is a proven technology that generates high thrust and has a specific impulse (Isp (is) approximately 900 s) twice that of today's best chemical rockets. During the Rover and NERVA (Nuclear Engine for Rocket Vehicle Applications) programs, twenty rocket reactors were designed, built and ground tested. These tests demonstrated: (1) a wide range of thrust; (2) high temperature carbide-based nuclear fuel; (3) sustained engine operation; (4) accumulated lifetime; and (5) restart capability - all the requirements needed for a human mission to Mars. Ceramic metal fuel was also evaluated as a backup option. In NASA's recent Mars Design reference Architecture (DRA) 5.0 study, the NTR was selected as the preferred propulsion option because of its proven technology, higher performance, lower launch mass, versatile vehicle design, simple assembly, and growth potential. In contrast to other advanced propulsion options, NTP requires no large technology scale-ups. In fact, the smallest engine tested during the Rover program - the 25 klbf 'Pewee' engine is sufficient for a human Mars mission when used in a clustered engine configuration. The 'Copernicus crewed NTR Mars transfer vehicle design developed for DRA 5.0 has significant capability that can enable reusable '1-year' round trip human missions to candidate near Earth asteroids (NEAs) like 1991 JW in 2027, or 2000 SG344 and Apophis in 2028. A robotic precursor mission to 2000 SG344 in late 2023 could provide an attractive Flight Technology Demonstration of a small NTR engine that is scalable to the 25 klbf-class engine used for human missions 5 years later. In addition to the detailed scientific data gathered from on-site inspection, human NEA missions would also provide a valuable 'check out' function for key elements of the NTR transfer vehicle (its propulsion module, TransHab and life support systems, etc.) in a 'deep space' environment prior to undertaking the longer duration Mars orbital and landing missions that would follow. The initial mass in low Earth orbit required for a mission to Apophis is approximately 323 t consisting of the NTR propulsion module ((is) approximately 138 t), the integrated saddle truss and LH2 drop tank assembly ((is) approximately 123 t), and the 6-crew payload element ((is) approximately 62 t). The later includes a multi-mission Space Excursion Vehicle (MMSEV) used for close-up examination and sample gathering. The total burn time and required restarts on the three 25 klbf 'Pewee-class' engines operating at Isp (is) approximately 906 s, are approximately 76.2 minutes and 4, respectively, well below the 2 hours and 27 restarts demonstrated on the NERVA eXperimental Engine, the NRX-XE. The paper examines the benefits, requirements and characteristics of using NTP for the above NEA missions. The impacts on vehicle design of HLV payload volume and lift capability, crew size, and reusability are also quantified.

Space reactor fuel element testing in upgraded TREAT

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

Todosow, M.; Bezler, P.; Ludewig, H.

1993-01-14

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

Space reactor fuel element testing in upgraded TREAT

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

Todosow, M.; Bezler, P.; Ludewig, H.

1993-05-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

Nuclear Propulsion in Space (1968)

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

None

Project NERVA was an acronym for Nuclear Engine for Rocket Vehicle Application, a joint program of the U.S. Atomic Energy Commission and NASA managed by the Space Nuclear Propulsion Office (SNPO) at the Nuclear Rocket Development Station in Jackass Flats, Nevada U.S.A. Between 1959 and 1972, the Space Nuclear Propulsion Office oversaw 23 reactor tests, both the program and the office ended at the end of 1972.

Nuclear Propulsion in Space (1968)

ScienceCinema

None

2018-01-16

Project NERVA was an acronym for Nuclear Engine for Rocket Vehicle Application, a joint program of the U.S. Atomic Energy Commission and NASA managed by the Space Nuclear Propulsion Office (SNPO) at the Nuclear Rocket Development Station in Jackass Flats, Nevada U.S.A. Between 1959 and 1972, the Space Nuclear Propulsion Office oversaw 23 reactor tests, both the program and the office ended at the end of 1972.

Exploratory study of several advanced nuclear-MHD power plant systems.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

1973-01-01

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

Nuclear Engine System Simulation (NESS) version 2.0

NASA Technical Reports Server (NTRS)

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

1993-01-01

The topics are presented in viewgraph form and include the following; nuclear thermal propulsion (NTP) engine system analysis program development; nuclear thermal propulsion engine analysis capability requirements; team resources used to support NESS development; expanded liquid engine simulations (ELES) computer model; ELES verification examples; NESS program development evolution; past NTP ELES analysis code modifications and verifications; general NTP engine system features modeled by NESS; representative NTP expander, gas generator, and bleed engine system cycles modeled by NESS; NESS program overview; NESS program flow logic; enabler (NERVA type) nuclear thermal rocket engine; prismatic fuel elements and supports; reactor fuel and support element parameters; reactor parameters as a function of thrust level; internal shield sizing; and reactor thermal model.

Nuclear modules for space electric propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1998-01-01

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

Nuclear thermal rocket workshop reference system Rover/NERVA

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.

1991-01-01

The Rover/NERVA engine system is to be used as a reference, against which each of the other concepts presented in the workshop will be compared. The following topics are reviewed: the operational characteristics of the nuclear thermal rocket (NTR); the accomplishments of the Rover/NERVA programs; and performance characteristics of the NERVA-type systems for both Mars and lunar mission applications. Also, the issues of ground testing, NTR safety, NASA's nuclear propulsion project plans, and NTR development cost estimates are briefly discussed.

Nuclear design of a vapor core reactor for space nuclear propulsion

NASA Astrophysics Data System (ADS)

Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

1993-01-01

Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

Cryochemical and CVD processing of shperical carbide fuels for propulsion reactors

NASA Astrophysics Data System (ADS)

Blair, H. Thomas; Carroll, David W.; Matthews, R. Bruce

1991-01-01

Many of the nuclear propulsion reactor concepts proposed for a manned mission to Mars use a coated spherical particle fuel form similar to that used in the Rover and NERVA propulsion reactors. The formation of uranium dicarbide microspheres using a cryochemical process and the coating of the UC2 spheres with zirconium carbide using chemical vapor deposition are being developed at Los Alamos National Laboratory. The cryochemical process is described with a discussion of the variables affecting the sphere formation and carbothermic reduction to produce UC2 spheres from UO2. Emphasis is placed on minimizing the wastes produced by the process. The ability to coat particles with ZrC was recaptured, and improvements in the process and equipment were developed. Volatile organometallic precursors were investigated as alternatives to the original ZrCl4 precursor.

Multi-megawatt power system trade study

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Schnitzler, Bruce G.; Parks, Benjamin T.

2002-01-01

A concept study was undertaken to evaluate potential multi-megawatt power sources for nuclear electric propulsion. The nominal electric power requirement was set at 15 MWe with an assumed mission profile of 120 days at full power, 60 days in hot standby, and another 120 days of full power, repeated several times for 7 years of service. Two configurations examined were (1) a gas-cooled reactor based on the NERVA Derivative design, operating a closed cycle Brayton power conversion system; and (2) a molten metal-cooled reactor based on SP-100 technology, driving a boiling potassium Rankine power conversion system. This study considered the relative merits of these two systems, seeking to optimize the specific mass. Conclusions were that either concept appeared capable of reaching the specific mass goal of 3-5 kg/kWe estimated to be needed for this class of mission, though neither could be realized without substantial development in reactor fuels technology, thermal radiator mass and volume efficiency, and power conversion and distribution electronics and systems capable of operating at high temperatures. The gas-Brayton system showed a specific mass advantage (3.17 vs 6.43 kg/kWe for the baseline cases) under the set of assumptions used and eliminated the need to deal with two-phase working fluid flows in the microgravity environment of space. .

The past as prologue - A look at historical flight qualifications for space nuclear systems

NASA Technical Reports Server (NTRS)

Bennett, Gary L.

1992-01-01

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

The past as prologue - A look at historical flight qualifications for space nuclear systems

NASA Astrophysics Data System (ADS)

Bennett, Gary L.

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

Nuclear safety

NASA Technical Reports Server (NTRS)

Buden, D.

1991-01-01

Topics dealing with nuclear safety are addressed which include the following: general safety requirements; safety design requirements; terrestrial safety; SP-100 Flight System key safety requirements; potential mission accidents and hazards; key safety features; ground operations; launch operations; flight operations; disposal; safety concerns; licensing; the nuclear engine for rocket vehicle application (NERVA) design philosophy; the NERVA flight safety program; and the NERVA safety plan.

An Historical Perspective of the NERVA Nuclear Rocket Engine Technology Program

NASA Technical Reports Server (NTRS)

Robbins, W. H.; Finger, H. B.

1991-01-01

Nuclear rocket research and development was initiated in the United States in 1955 and is still being pursued to a limited extent. The major technology emphasis occurred in the decade of the 1960s and was primarily associated with the Rover/NERVA Program where the technology for a nuclear rocket engine system for space application was developed and demonstrated. The NERVA (Nuclear Engine for Rocket Vehicle Application) technology developed twenty years ago provides a comprehensive and viable propulsion technology base that can be applied and will prove to be valuable for application to the NASA Space Exploration Initiative (SEI). This paper, which is historical in scope, provides an overview of the conduct of the NERVA Engine Program, its organization and management, development philosophy, the engine configuration, and significant accomplishments.

Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

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

Bruce G. Schnitzler

Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selectedmore » over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse ({approx}900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as open loop systems for direct nuclear thermal propulsion. Although a number of fast spectrum reactor and engine designs suitable for direct nuclear thermal propulsion were proposed and designed, none were built. This report summarizes status results of evaluations of small nuclear reactor designs suitable for direct nuclear thermal propulsion.« less

Neurexin dysfunction in adult neurons results in autistic-like behavior in mice.

PubMed

Rabaneda, Luis G; Robles-Lanuza, Estefanía; Nieto-González, José Luis; Scholl, Francisco G

2014-07-24

Autism spectrum disorders (ASDs) comprise a group of clinical phenotypes characterized by repetitive behavior and social and communication deficits. Autism is generally viewed as a neurodevelopmental disorder where insults during embryonic or early postnatal periods result in aberrant wiring and function of neuronal circuits. Neurexins are synaptic proteins associated with autism. Here, we generated transgenic βNrx1ΔC mice in which neurexin function is selectively impaired during late postnatal stages. Whole-cell recordings in cortical neurons show an impairment of glutamatergic synaptic transmission in the βNrx1ΔC mice. Importantly, mutant mice exhibit autism-related symptoms, such as increased self-grooming, deficits in social interactions, and altered interaction for nonsocial olfactory cues. The autistic-like phenotype of βNrx1ΔC mice can be reversed after removing the mutant protein in aged animals. The defects resulting from disruption of neurexin function after the completion of embryonic and early postnatal development suggest that functional impairment of mature circuits can trigger autism-related phenotypes. Copyright © 2014 The Authors. Published by Elsevier Inc. All rights reserved.

Historical flight qualifications of space nuclear systems

NASA Astrophysics Data System (ADS)

Bennett, Gary L.

1997-01-01

An overview is presented of the qualification programs for the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions; the SNAP-10A space reactor; the Nuclear Engine for Rocket Vehicle Applications (NERVA); the F-1 chemical rocket engine used on the Saturn-V Apollo lunar missions; and the Space Shuttle Main Engines (SSMEs). Some similarities and contrasts between the qualification testing employed on these five programs will be noted. One common thread was that in each of these successful programs there was an early focus on component and subsystem tests to uncover and correct problems.

The response of claudin-like transmembrane septate junction proteins to altered environmental ion levels in the larval mosquito Aedes aegypti.

PubMed

Jonusaite, Sima; Kelly, Scott P; Donini, Andrew

2016-07-01

Septate junctions (SJs) occlude the paracellular pathway and function as paracellular diffusion barriers within invertebrate epithelia. However, integral components of SJs and their contribution to barrier properties have received considerably less attention than those of vertebrate occluding junctions. In arthropods, SJ proteins have only been identified in Drosophila and among these are three integral claudin-like proteins, Megatrachea (Mega), Sinuous (Sinu) and Kune-kune (Kune), as well as a receptor-like transmembrane SJ protein known as Neurexin IV (Nrx IV). In this study, mega, sinu, kune and nrx IV are identified and characterized in aquatic larvae of the mosquito Aedes aegypti and a role for these proteins in ionoregulatory homeostasis is considered. Transcripts encoding Mega, Sinu, Kune and Nrx IV were found in iono/osmoregulatory tissues such as the midgut, Malpighian tubules, hindgut and anal papillae, but abundance was greater in the hindgut and anal papillae. Using immunohistochemical and western blot analysis it was found that Kune localized to the regions of intercellular contact between epithelial cells of the rectum and posterior midgut and in the apical membrane domain of the syncytial epithelium of anal papillae. To investigate a potential role for integral SJ proteins in larval A. aegypti iono/osmoregulation, abundance was examined in animals reared in freshwater or brackish water (30 % seawater). In iono/osmoregulatory epithelia, larvae exhibited tissue-specific alterations in mega mRNA and Kune protein abundance, but not sinu or nrx IV mRNA. These studies provide a first look at the potential contribution of integral SJ components to iono/osmoregulatory homeostasis in an aquatic invertebrate.

Drosophila contactin, a homolog of vertebrate contactin, is required for septate junction organization and paracellular barrier function.

PubMed

Faivre-Sarrailh, Catherine; Banerjee, Swati; Li, Jingjun; Hortsch, Michael; Laval, Monique; Bhat, Manzoor A

2004-10-01

Septate junctions (SJs) in epithelial and neuronal cells play an important role in the formation and maintenance of charge and size selective barriers. They form the basis for the ensheathment of nerve fibers in Drosophila and for the attachment of myelin loops to axonal surface in vertebrates. The cell-adhesion molecules NRX IV/Caspr/Paranodin (NCP1), contactin and Neurofascin-155 (NF-155) are all present at the vertebrate axo-glial SJs. Mutational analyses have shown that vertebrate NCP1 and its Drosophila homolog, Neurexin IV (NRX IV) are required for the formation of SJs. In this study, we report the genetic, molecular and biochemical characterization of the Drosophila homolog of vertebrate contactin, CONT. Ultrastructural and dye-exclusion analyses of Cont mutant embryos show that CONT is required for organization of SJs and paracellular barrier function. We show that CONT, Neuroglian (NRG) (Drosophila homolog of NF-155) and NRX IV are interdependent for their SJ localization and these proteins form a tripartite complex. Hence, our data provide evidence that the organization of SJs is dependent on the interactions between these highly conserved cell-adhesion molecules.

Nuclear Engine System Simulation (NESS). Version 2.0: Program user's guide

NASA Technical Reports Server (NTRS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman

1993-01-01

This Program User's Guide discusses the Nuclear Thermal Propulsion (NTP) engine system design features and capabilities modeled in the Nuclear Engine System Simulation (NESS): Version 2.0 program (referred to as NESS throughout the remainder of this document), as well as its operation. NESS was upgraded to include many new modeling capabilities not available in the original version delivered to NASA LeRC in Dec. 1991, NESS's new features include the following: (1) an improved input format; (2) an advanced solid-core NERVA-type reactor system model (ENABLER 2); (3) a bleed-cycle engine system option; (4) an axial-turbopump design option; (5) an automated pump-out turbopump assembly sizing option; (6) an off-design gas generator engine cycle design option; (7) updated hydrogen properties; (8) an improved output format; and (9) personal computer operation capability. Sample design cases are presented in the user's guide that demonstrate many of the new features associated with this upgraded version of NESS, as well as design modeling features associated with the original version of NESS.

Fluid dynamics computer programs for NERVA turbopump

NASA Technical Reports Server (NTRS)

Brunner, J. J.

1972-01-01

During the design of the NERVA turbopump, numerous computer programs were developed for the analyses of fluid dynamic problems within the machine. Program descriptions, example cases, users instructions, and listings for the majority of these programs are presented.

The Satellite Nuclear Power Station - An option for future power generation.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.

1973-01-01

A new concept in nuclear power generation is being explored which essentially eliminates major objections to nuclear power. The Satellite Nuclear Power Station, remotely operated in synchronous orbit, would transmit power safely to the ground by a microwave beam. Fuel reprocessing would take place in space and no radioactive materials would ever be returned to earth. Even the worst possible accident to such a plant should have negligible effect on the earth. An exploratory study of a satellite nuclear power station to provide 10,000 MWe to the earth has shown that the system could weigh about 20 million pounds and cost less than $1000/KWe. An advanced breeder reactor operating with an MHD power cycle could achieve an efficiency of about 50% with a 1100 K radiator temperature. If a hydrogen moderated gas core reactor is used, its breeding ratio of 1.10 would result in a fuel doubling time of a few years. A rotating fluidized bed or NERVA type reactor might also be used. The efficiency of power transmission from synchronous orbit would range from 70% to 80%.

Subscale Validation of the Subsurface Active Filtration of Exhaust (SAFE) Approach to NTP Ground Testing

NASA Technical Reports Server (NTRS)

Marshall, William M.; Borowski, Stanley K.; Bulman, Mel; Joyner, Russell; Martin, Charles R.

2015-01-01

Brief History of NTP: Project Rover Began in 1950s by Los Alamos Scientific Labs (now Los Alamos National Labs) and ran until 1970s Tested a series of nuclear reactor engines of varying size at Nevada Test Site (now Nevada National Security Site) Ranged in scale from 111 kN (25 klbf) to 1.1 MN (250 klbf) Included Nuclear Furnace-1 tests Demonstrated the viability and capability of a nuclear rocket engine test program One of Kennedys 4 goals during famous moon speech to Congress Nuclear Engines for Rocket Vehicle Applications (NERVA) Atomic Energy Commission and NASA joint venture started in 1964 Parallel effort to Project Rover was focused on technology demonstration Tested XE engine, a 245-kN (55-klbf) engine to demonstrate startup shutdown sequencing. Hot-hydrogen stream is passed directly through fuel elements potential for radioactive material to be eroded into gaseous fuel flow as identified in previous programs NERVA and Project Rover (1950s-70s) were able to test in open atmosphere similar to conventional rocket engine test stands today Nuclear Furance-1 tests employed a full scrubber system Increased government and environmental regulations prohibit the modern testing in open atmosphere. Since the 1960s, there has been an increasing cessation on open air testing of nuclear material Political and national security concerns further compound the regulatory environment

Historical flight qualifications of space nuclear systems

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

Bennett, G.L.

1997-01-01

An overview is presented of the qualification programs for the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions; the SNAP-10A space reactor; the Nuclear Engine for Rocket Vehicle Applications (NERVA); the F-1 chemical rocket engine used on the Saturn-V Apollo lunar missions; and the Space Shuttle Main Engines (SSMEs). Some similarities and contrasts between the qualification testing employed on these five programs will be noted. One common thread was that in each of these successful programs there was an early focus on component and subsystem tests to uncover and correct problems. {copyright} {italmore » 1997 American Institute of Physics.}« less

Science in Flux: NASA's Nuclear Program at Plum Brook Station 1955-2005

NASA Technical Reports Server (NTRS)

Bowles, Mark D.

2006-01-01

Science in Flux traces the history of one of the most powerful nuclear test reactors in the United States and the only nuclear facility ever built by NASA. In the late 1950's NASA constructed Plum Brook Station on a vast tract of undeveloped land near Sandusky, Ohio. Once fully operational in 1963, it supported basic research for NASA's nuclear rocket program (NERVA). Plum Brook represents a significant, if largely forgotten, story of nuclear research, political change, and the professional culture of the scientists and engineers who devoted their lives to construct and operate the facility. In 1973, after only a decade of research, the government shut Plum Brook down before many of its experiments could be completed. Even the valiant attempt to redefine the reactor as an environmental analysis tool failed, and the facility went silent. The reactors lay in costly, but quiet standby for nearly a quarter-century before the Nuclear Regulatory Commission decided to decommission the reactors and clean up the site. The history of Plum Brook reveals the perils and potentials of that nuclear technology. As NASA, Congress, and space enthusiasts all begin looking once again at the nuclear option for sending humans to Mars, the echoes of Plum Brook's past will resonate with current policy and space initiatives.

Neurexin and Neuroligin Mediate Retrograde Synaptic Inhibition in C. elegans

PubMed Central

Hu, Zhitao; Hom, Sabrina; Kudze, Tambudzai; Tong, Xia-Jing; Choi, Seungwon; Aramuni, Gayane; Zhang, Weiqi; Kaplan, Joshua M.

2013-01-01

The synaptic adhesion molecules Neurexin and Neuroligin alter the development and function of synapses and are linked to autism in humans. We find that C. elegans Neurexin (NRX-1) and Neuroligin (NLG-1) mediate a retrograde synaptic signal that inhibits neurotransmitter release at neuromuscular junctions. Retrograde signaling was induced in mutants lacking a muscle microRNA (miR-1) and was blocked in mutants lacking NLG-1 or NRX-1. Release was rapid and abbreviated when the retrograde signal was on whereas release was slow and prolonged when retrograde signaling was blocked. The retrograde signal adjusted release kinetics by inhibiting exocytosis of synaptic vesicles (SVs) that are distal to the site of calcium entry. Inhibition of release was mediated by increased pre-synaptic levels of Tomosyn, an inhibitor of SV fusion. PMID:22859820

Three-dimensional particle tracking velocimetry algorithm based on tetrahedron vote

NASA Astrophysics Data System (ADS)

Cui, Yutong; Zhang, Yang; Jia, Pan; Wang, Yuan; Huang, Jingcong; Cui, Junlei; Lai, Wing T.

2018-02-01

A particle tracking velocimetry algorithm based on tetrahedron vote, which is named TV-PTV, is proposed to overcome the limited selection problem of effective algorithms for 3D flow visualisation. In this new cluster-matching algorithm, tetrahedrons produced by the Delaunay tessellation are used as the basic units for inter-frame matching, which results in a simple algorithmic structure of only two independent preset parameters. Test results obtained using the synthetic test image data from the Visualisation Society of Japan show that TV-PTV presents accuracy comparable to that of the classical algorithm based on new relaxation method (NRX). Compared with NRX, TV-PTV possesses a smaller number of loops in programming and thus a shorter computing time, especially for large particle displacements and high particle concentration. TV-PTV is confirmed practically effective using an actual 3D wake flow.

Engine System Model Development for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Nelson, Karl W.; Simpson, Steven P.

2006-01-01

In order to design, analyze, and evaluate conceptual Nuclear Thermal Propulsion (NTP) engine systems, an improved NTP design and analysis tool has been developed. The NTP tool utilizes the Rocket Engine Transient Simulation (ROCETS) system tool and many of the routines from the Enabler reactor model found in Nuclear Engine System Simulation (NESS). Improved non-nuclear component models and an external shield model were added to the tool. With the addition of a nearly complete system reliability model, the tool will provide performance, sizing, and reliability data for NERVA-Derived NTP engine systems. A new detailed reactor model is also being developed and will replace Enabler. The new model will allow more flexibility in reactor geometry and include detailed thermal hydraulics and neutronics models. A description of the reactor, component, and reliability models is provided. Another key feature of the modeling process is the use of comprehensive spreadsheets for each engine case. The spreadsheets include individual worksheets for each subsystem with data, plots, and scaled figures, making the output very useful to each engineering discipline. Sample performance and sizing results with the Enabler reactor model are provided including sensitivities. Before selecting an engine design, all figures of merit must be considered including the overall impacts on the vehicle and mission. Evaluations based on key figures of merit of these results and results with the new reactor model will be performed. The impacts of clustering and external shielding will also be addressed. Over time, the reactor model will be upgraded to design and analyze other NTP concepts with CERMET and carbide fuel cores.

Pratt & Whitney ESCORT derivative for mars surface power

NASA Astrophysics Data System (ADS)

Feller, Gerald J.; Joyner, Russell

1999-01-01

The purpose of this paper is to address the applicability of a common reactor system design from the Pratt & Whitney ESCORT nuclear thermal rocket engine concept to support current NASA mars surface-based power requirements. The ESCORT is a bimodal engine capable of supporting a wide range of propulsive thermal and vehicle electrical power requirements. The ESCORT engine is powered by a fast-spectrum beryllium-reflected CERMET-fueled nuclear reactor. In addition to an expander cycle propulsive mode, the ESCORT is capable of operating in an electrical power mode. In this mode, the reactor is used to heat a mixture of helium and xenon to drive a closed-loop Brayton cycle in order to generate electrical energy. Recent Design Reference Mission requirements (DRM) from NASA Johnson Space Center and NASA Lewis Research Center studies in 1997 and 1998 have detailed upgraded requirements for potential mars transfer missions. The current NASA DRM requires a nuclear thermal propulsion system capable of delivering total mission requirements of 200170 N (45000 lbf) thrust and 50 kWe of spacecraft electrical power. Additionally, these requirements detailed a surface power system capable of providing approximately 160 kW of electrical energy over an approximate 10 year period within a given weight and volume envelope. Current NASA studies use a SP-100 reactor (0.8 MT) and a NERVA derivative (1.6 MT) as baseline systems. A mobile power cart of approximate dimensions 1.7 m×4.5 m×4.4 m has been conceptualized to transport the reactor power system on the Mars Surface. The 63.25 cm diameter and 80.25 cm height of the ESCORT and its 1.3 MT of weight fit well within the current weight and volume target range of the NASA DRM requirements. The modifications required to the ESCORT reactor system to support this upgraded electrical power requirements along with operation in the Martian atmospheric conditions are addressed in this paper. Sufficient excess reactivity and burnup capability were designed into the ESCORT reactor system to support these upgraded requirements. Only slight modifications to reactor hardware were required to address any environmental considerations. These modifications involved sealing any refractory metal alloy components from the CO2 in the Martian Atmosphere. Also, the Brayton cycle Power Conversion Unit (PCU) hardware was modified to support the upgraded requirements. This paper discusses the design analysis performed and provides information on the final common reactor concept to be used on the Mars surface to support manned missions.

An historical collection of papers on nuclear thermal propulsion

NASA Astrophysics Data System (ADS)

The present volume of historical papers on nuclear thermal propulsion (NTP) encompasses NTP technology development regarding solid-core NTP technology, advanced concepts from the early years of NTP research, and recent activities in the field. Specific issues addressed include NERVA rocket-engine technology, the development of nuclear rocket propulsion at Los Alamos, fuel-element development, reactor testing for the Rover program, and an overview of NTP concepts and research emphasizing two decades of NASA research. Also addressed are the development of the 'nuclear light bulb' closed-cycle gas core and a demonstration of a fissioning UF6 gas in an argon vortex. The recent developments reviewed include the application of NTP to NASA's Lunar Space Transportation System, the use of NTP for the Space Exploration Initiative, and the development of nuclear rocket engines in the former Soviet Union.

Modeling Transients and Designing a Passive Safety System for a Nuclear Thermal Rocket Using Relap5

NASA Astrophysics Data System (ADS)

Khatry, Jivan

Long-term high payload missions necessitate the need for nuclear space propulsion. Several nuclear reactor types were investigated by the Nuclear Engine for Rocket Vehicle Application (NERVA) program of National Aeronautics and Space Administration (NASA). Study of planned/unplanned transients on nuclear thermal rockets is important due to the need for long-term missions. A NERVA design known as the Pewee I was selected for this purpose. The following transients were run: (i) modeling of corrosion-induced blockages on the peripheral fuel element coolant channels and their impact on radiation heat transfer in the core, and (ii) modeling of loss-of-flow-accidents (LOFAs) and their impact on radiation heat transfer in the core. For part (i), the radiation heat transfer rate of blocked channels increases while their neighbors' decreases. For part (ii), the core radiation heat transfer rate increases while the flow rate through the rocket system is decreased. However, the radiation heat transfer decreased while there was a complete LOFA. In this situation, the peripheral fuel element coolant channels handle the majority of the radiation heat transfer. Recognizing the LOFA as the most severe design basis accident, a passive safety system was designed in order to respond to such a transient. This design utilizes the already existing tie rod tubes and connects them to a radiator in a closed loop. Hence, this is basically a secondary loop. The size of the core is unchanged. During normal steady-state operation, this secondary loop keeps the moderator cool. Results show that the safety system is able to remove the decay heat and prevent the fuel elements from melting, in response to a LOFA and subsequent SCRAM.

Mars Mission Analysis Trades Based on Legacy and Future Nuclear Propulsion Options

NASA Astrophysics Data System (ADS)

Joyner, Russell; Lentati, Andrea; Cichon, Jaclyn

2007-01-01

The purpose of this paper is to discuss the results of mission-based system trades when using a nuclear thermal propulsion (NTP) system for Solar System exploration. The results are based on comparing reactor designs that use a ceramic-metallic (CERMET), graphite matrix, graphite composite matrix, or carbide matrix fuel element designs. The composite graphite matrix and CERMET designs have been examined for providing power as well as propulsion. Approaches to the design of the NTP to be discussed will include an examination of graphite, composite, carbide, and CERMET core designs and the attributes of each in regards to performance and power generation capability. The focus is on NTP approaches based on tested fuel materials within a prismatic fuel form per the Argonne National Laboratory testing and the ROVER/NERVA program. NTP concepts have been examined for several years at Pratt & Whitney Rocketdyne for use as the primary propulsion for human missions beyond earth. Recently, an approach was taken to examine the design trades between specific NTP concepts; NERVA-based (UC)C-Graphite, (UC,ZrC)C-Composite, (U,Zr)C-Solid Carbide and UO2-W CERMET. Using Pratt & Whitney Rocketdyne's multidisciplinary design analysis capability, a detailed mission and vehicle model has been used to examine how several of these NTP designs impact a human Mars mission. Trends for the propulsion system mass as a function of power level (i.e. thrust size) for the graphite-carbide and CERMET designs were established and correlated against data created over the past forty years. These were used for the mission trade study. The resulting mission trades presented in this paper used a comprehensive modeling approach that captures the mission, vehicle subsystems, and NTP sizing.

Propulsion at the Marshall Space Flight Center - A brief history

NASA Technical Reports Server (NTRS)

Jones, L. W.; Fisher, M. F.; Mccool, A. A.; Mccarty, J. P.

1991-01-01

The history of propulsion development at the NASA Marshall Space Flight Center is summarized, beginning with the development of the propulsion system for the Redstone missile. This course of propulsion development continues through the Jupiter IRBM, the Saturn family of launch vehicles and the engines that powered them, the Centaur upper stage and RL-10 engine, the Reactor In-Flight Test stage and the NERVA nuclear engine. The Space Shuttle Main Engine and Solid Rocket Boosters are covered, as are spacecraft propulsion systems, including the reaction control systems for the High Energy Astronomy Observatory and the Space Station. The paper includes a description of several technology efforts such as those in high pressure turbomachinery, aerospike engines, and the AS203 cyrogenic fluid management flight experiment. These and other propulsion projects are documented, and the scope of activities in support of these efforts at Marshall delineated.

NERVA dynamic analysis methodology, SPRVIB

NASA Technical Reports Server (NTRS)

Vronay, D. F.

1972-01-01

The general dynamic computer code called SPRVIB (Spring Vib) developed in support of the NERVA (nuclear engine for rocket vehicle application) program is described. Using normal mode techniques, the program computes kinematical responses of a structure caused by various combinations of harmonic and elliptic forcing functions or base excitations. Provision is made for a graphical type of force or base excitation input to the structure. A description of the required input format and a listing of the program are presented, along with several examples illustrating the use of the program. SPRVIB is written in FORTRAN 4 computer language for use on the CDC 6600 or the IBM 360/75 computers.

Low Thrust, Deep Throttling, US/CIS Integrated NTRE

NASA Astrophysics Data System (ADS)

Culver, Donald W.; Kolganov, Vyacheslav; Rochow, Richard F.

1994-07-01

In 1993 our international team performed a follow-on ``Nuclear Thermal Rocket Engine (NTRE) Extended Life Feasibility Assessment'' study for the Nuclear Propulsion Office (NPO) at NASAs Lewis Research Center. The main purpose of this study was to complete the 1992 study matrix to assess NTRE designs at thrust levels of 22.5, 11.3, and 6.8 tonnes, using Commonwealth of Independent States (CIS) reactor technology. An additional Aerojet goal was to continue improving the NTRE concept we had generated. Deep throttling, mission performance optimized engine design parametrics, and reliability/cost enhancing engine system simplifications were studied, because they seem to be the last three basic design improvements sorely needed by post-NERVA NTRE. Deep throttling improves engine life by eliminating damaging thermal and mechanical shocks caused by after-cooling with pulsed coolant flow. Alternately, it improves mission performance with steady flow after-cooling by minimizing reactor over-cooling. Deep throttling also provides a practical transition from high pressures and powers of the high thrust power cycle to the low pressures and powers of our electric power generating mode. Two deep throttling designs are discussed; a workable system that was studied and a simplified system that is recommended for future study. Mission-optimized engine thrust/weight (T/W) and Isp predictions are included along with system flow schemes and concept sketches.

Nuclear thermal propulsion transportation systems for lunar/Mars exploration

NASA Technical Reports Server (NTRS)

Clark, John S.; Borowski, Stanley K.; Mcilwain, Melvin C.; Pellaccio, Dennis G.

1992-01-01

Nuclear thermal propulsion technology development is underway at NASA and DoE for Space Exploration Initiative (SEI) missions to Mars, with initial near-earth flights to validate flight readiness. Several reactor concepts are being considered for these missions, and important selection criteria will be evaluated before final selection of a system. These criteria include: safety and reliability, technical risk, cost, and performance, in that order. Of the concepts evaluated to date, the Nuclear Engine for Rocket Vehicle Applications (NERVA) derivative (NDR) is the only concept that has demonstrated full power, life, and performance in actual reactor tests. Other concepts will require significant design work and must demonstrate proof-of-concept. Technical risk, and hence, development cost should therefore be lowest for the concept, and the NDR concept is currently being considered for the initial SEI missions. As lighter weight, higher performance systems are developed and validated, including appropriate safety and astronaut-rating requirements, they will be considered to support future SEI application. A space transportation system using a modular nuclear thermal rocket (NTR) system for lunar and Mars missions is expected to result in significant life cycle cost savings. Finally, several key issues remain for NTR's, including public acceptance and operational issues. Nonetheless, NTR's are believed to be the 'next generation' of space propulsion systems - the key to space exploration.

Peat δ13Ccelluose-recorded wetting trend during the past 8000 years in the southern Altai Mountains, northern Xinjiang, NW China

NASA Astrophysics Data System (ADS)

Zhang, Dongliang; Feng, Zhaodong; Yang, Yunpeng; Lan, Bo; Ran, Min; Mu, Guijin

2018-05-01

There have been large discrepancies in the proposed mechanisms accounting for the wetting trend since ∼8.0 cal. kyr BP in the Altai Mountains and the surrounding areas. To validate or invalidate the widely reported wetting trend, we obtained a carbon isotope of cellulose (δ13Ccelluose)-recorded warm-season moisture history from a Narenxia (NRX) peat core in the southern Altai Mountains, northern Xinjiang, NW China. The δ13Ccelluose-recorded warm-season moisture reconstruction of the NRX peat core provides a strong support to the widely-reported proposition that the climate was generally dry before ∼8.0 cal. kyr BP and was changed to a wetting trend during the past ∼8000 years in the Altai Mountains and the surrounding areas. The wetting trend since ∼8.0 cal. kyr BP well resembles the increasing trend of the reconnaissance drought index (RDI) that was calculated on the basis of pollen-inferred temperature and precipitation data from the same core. The resemblance implies that the wetting trend during the past ∼8000 years resulted from the combined effect of temperature and precipitation.

Carbide fuels for nuclear thermal propulsion

NASA Astrophysics Data System (ADS)

Matthews, R. B.; Blair, H. T.; Chidester, K. M.; Davidson, K. V.; Stark, W. E.; Storms, E. K.

1991-09-01

A renewed interest in manned exploration of space has revitalized interest in the potential for advancing nuclear rocket technology developed during the 1960's. Carbide fuel performance, melting point, stability, fabricability and compatibility are key technology issues for advanced Nuclear Thermal Propulsion reactors. The Rover fuels development ended with proven carbide fuel forms with demonstrated operating temperatures up to 2700 K for over 100 minutes. The next generation of nuclear rockets will start where the Rover technology ended, but with a more rigorous set of operating requirements including operating lifetime to 10 hours, operating temperatures greater that 3000 K, low fission product release, and compatibility. A brief overview of Rover/NERVA carbide fuel development is presented. A new fuel form with the highest potential combination of operating temperature and lifetime is proposed that consists of a coated uranium carbide fuel sphere with built-in porosity to contain fission products. The particles are dispersed in a fiber reinforced ZrC matrix to increase thermal shock resistance.

An integral nuclear power and propulsion system concept

NASA Astrophysics Data System (ADS)

Choong, Phillip T.; Teofilo, Vincent L.; Begg, Lester L.; Dunn, Charles; Otting, William

An integral space power concept provides both the electrical power and propulsion from a common heat source and offers superior performance capabilities over conventional orbital insertion using chemical propulsion systems. This paper describes a hybrid (bimodal) system concept based on a proven, inherently safe solid fuel form for the high temperature reactor core operation and rugged planar thermionic energy converter for long-life steady state electric power production combined with NERVA-based rocket technology for propulsion. The integral system is capable of long-life power operation and multiple propulsion operations. At an optimal thrust level, the integral system can maintain the minimal delta-V requirement while minimizing the orbital transfer time. A trade study comparing the overall benefits in placing large payloads to GEO with the nuclear electric propulsion option shows superiority of nuclear thermal propulsion. The resulting savings in orbital transfer time and the substantial reduction of overall lift requirement enables the use of low-cost launchers for several near-term military satellite missions.

The influence of radiation shielding on reusable nuclear shuttle design

NASA Technical Reports Server (NTRS)

Littman, T. M.; Garcia, D.

1972-01-01

Alternate reusable nuclear shuttle configurations were synthesized and evaluated. Particular attention was given to design factors which reduced tank exposure to direct and scattered radiation, increased payload-engine separation, and improved self-shielding by the LH2 propellant. The most attractive RNS concept in terms of cost effectiveness consists of a single conical aft bulkhead tank with a high fineness ratio. Launch is accomplished by the INT-21 with the tank positioned in the inverted attitude. The NERVA engine is delivered to orbit separately where final stage assembly and checkout are accomplished. This approach is consistent with NERVA definition criteria and required operating procedures to support an economically viable nuclear shuttle transportation program in the post-1980 period.

The Rationale/Benefits of Nuclear Thermal Rocket Propulsion for NASA's Lunar Space Transportation System

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.

1994-01-01

The solid core nuclear thermal rocket (NTR) represents the next major evolutionary step in propulsion technology. With its attractive operating characteristics, which include high specific impulse (approximately 850-1000 s) and engine thrust-to-weight (approximately 4-20), the NTR can form the basis for an efficient lunar space transportation system (LTS) capable of supporting both piloted and cargo missions. Studies conducted at the NASA Lewis Research Center indicate that an NTR-based LTS could transport a fully-fueled, cargo-laden, lunar excursion vehicle to the Moon, and return it to low Earth orbit (LEO) after mission completion, for less initial mass in LEO than an aerobraked chemical system of the type studied by NASA during its '90-Day Study.' The all-propulsive NTR-powered LTS would also be 'fully reusable' and would have a 'return payload' mass fraction of approximately 23 percent--twice that of the 'partially reusable' aerobraked chemical system. Two NTR technology options are examined--one derived from the graphite-moderated reactor concept developed by NASA and the AEC under the Rover/NERVA (Nuclear Engine for Rocket Vehicle Application) programs, and a second concept, the Particle Bed Reactor (PBR). The paper also summarizes NASA's lunar outpost scenario, compares relative performance provided by different LTS concepts, and discusses important operational issues (e.g., reusability, engine 'end-of life' disposal, etc.) associated with using this important propulsion technology.

Neuroglian, Gliotactin, and the Na+/K+ ATPase are essential for septate junction function in Drosophila

PubMed Central

Genova, Jennifer L.; Fehon, Richard G.

2003-01-01

One essential function of epithelia is to form a barrier between the apical and basolateral surfaces of the epithelium. In vertebrate epithelia, the tight junction is the primary barrier to paracellular flow across epithelia, whereas in invertebrate epithelia, the septate junction (SJ) provides this function. In this study, we identify new proteins that are required for a functional paracellular barrier in Drosophila. In addition to the previously known components Coracle (COR) and Neurexin (NRX), we show that four other proteins, Gliotactin, Neuroglian (NRG), and both the α and β subunits of the Na+/K+ ATPase, are required for formation of the paracellular barrier. In contrast to previous reports, we demonstrate that the Na pump is not localized basolaterally in epithelial cells, but instead is concentrated at the SJ. Data from immunoprecipitation and somatic mosaic studies suggest that COR, NRX, NRG, and the Na+/K+ ATPase form an interdependent complex. Furthermore, the observation that NRG, a Drosophila homologue of vertebrate neurofascin, is an SJ component is consistent with the notion that the invertebrate SJ is homologous to the vertebrate paranodal SJ. These findings have implications not only for invertebrate epithelia and barrier functions, but also for understanding of neuron–glial interactions in the mammalian nervous system. PMID:12782686

Neuroglian, Gliotactin, and the Na+/K+ ATPase are essential for septate junction function in Drosophila.

PubMed

Genova, Jennifer L; Fehon, Richard G

2003-06-09

One essential function of epithelia is to form a barrier between the apical and basolateral surfaces of the epithelium. In vertebrate epithelia, the tight junction is the primary barrier to paracellular flow across epithelia, whereas in invertebrate epithelia, the septate junction (SJ) provides this function. In this study, we identify new proteins that are required for a functional paracellular barrier in Drosophila. In addition to the previously known components Coracle (COR) and Neurexin (NRX), we show that four other proteins, Gliotactin, Neuroglian (NRG), and both the alpha and beta subunits of the Na+/K+ ATPase, are required for formation of the paracellular barrier. In contrast to previous reports, we demonstrate that the Na pump is not localized basolaterally in epithelial cells, but instead is concentrated at the SJ. Data from immunoprecipitation and somatic mosaic studies suggest that COR, NRX, NRG, and the Na+/K+ ATPase form an interdependent complex. Furthermore, the observation that NRG, a Drosophila homologue of vertebrate neurofascin, is an SJ component is consistent with the notion that the invertebrate SJ is homologous to the vertebrate paranodal SJ. These findings have implications not only for invertebrate epithelia and barrier functions, but also for understanding of neuron-glial interactions in the mammalian nervous system.

Alternate Fuel Cycle Technologies/Thorium Fuel Cycle Technology Programs. Quarterly report for period 1 April--30 June 1978

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

Vondra, B.L.

1978-08-01

Voloxidation and dissolution studies: rotary-kiln heat-transfer tests are under way using a small rotary kiln along with the development of a mathematical model to determine kiln-heat-flux profiles necessary to maintain a desired temperature gradient. The erosion/corrosion test for evaluating materials of construction is operational. Fuel from a BWR (Big Rock Point) yielded more fine solid residue on dissolution than in previous tests with PWR fuel. Two additional parametric voloxidation tests with H.B. Robinson fuel compared air vs pure oxygen atmospheres at 550{sup 0}C; overall tritium release and subsequent fuel dissolution were equivalent. Thorium dissolution studies: the dissolution rate of thoriamore » in fluoride-catalyzed 8 to 14 M HNO{sub 3} (100{sup 0}C) was max between 0.04 to 0.06 M HF; at higher fluoride concentrations, ThF{sub 4}.5H{sub 2}O precipitated. The rate of zircaloy dissolution continued to increase with increasing fluoride concentration. Stainless-steel-clad (Th,U)0{sub 2} fuel rods irradiated in the NRX reactor were sheared, voloxidized, and dissolved. {le}10% of the tritium was released during voloxidation in air at 600{sup 0}C. Carbon-14 removal from off-gas and fixation: carbon dioxide removal with Linde 13X molecular sieves to less than 100 ppB was experimentally verified using 300 ppM CO in air. Decontamination factors from 3000 to 7500 were obtained for CO{sub 2} removal in the gas-slurry stirred-tank reactor with CA(OH){sub 2}.or Ba(0H){sub 2}/sup .8H2O./. With Ba(OH){sub 2}.H{sub 2}0{sup 2} in a fixed-bed column, decontamination factors of about 30,000 were obtained.« less

Nuclear Lunar Logistics Study

NASA Technical Reports Server (NTRS)

1963-01-01

This document has been prepared to incorporate all presentation aid material, together with some explanatory text, used during an oral briefing on the Nuclear Lunar Logistics System given at the George C. Marshall Space Flight Center, National Aeronautics and Space Administration, on 18 July 1963. The briefing and this document are intended to present the general status of the NERVA (Nuclear Engine for Rocket Vehicle Application) nuclear rocket development, the characteristics of certain operational NERVA-class engines, and appropriate technical and schedule information. Some of the information presented herein is preliminary in nature and will be subject to further verification, checking and analysis during the remainder of the study program. In addition, more detailed information will be prepared in many areas for inclusion in a final summary report. This work has been performed by REON, a division of Aerojet-General Corporation under Subcontract 74-10039 from the Lockheed Missiles and Space Company. The presentation and this document have been prepared in partial fulfillment of the provisions of the subcontract. From the inception of the NERVA program in July 1961, the stated emphasis has centered around the demonstration of the ability of a nuclear rocket to perform safely and reliably in the space environment, with the understanding that the assignment of a mission (or missions) would place undue emphasis on performance and operational flexibility. However, all were aware that the ultimate justification for the development program must lie in the application of the nuclear propulsion system to the national space objectives.

Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide

NASA Astrophysics Data System (ADS)

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

1993-03-01

A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner. The following discussion summarizes the overall analysis methodology, key assumptions, and capabilities associated with the NESS presents an example problem, and compares the results to related NTP engine system designs. Initial installation instructions and program disks are in Volume 2 of the NESS Program User's Guide.

Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide

NASA Technical Reports Server (NTRS)

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

1993-01-01

A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner. The following discussion summarizes the overall analysis methodology, key assumptions, and capabilities associated with the NESS presents an example problem, and compares the results to related NTP engine system designs. Initial installation instructions and program disks are in Volume 2 of the NESS Program User's Guide.

Ultra high vacuum adhesion testing of NERVA engine materials

NASA Technical Reports Server (NTRS)

1970-01-01

The primary objective of this research program was to determine the effects of surface cleaning and deliberate gaseous contamination on the adhesion behavior of selected candidate materials for use in the NERVA nuclear rocket engine program. Using a torsion balance technique, the relationship between the normal compressive load applied to crossed rod samples and the resultant contact resistance was used to ascertain the extent of adhesion under each set of experimental conditions. In addition to an evaluation of the static adhesion behavior of selected materials combinations, the experimental apparatus was modified to permit a similar investigation relating to the effects of specific tangential displacements of the sample wires, i.e., their sliding friction behavior. During the course of this subcontract, the materials combinations 440 C vs. 440 C. pyrographite vs ZTA graphite, Nbc (graphite) vs. Nbc (graphite), and Electrolize Inconel 718 vs. Au electroplated 302 S/S were evaluated.

NERVA 400E thrust train dynamic analysis

NASA Technical Reports Server (NTRS)

Vronay, D. F.

1972-01-01

The natural frequencies and dynamic responses of the NERVA 400E engine thrust train were determined for nuclear space operations (NSO), and earth-orbital shuttle (EOS) during launch and boost conditions. For NSO, a mini-tank configuration was analyzed with the forward end of the upper truss assumed fixed at the stage/mini-tank interface. For EOS, both a mini-tank and an engine only configuration were analyzed for a specific engine assembly support (EAS) stiffness. For all cases the effect of the shield on dynamic response characteristics was determined by performing parallel analyses with and without the shield. Gimbaling loads were not generated as that effort was scheduled after the termination date. The analysis, while demonstrating the adequacy of the engine design, revealed serious deficiencies in the EAS. Responses at the unsupported ends of the engine are excessive. Responses at the nuclear subsystem interface appear acceptable. It is recommended that additional analysis and design effort be expended upon the EAS to ensure that all engine responses stay within reasonable bounds.

Exhaust gas treatment in testing nuclear rocket engines

NASA Astrophysics Data System (ADS)

Zweig, Herbert R.; Fischler, Stanley; Wagner, William R.

1993-01-01

With the exception of the last test series of the Rover program, Nuclear Furnace 1, test-reactor and rocket engine hydrogen gas exhaust generated during the Rover/NERVA program was released directly to the atmosphere, without removal of the associated fission products and other radioactive debris. Current rules for nuclear facilities (DOE Order 5480.6) are far more protective of the general environment; even with the remoteness of the Nevada Test Site, introduction of potentially hazardous quantities of radioactive waste into the atmosphere must be scrupulously avoided. The Rocketdyne treatment concept features a diffuser to provide altitude simulation and pressure recovery, a series of heat exchangers to gradually cool the exhaust gas stream to 100 K, and an activated charcoal bed for adsorption of inert gases. A hydrogen-gas fed ejector provides auxiliary pumping for startup and shutdown of the engine. Supplemental filtration to remove particulates and condensed phases may be added at appropriate locations in the system. The clean hydrogen may be exhausted to the atmosphere and flared, or the gas may be condensed and stored for reuse in testing. The latter approach totally isolates the working gas from the environment.

Fast-Turnoff Transient Electromagnetic (TEM) Field Study at the Mars Analog Site of Rio Tinto, Spain

NASA Astrophysics Data System (ADS)

Jernsletten, J. A.

2005-03-01

This report describes a Fast-Turnoff Transient Electromagnetic (TEM) study at the Peña de Hierro ("Berg of Iron") field area of the Mars Analog Research and Technology Experiment (MARTE), near the towns Rio Tinto and Nerva, Andalucia region, Spain.

Research Technology

NASA Image and Video Library

1960-01-01

Originally investigated in the 1960's by Marshall Space Flight Center plarners as part of the Nuclear Energy for Rocket Vehicle Applications (NERVA) program, nuclear-thermal rocket propulsion has been more recently considered in spacecraft designs for interplanetary human exploration. This artist's concept illustrates a nuclear-thermal rocket with an aerobrake disk as it orbits Mars.

Mechanism-based Proteomic Screening Identifies Targets of Thioredoxin-like Proteins*

PubMed Central

Nakao, Lia S.; Everley, Robert A.; Marino, Stefano M.; Lo, Sze M.; de Souza, Luiz E.; Gygi, Steven P.; Gladyshev, Vadim N.

2015-01-01

Thioredoxin (Trx)-fold proteins are protagonists of numerous cellular pathways that are subject to thiol-based redox control. The best characterized regulator of thiols in proteins is Trx1 itself, which together with thioredoxin reductase 1 (TR1) and peroxiredoxins (Prxs) comprises a key redox regulatory system in mammalian cells. However, there are numerous other Trx-like proteins, whose functions and redox interactors are unknown. It is also unclear if the principles of Trx1-based redox control apply to these proteins. Here, we employed a proteomic strategy to four Trx-like proteins containing CXXC motifs, namely Trx1, Rdx12, Trx-like protein 1 (Txnl1) and nucleoredoxin 1 (Nrx1), whose cellular targets were trapped in vivo using mutant Trx-like proteins, under conditions of low endogenous expression of these proteins. Prxs were detected as key redox targets of Trx1, but this approach also supported the detection of TR1, which is the Trx1 reductant, as well as mitochondrial intermembrane proteins AIF and Mia40. In addition, glutathione peroxidase 4 was found to be a Rdx12 redox target. In contrast, no redox targets of Txnl1 and Nrx1 could be detected, suggesting that their CXXC motifs do not engage in mixed disulfides with cellular proteins. For some Trx-like proteins, the method allowed distinguishing redox and non-redox interactions. Parallel, comparative analyses of multiple thiol oxidoreductases revealed differences in the functions of their CXXC motifs, providing important insights into thiol-based redox control of cellular processes. PMID:25561728

Common radiation analysis model for 75,000 pound thrust NERVA engine (1137400E)

NASA Technical Reports Server (NTRS)

Warman, E. A.; Lindsey, B. A.

1972-01-01

The mathematical model and sources of radiation used for the radiation analysis and shielding activities in support of the design of the 1137400E version of the 75,000 lbs thrust NERVA engine are presented. The nuclear subsystem (NSS) and non-nuclear components are discussed. The geometrical model for the NSS is two dimensional as required for the DOT discrete ordinates computer code or for an azimuthally symetrical three dimensional Point Kernel or Monte Carlo code. The geometrical model for the non-nuclear components is three dimensional in the FASTER geometry format. This geometry routine is inherent in the ANSC versions of the QAD and GGG Point Kernal programs and the COHORT Monte Carlo program. Data are included pertaining to a pressure vessel surface radiation source data tape which has been used as the basis for starting ANSC analyses with the DASH code to bridge into the COHORT Monte Carlo code using the WANL supplied DOT angular flux leakage data. In addition to the model descriptions and sources of radiation, the methods of analyses are briefly described.

Investigation of a Tricarbide Grooved Ring Fuel Element for a Nuclear Thermal Rocket

NASA Technical Reports Server (NTRS)

Taylor, Brian D.; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2017-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the temperature limitations of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment.

Fine-Scale Linkage Mapping Reveals a Small Set of Candidate Genes Influencing Honey Bee Grooming Behavior in Response to Varroa Mites

PubMed Central

Arechavaleta-Velasco, Miguel E.; Alcala-Escamilla, Karla; Robles-Rios, Carlos; Tsuruda, Jennifer M.; Hunt, Greg J.

2012-01-01

Populations of honey bees in North America have been experiencing high annual colony mortality for 15–20 years. Many apicultural researchers believe that introduced parasites called Varroa mites (V. destructor) are the most important factor in colony deaths. One important resistance mechanism that limits mite population growth in colonies is the ability of some lines of honey bees to groom mites from their bodies. To search for genes influencing this trait, we used an Illumina Bead Station genotyping array to determine the genotypes of several hundred worker bees at over a thousand single-nucleotide polymorphisms in a family that was apparently segregating for alleles influencing this behavior. Linkage analyses provided a genetic map with 1,313 markers anchored to genome sequence. Genotypes were analyzed for association with grooming behavior, measured as the time that individual bees took to initiate grooming after mites were placed on their thoraces. Quantitative-trait-locus interval mapping identified a single chromosomal region that was significant at the chromosome-wide level (p<0.05) on chromosome 5 with a LOD score of 2.72. The 95% confidence interval for quantitative trait locus location contained only 27 genes (honey bee official gene annotation set 2) including Atlastin, Ataxin and Neurexin-1 (AmNrx1), which have potential neurodevelopmental and behavioral effects. Atlastin and Ataxin homologs are associated with neurological diseases in humans. AmNrx1 codes for a presynaptic protein with many alternatively spliced isoforms. Neurexin-1 influences the growth, maintenance and maturation of synapses in the brain, as well as the type of receptors most prominent within synapses. Neurexin-1 has also been associated with autism spectrum disorder and schizophrenia in humans, and self-grooming behavior in mice. PMID:23133594

Fine-scale linkage mapping reveals a small set of candidate genes influencing honey bee grooming behavior in response to Varroa mites.

PubMed

Arechavaleta-Velasco, Miguel E; Alcala-Escamilla, Karla; Robles-Rios, Carlos; Tsuruda, Jennifer M; Hunt, Greg J

2012-01-01

Populations of honey bees in North America have been experiencing high annual colony mortality for 15-20 years. Many apicultural researchers believe that introduced parasites called Varroa mites (V. destructor) are the most important factor in colony deaths. One important resistance mechanism that limits mite population growth in colonies is the ability of some lines of honey bees to groom mites from their bodies. To search for genes influencing this trait, we used an Illumina Bead Station genotyping array to determine the genotypes of several hundred worker bees at over a thousand single-nucleotide polymorphisms in a family that was apparently segregating for alleles influencing this behavior. Linkage analyses provided a genetic map with 1,313 markers anchored to genome sequence. Genotypes were analyzed for association with grooming behavior, measured as the time that individual bees took to initiate grooming after mites were placed on their thoraces. Quantitative-trait-locus interval mapping identified a single chromosomal region that was significant at the chromosome-wide level (p<0.05) on chromosome 5 with a LOD score of 2.72. The 95% confidence interval for quantitative trait locus location contained only 27 genes (honey bee official gene annotation set 2) including Atlastin, Ataxin and Neurexin-1 (AmNrx1), which have potential neurodevelopmental and behavioral effects. Atlastin and Ataxin homologs are associated with neurological diseases in humans. AmNrx1 codes for a presynaptic protein with many alternatively spliced isoforms. Neurexin-1 influences the growth, maintenance and maturation of synapses in the brain, as well as the type of receptors most prominent within synapses. Neurexin-1 has also been associated with autism spectrum disorder and schizophrenia in humans, and self-grooming behavior in mice.

"Bimodal" Nuclear Thermal Rocket (BNTR) Propulsion for Future Human Mars Exploration Missions

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.

2004-01-01

The Nuclear Thermal Rocket (NTR) Propulsion program is discussed. The Rover/NERVA program from 1959-1972 is compared with the current program. A key technology description, bimodal vehicle design for Mars Cargo and the crew transfer vehicle with inflatable module and artificial gravity capability, including diagrams are included. The LOX-Augmented NTR concept/operational features and characteristics are discussed.

Induction graphitizing furnace acceptance test report

NASA Technical Reports Server (NTRS)

1972-01-01

The induction furnace was designed to provide the controlled temperature and environment required for the post-cure, carbonization and graphitization processes for the fabrication of a fibrous graphite NERVA nozzle extension. The acceptance testing required six tests and a total operating time of 298 hrs. Low temperature mode operations, 120 to 850 C, were completed in one test run. High temperature mode operations, 120 to 2750 C, were completed during five tests.

Library Staff operate a Microfilm Reader at the Lewis Research Center

NASA Image and Video Library

1961-04-21

Jean Neidengard and George Mandel operate a Kodak Recordak microfilm reader in the library at the National Aeronautics and Space Administration (NASA) Lewis Research Center. The library was located in the Administration Building until the mid-1960s. It was then moved to the Propulsion Systems Laboratory Office Building. In 2008 the library was moved once again, to the Research Analysis Center. At the time of this photograph, the Lewis library claimed to possess “One of the most complete aero-technical collections in the world.” It was doing a brisk business in the early 1960s. During 1960 alone the library acquired 19,000 new documents and provided 100,000 documents to customers. The library’s eleven-person staff provided reference services, archived technical reports, and supplied periodicals. The staff also included Sam Reiss, a full-time translator who could read 30 languages. He translated technical reports from all over the world for the Lewis research staff. Jean Neidengard oversaw the secret Atomic Energy Commission (AEC) documents in the collection. NASA was partnering with the AEC at the time on Nuclear Engine for Rocket Vehicle Application (NERVA) program. NASA Lewis was the agency’s lead center in the NERVA program. Neidengard’s husband Bill was the head mechanic in the Propulsion Systems Laboratory. George Mandel led the library staff from 1955 to 1968.

Comparison of mission design options for manned Mars missions

NASA Technical Reports Server (NTRS)

Babb, Gus R.; Stump, William R.

1986-01-01

A number of manned Mars mission types, propulsion systems, and operational techniques are compared. Conjunction and opposition class missions for cryogenic, hybrid (cryo/storable), and NERVA propulsion concepts are addressed. In addition, both Earth and Mars orbit aerobraking, direct entry of landers, hyperbolic rendezvous, and electric propulsion cases are examined. A common payload to Mars was used for all cases. The basic figure of merit used was weight in low Earth orbit (LEO) at mission initiation. This is roughly proportional to launch costs.

Nuclear propulsion technology development - A joint NASA/Department of Energy project

NASA Technical Reports Server (NTRS)

Clark, John S.

1992-01-01

NASA-Lewis has undertaken the conceptual development of spacecraft nuclear propulsion systems with DOE support, in order to establish the bases for Space Exploration Initiative lunar and Mars missions. This conceptual evolution project encompasses nuclear thermal propulsion (NTP) and nuclear electric propulsion (NEP) systems. A technology base exists for NTP in the NERVA program files; more fundamental development efforts are entailed in the case of NEP, but this option is noted to offer greater advantages in the long term.

Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

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

Trammell, Michael P; Jolly, Brian C; Miller, James Henry

ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

Revised Point of Departure Design Options for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Fittje, James E.; Schnitzler, Bruce G.; Borowoski, Stanley

2015-01-01

Four Revised Point of Departure NTR Engines were Designed and Analyzed using MCNP and NESS. All Four Engines Have Thermodynamically Closed Cycles at Nominal Chamber Pressures. 111 kilonewton (25 kip-force) Cermet Design Required Dedicated Heater Elements to Close the Cycle. Cermet Based Designs had Slightly Higher TW Ratios, but Required Substantially More U-235. NERVA Derived Criticality Limited Engine Could Operate at Lower Power and Thrust Levels Compared to the Criticality Limited Cermet Design.

NERVA turbopump bearing retainer fabrication on nonmetallic retainer

NASA Technical Reports Server (NTRS)

Accinelli, J. B.

1972-01-01

The need for a low-wear, lightweight, high strength bearing retainer material with a radiation degradation threshold of 10 to the 9th power rads (C) prompted development of nonmetallic reinforced polymers of the following types: (1) polybenzimidazole, (2) polyimide, and (3) polyquinoxaline. Retainers were machined from tubular laminates (billets), including reinforcement by either glass or graphite fabric or filament. Fabrication of billets involves hot preimpregnation of the reinforcement fabric or filament with polymer followed by wrapping this prepreg over a heated mandrel to form a tube with the required thickness and length.

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are discussed. The authors demonstrated success in reaching desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and define a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

2018-01-01

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nu- clear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and de ne a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Payload dose rate from direct beam radiation and exhaust gas fission products. [for nuclear engine for rocket vehicles

NASA Technical Reports Server (NTRS)

Capo, M. A.; Mickle, R.

1975-01-01

A study was made to determine the dose rate at the payload position in the NERVA System (1) due to direct beam radiation and (2) due to the possible effect of fission products contained in the exhaust gases for various amounts of hydrogen propellant in the tank. Results indicate that the gamma radiation is more significant than the neutron flux. Under different assumptions the gamma contribution from the exhaust gases was 10 to 25 percent of total gamma flux.

A Historical Review of Cermet Fuel Development and the Engine Performance Implications

NASA Technical Reports Server (NTRS)

Stewart, Mark E.

2015-01-01

To better understand Cermet engine performance, examined historical material development reports two issues: High vaporization rate of UO2, High temperature chemical stability of UO2. Cladding and chemical stabilizers each result in large, order of magnitude improvements in high temperature performance. Few samples were tested above 2770 K. Results above 2770 K are ambiguous. Contemporary testing may clarify performance. Cermet sample testing during the NERVA Rover era. Important properties, melting temperature, vaporization rate, strength, Brittle-to-Ductile Transition, cermet sample test results, engine performance, location, peak temperature.

Nuclear Thermal Propulsion (NTP): A Proven Growth Technology for Human NEO/Mars Exploration Missions

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.; McCurdy, David R.; Packard, Thomas W.

2012-01-01

The nuclear thermal rocket (NTR) represents the next "evolutionary step" in high performance rocket propulsion. Unlike conventional chemical rockets that produce their energy through combustion, the NTR derives its energy from fission of Uranium-235 atoms contained within fuel elements that comprise the engine s reactor core. Using an "expander" cycle for turbopump drive power, hydrogen propellant is raised to a high pressure and pumped through coolant channels in the fuel elements where it is superheated then expanded out a supersonic nozzle to generate high thrust. By using hydrogen for both the reactor coolant and propellant, the NTR can achieve specific impulse (Isp) values of 900 seconds (s) or more - twice that of today s best chemical rockets. From 1955 - 1972, twenty rocket reactors were designed, built and ground tested in the Rover and NERVA (Nuclear Engine for Rocket Vehicle Applications) programs. These programs demonstrated: (1) high temperature carbide-based nuclear fuels; (2) a wide range of thrust levels; (3) sustained engine operation; (4) accumulated lifetime at full power; and (5) restart capability - all the requirements needed for a human Mars mission. Ceramic metal "cermet" fuel was pursued as well, as a backup option. The NTR also has significant "evolution and growth" capability. Configured as a "bimodal" system, it can generate its own electrical power to support spacecraft operational needs. Adding an oxygen "afterburner" nozzle introduces a variable thrust and Isp capability and allows bipropellant operation. In NASA s recent Mars Design Reference Architecture (DRA) 5.0 study, the NTR was selected as the preferred propulsion option because of its proven technology, higher performance, lower launch mass, versatile vehicle design, simple assembly, and growth potential. In contrast to other advanced propulsion options, no large technology scale-ups are required for NTP either. In fact, the smallest engine tested during the Rover program - the 25,000 lbf (25 klbf) "Pewee" engine is sufficient when used in a clustered engine arrangement. The "Copernicus" crewed spacecraft design developed in DRA 5.0 has significant capability and a human exploration strategy is outlined here that uses Copernicus and its key components for precursor near Earth object (NEO) and Mars orbital missions prior to a Mars landing mission. The paper also discusses NASA s current activities and future plans for NTP development that include system-level Technology Demonstrations - specifically ground testing a small, scalable NTR by 2020, with a flight test shortly thereafter.

9 CFR 78.7 - Brucellosis reactor cattle.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 9 Animals and Animal Products 1 2011-01-01 2011-01-01 false Brucellosis reactor cattle. 78.7... Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only for immediate slaughter as follows: (1...

9 CFR 78.7 - Brucellosis reactor cattle.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Brucellosis reactor cattle. 78.7... Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only for immediate slaughter as follows: (1...

Monopole Element at the Center of a Circular Groundplane of Arbitrary Radius. Volume 2. Appendices

DTIC Science & Technology

1986-03-31

ocM’T-o — o-iMin-o M*riiM/)«rNn« fMD -OO^O^MD<iino*nJ-JJO’*niMI)IDo(HDi>«.flriiot>o^rijNiD<r«i’-H zoo — JJSMI- rM»n-oomr)»o — m — — N — CM — i>ini>nj...o-o-orj-iNrxOO-o-oa) fta )CMOOft•o->rv«0𔃺-inn(0>fODnwoinin-.^)Oftino-ininj)o--o K<rry-<^ODoa)**mrtN«^naiooor)*(D^,NO’^inr>«^orvor)-H’OrvinN...C SUBROUTINE QMM * SUBROUTINE QMM<AK,DKD,DKW,CDKD,SDKD,CDK,SDK,TK. IWZ,NPH. Zll> IMPLICIT REAL*8 <A-H>,(P-Z) C0MPLEX*16 FDM, FMD

Assessment of the advantages and feasibility of a nuclear rocket for a manned Mars mission

NASA Technical Reports Server (NTRS)

Howe, Steven D.

1986-01-01

The feasibility of rebuilding and testing a nuclear thermal rocket (NTR) for the Mars mission was investigted. Calculations indicate that an NTR would substantially reduce the Earth-orbit assemble mass compared to LOX/LH2 systems. The mass savings were 36 and 65% for the cases of total aerobraking and of total propulsive braking respectively. Consequently, the cost savings for a single mission of using an NTR, if aerobraking is feasible, are probably insufficient to warrant the NTR development. If multiple missions are planned or if propulsive braking is desired at Mars and/or at Earth, then the savings of about $7 billion will easily pay for the NTR. Estimates of the cost of rebuilding a NTR were based on the previous NERVA program's budget plus additional costs to develop a flight ready engine. The total cost to build the engine would be between $4 to 5 billion. The concept of developing a full-power test stand at Johnston Atoll in the Pacific appears very feasible. The added expense of building facilities on the island should be less than $1.4 billion.

Assessment of the advantages and feasibility of a nuclear rocket for a manned Mars mission

NASA Astrophysics Data System (ADS)

Howe, Steven D.

1986-05-01

The feasibility of rebuilding and testing a nuclear thermal rocket (NTR) for the Mars mission was investigted. Calculations indicate that an NTR would substantially reduce the Earth-orbit assemble mass compared to LOX/LH2 systems. The mass savings were 36 and 65% for the cases of total aerobraking and of total propulsive braking respectively. Consequently, the cost savings for a single mission of using an NTR, if aerobraking is feasible, are probably insufficient to warrant the NTR development. If multiple missions are planned or if propulsive braking is desired at Mars and/or at Earth, then the savings of about $7 billion will easily pay for the NTR. Estimates of the cost of rebuilding a NTR were based on the previous NERVA program's budget plus additional costs to develop a flight ready engine. The total cost to build the engine would be between $4 to 5 billion. The concept of developing a full-power test stand at Johnston Atoll in the Pacific appears very feasible. The added expense of building facilities on the island should be less than $1.4 billion.

Design and development of LH2 cooled rolling element radial bearings for the NERVA engine turbopump. Volume 3: Phase 2: Tests on build-ups 16, 17, and 18 at NRDS, Jackass Flats, Nevada, December 1971 - March 1972

NASA Technical Reports Server (NTRS)

Accinelli, J. B.; Koch, D. A.; Reuter, F.

1972-01-01

The use of liquid hydrogen to cool the rolling element radial bearings in the nuclear engine for rocket vehicles is discussed. The fifteen hour service life goal was obtained during the tests. The increase in bearing life was also considered to be produced by: (1) improvements in bearing material, (2) bearing retainer configuration and manufacturing changes, and (3) better control of operating parameters.

Identification of Hydrated Sulfates Collected in the Northern Rio Tinto Valley by Reflectance and Raman Spectroscopy

NASA Technical Reports Server (NTRS)

Chemtob, S. M.; Arvidson, R. E.; Fernandez-Remolar, D. C.; Amils, R.; Morris, R. V.; Ming, D. W.; Prieto-Ballesteros, O.; Mustard, J. F.; Hutchinson, L.; Stein, T. C.;

9 CFR 78.7 - Brucellosis reactor cattle.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 9 Animals and Animal Products 1 2014-01-01 2014-01-01 false Brucellosis reactor cattle. 78.7... AGRICULTURE INTERSTATE TRANSPORTATION OF ANIMALS (INCLUDING POULTRY) AND ANIMAL PRODUCTS BRUCELLOSIS Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a...

9 CFR 78.7 - Brucellosis reactor cattle.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 9 Animals and Animal Products 1 2013-01-01 2013-01-01 false Brucellosis reactor cattle. 78.7... AGRICULTURE INTERSTATE TRANSPORTATION OF ANIMALS (INCLUDING POULTRY) AND ANIMAL PRODUCTS BRUCELLOSIS Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a...

9 CFR 78.7 - Brucellosis reactor cattle.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 9 Animals and Animal Products 1 2012-01-01 2012-01-01 false Brucellosis reactor cattle. 78.7... AGRICULTURE INTERSTATE TRANSPORTATION OF ANIMALS (INCLUDING POULTRY) AND ANIMAL PRODUCTS BRUCELLOSIS Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a...

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

NASA Astrophysics Data System (ADS)

Shen, Yongjun; Lei, Lecheng; Zhang, Xingwang; Ding, Jiandong

2014-11-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10-9 mol/L and 0.61 × 10-9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10-2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10-2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation pathways were elucidated in light of the analyzed degradation products.

Redox sensitivity of the MyD88 immune signaling adapter.

PubMed

Stottmeier, Benjamin; Dick, Tobias P

2016-12-01

The transcription factor nuclear factor-κB (NF-κB) mediates expression of key genes involved in innate immunity and inflammation. NF-κB activation has been repeatedly reported to be modulated by hydrogen peroxide (H 2 O 2 ). Here, we show that the NF-κB-activating signaling adapter myeloid differentiation primary response gene 88 (MyD88) is highly sensitive to oxidation by H 2 O 2 and may be redox-regulated in its function, thus facilitating an influence of H 2 O 2 on the NF-κB signaling pathway. Upon oxidation, MyD88 forms distinct disulfide-linked conjugates which are reduced by the MyD88-interacting oxidoreductase nucleoredoxin (Nrx). MyD88 cysteine residues functionally modulate MyD88-dependent NF-κB activation, suggesting a link between MyD88 thiol oxidation state and immune signaling. Copyright © 2016 Elsevier Inc. All rights reserved.

Affordable Development and Qualification Strategy for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Gerrish, Harold P., Jr.; Doughty, Glen E.; Bhattacharyya, Samit K.

2013-01-01

Nuclear Thermal Propulsion (NTP) is a concept which uses a nuclear reactor to heat a propellant to high temperatures without combustion and can achieve significantly greater specific impulse than chemical engines. NTP has been considered many times for human and cargo missions beyond low earth orbit. A lot of development and technical maturation of NTP components took place during the Rover/NERVA program of the 60's and early 70's. Other NTP programs and studies followed attempting to further mature the NTP concept and identify a champion customer willing to devote the funds and support the development schedule to a demonstration mission. Budgetary constraints require the use of an affordable development and qualification strategy that takes into account all the previous work performed on NTP to construct an existing database, and include lessons learned and past guidelines followed. Current guidelines and standards NASA uses for human rating chemical rocket engines is referenced. The long lead items for NTP development involve the fuel elements of the reactor and ground testing the engine system, subsystem, and components. Other considerations which greatly impact the development plans includes the National Space Policy, National Environmental Policy Act, Presidential Directive/National Security Council Memorandum #25 (Scientific or Technological Experiments with Possible Large-Scale Adverse Environmental Effects and Launch of Nuclear Systems into Space), and Safeguards and Security. Ground testing will utilize non-nuclear test capabilities to help down select components and subsystems before testing in a nuclear environment to save time and cost. Existing test facilities with minor modifications will be considered to the maximum extent practical. New facilities will be designed to meet minimum requirements. Engine and test facility requirements are based on the driving mission requirements with added factors of safety for better assurance and reliability. Emphasis will be placed on small engines, since the smaller the NTP engine, the easier it is to transport, assemble/disassemble, and filter the exhaust during tests. A new ground test concept using underground bore holes (modeled after the underground nuclear test program) to filter the NTP engine exhaust is being considered. The NTP engine system design, development, test, and evaluation plan includes many engine components and subsystems, which are very similar to those used in chemical engines, and can be developed in conjunction with them Other less mature NTP engine components and subsystems (e.g., reactor) will be thoroughly analyzed and tested to acceptable levels recommended by the referenced standards and guidelines. The affordable development strategy also considers a prototype flight test, as a final step in the development process. Preliminary development schedule estimates show that an aggressive development schedule (without much margin) will be required to be flight ready for a 2033 human mission to Mars.

NERVA nozzle design status report

NASA Technical Reports Server (NTRS)

Williams, J. J.; Pickering, J. L.; Ackerman, R. G.

1972-01-01

The results of the design analyses are presented along with the status of the attained design maturity of the structural elements of the nozzle jacket and various aspects of the coolant passages. The design analyses relating to the nozzle shell were based on design allowables as supported by cursory values obtained from ARMCO 22-13-5 nozzle forgings. The major aspects of the coolant passages considered include: low cycle thermal fatigue, ability to operate at 4500 R gas temperature, tube buckling, and susceptibility to erosion. The scope of the analysis is limited to processes leading to reliability assessments of failure mechanisms.

Evaluation of insulation materials and composites for use in a nuclear radiation environment, phase 2

NASA Technical Reports Server (NTRS)

Westerheide, D. E.; Carter, H. G.; Erickson, R. C.; Kerlin, E. E.

1972-01-01

The nuclear heating of the propellant in all of the four baseline RNS configurations studied was much lower than that of the nuclear flight module configuration with the 5000-MW NERVA analyzed previously. Although the nuclear heating has been reduced, the effect of nuclear heating on the propellant as well as the effect of nuclear heating on internal structures such as antivortex baffles, screens, and sump components cannot be neglected. In addition, it was found that the present analytical precedures were not able to predict boundary layer initiation and breakoff points with the accuracy necessary to predict propellant thermodynamic nonequilibrium (stratification) and/or mixing.

Culture-independent detection of 'TM7' bacteria in a streptomycin-resistant acidophilic nitrifying process

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

Kurogi, T.; Linh, N. T. T.; Kuroki, T.

2014-02-20

Nitrification in biological wastewater treatment processes has been believed for long time to take place under neutral conditions and is inhibited under acidic conditions. However, we previously constructed acidophilic nitrifying sequencing-batch reactors (ANSBRs) being capable of nitrification at < pH 4 and harboring bacteria of the candidate phylum 'TM7' as the major constituents of the microbial community. In light of the fact that the 16S rRNA of TM7 bacteria has a highly atypical base substitution possibly responsible for resistance to streptomycin at the ribosome level, this study was undertaken to construct streptomycin-resistant acidophilic nitrifying (SRAN) reactors and to demonstrate whethermore » TM7 bacteria are abundant in these reactors. The SRAN reactors were constructed by seeding with nitrifying sludge from an ANSBR and cultivating with ammonium-containing mineral medium (pH 4.0), to which streptomycin at a concentration of 10, 30 and 50 mg L{sup −1} was added. In all reactors, the pH varied between 2.7 and 4.0, and ammonium was completely converted to nitrate in every batch cycle. PCR-aided denaturing gradient gel electrophoresis (DGGE) targeting 16S rRNA genes revealed that some major clones assigned to TM7 bacteria and Gammaproteobacteria were constantly present during the overall period of operation. Fluorescence in situ hybridization (FISH) with specific oligonucleotide probes also showed that TM7 bacteria predominated in all SRAN reactors, accounting for 58% of the total bacterial population on average. Although the biological significance of the TM7 bacteria in the SRAN reactors are unknown, our results suggest that these bacteria are possibly streptomycin-resistant and play some important roles in the acidophilic nitrifying process.« less

RELAP-7 Software Verification and Validation Plan

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

Smith, Curtis L.; Choi, Yong-Joon; Zou, Ling

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty yearsmore » of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.« less

Current Ground Test Options for Nuclear Thermal Propulsion (NTP)

NASA Technical Reports Server (NTRS)

Gerrish, Harold P., Jr.

2014-01-01

About 20 different NTP engines/ reactors were tested from 1959 to 1972 as part of the Rover and Nuclear Engine for Rocket Vehicle Application (NERVA) program. Most were tested in open air at test cell A or test cell C, at the Nevada Test Site (NTS). Even after serious engine breakdowns of the reactor (e.g., Phoebus 1A), the test cells were cleaned up for other engine tests. The engine test stand (ETS) was made for high altitude (approximately 1 psia) testing of an NTP engine with a flight configuration, but still had the exhaust released to open air. The Rover/NERVA program became aware of new environmental regulations which would prohibit the release of any significant quantity of radioactive particulates and noble gases into the open air. The nuclear furnace (NF-1) was the last reactor tested before the program was cancelled in 1973, but successfully demonstrated a scrubber concept on how to filter the NTP exhaust. The NF-1 was demonstrated in the summer of 1972. The NF-1 used a 44MW reactor and operated each run for approximately 90 minutes. The system cooled the hot hydrogen exhaust from the engine with a water spray before entering a particle filter. The exhaust then passed through a series of heat exchangers and water separators to help remove water from the exhaust and further reduce the exhaust temperatures. The exhaust was next prepared for the charcoal trap by passing through a dryer and effluent cooler to bring exhaust temperatures close to liquid nitrogen. At those low temperatures, most of the noble gases (e.g., Xe and Kr made from fission products) get captured in the charcoal trap. The filtered hydrogen is finally passed through a flare stack and released to the air. The concept was overall successful but did show a La plating on some surfaces and had multiple recommendations for improvement. The most recent detailed study on the NTP scrubber concept was performed by the ARES Corporation in 2006. The concept is based on a 50,000 lbf thrust engine (approximately 1 GW) with a maximum burn time of 1 hour. The concept utilized lessons learned from NF-1. The strategy breaks down the exhaust into parallel paths to allow flexibility with engine size and mass flow of exhaust. Similar to NF-1, the exhaust is slowed down, cooled, filtered of particulates, filtered of noble gases, and then the clean hydrogen is flared to open air. Another concept proposed by Steve Howe (currently Director of the Center for Space Nuclear Research) to simplify the NTP exhaust filtering is to run the hydrogen exhaust into boreholes underground to filter the exhaust. The two borehole site locations proposed are at the NTS and at the Idaho National Laboratory (INL). At NTS, the boreholes are 8' diameter and 1200' deep. The permeability of hydrogen through the soil and its buoyancy will allow it to rise up through the soil and allow the filtering of noble gases and radioactive particulates. The exhaust needs to be cooled to 600C before entering the borehole to avoid soil glazing. Preliminary analysis shows a small buildup of back pressure with time which depends on permeability. Noble gases entering the borehole walls deep can take a long time before reaching the surface. Other factors affecting permeability include borehole pressure, water saturation, and turbulence. Also, a possible need to pump out contaminated water collected at the bottom of the borehole. At INL, the borehole concept is slightly different. The underground borehole has openings to the soil at special depths which have impermeable interbeds above the water table and below the surface to allow the exhaust to travel horizontal between the impermeable layers. Preliminary results indicate better permeability than at NTS. The last option is total containment of the exhaust during the test run. The concept involves slowing down the flow to subsonic in a water cooled diffuser. The hydrogen is burned off in an oxygen rich afterburner with the only products being steam, oxygen, and some noble gases. A heat exchanger and water spray pulls heat from the steam and lowers the temperature for condensation. The optimum ratio between the two is being investigated, with a goal to minimize the total volume of the water hold tanks. A water tank farm collects the contaminated water. The amount of water produced from burning the hydrogen is approximately 100,000 gallons (not including cooling water) for a 25k lbf engine operating for 50 minutes. Residual gases (e.g., oxygen and some noble gases) can be captured at cryogenic levels with a liquid nitrogen cooled dewar. After a few weeks post-test, the radiation levels can drop to more favorable levels before slowly draining each capture tank and using existing filters. With today's environmental regulations, the NTP exhaust is filtered to meet 10 mrem/year exposure to the general public (at a DOE site) or 100 mrem/year (via NRC when tested elsewhere), when natural background radiation exposure to the general public is 300- 600 mrem per year. The current society feels more comfortable with filtering even lower to as low as reasonably achievable (ALARA).

Quick trips to Mars

NASA Technical Reports Server (NTRS)

Hornung, R.

1991-01-01

The design of a Mars Mission Vehicle that would have to be launched by two very heavy lift launch vehicles is described along with plans for a mission to Mars. The vehicle has three nuclear engine for rocket vehicle application (NERVA) boosters with a fourth in the center that acts as a dual mode system. The fourth generates electrical power while in route, but it also helps lift the vehicle out of earth orbit. A Mars Ascent Vehicle (MAV), a Mars transfer vehicle stage, and a Mars Excursion Vehicle (MEV) are located on the front end of this vehicle. Other aspects of this research including aerobraking, heat shielding, nuclear thermal rocket engines, a mars mission summary, closed Brayton cycle with and without regeneration, liquid hydrogen propellant storage, etc. are addressed.

Mars lander survey

NASA Technical Reports Server (NTRS)

Stump, William R.; Babb, Gus R.; Davis, Hubert P.

1986-01-01

The requirements, issues, and design options are reviewed for manned Mars landers. Issues such as high 1/d versus low 1/d shape, parking orbit, and use of a small Mars orbit transfer vehicle to move the lander from orbit to orbit are addressed. Plots of lander mass as a function of Isp, destination orbit, and cargo up and down, plots of initial stack mass in low Earth orbit as a function of lander mass and parking orbit, detailed weight statements, and delta V tables for a variety of options are included. Lander options include a range from minimum landers up to a single stage reusable design. Mission options include conjunction and Venus flyby trajectories using all-cryogenic, hybrid, NERVA, and Mars orbit aerobraking propulsion concepts.

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

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

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

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

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

Block, R.C.; Feiner, F.

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

77 FR 41206 - Guidelines for Preparing and Reviewing Licensing Applications for Instrumentation and Control...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-12

... Applications for Instrumentation and Control Upgrades for Non-Power Reactors AGENCY: Nuclear Regulatory... (NRC or the Commission) is requesting public comment on Chapter 7, Section 7.3, Reactor Control System...-Power Reactors: Format and Content,'' for instrumentation and control (I&C) upgrades and NUREG-1537...

Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

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

Bruce G. Schnitzler; Stanley K. Borowski

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified asmore » the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were constrained to fit within the payload volume of the then planned space shuttle. The SNRE core design utilized hexagonal fuel elements and hexagonal structural support elements. The total number of elements can be varied to achieve engine designs of higher or lower thrust levels. Some variation in the ratio of fuel elements to structural elements is also possible. Options for SNRE-based engine designs in the 25,000-lbf thrust range were described in a recent (2010) Joint Propulsion Conference paper. The reported designs met or exceeded the performance characteristics baselined in the DRA 5.0 Study. Lower thrust SNRE-based designs were also described in a recent (2011) Joint Propulsion Conference paper. Recent activities have included parallel evaluation and design efforts on fast spectrum engines employing refractory metal alloy fuels. These efforts include evaluation of both heritage designs from the Argonne National Laboratory (ANL) and General Electric Company GE-710 Programs as well as more recent designs. Results are presented for a number of not-yet optimized fast spectrum engine options.« less

Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Schnitzler, Bruce G.; Borowski, Stanley K.

2012-01-01

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were constrained to fit within the payload volume of the then planned space shuttle. The SNRE core design utilized hexagonal fuel elements and hexagonal structural support elements. The total number of elements can be varied to achieve engine designs of higher or lower thrust levels. Some variation in the ratio of fuel elements to structural elements is also possible. Options for SNRE-based engine designs in the 25,000-lbf thrust range were described in a recent (2010) Joint Propulsion Conference paper. The reported designs met or exceeded the performance characteristics baselined in the DRA 5.0 Study. Lower thrust SNRE-based designs were also described in a recent (2011) Joint Propulsion Conference paper. Recent activities have included parallel evaluation and design efforts on fast spectrum engines employing refractory metal alloy fuels. These efforts include evaluation of both heritage designs from the Argonne National Laboratory (ANL) and General Electric Company GE-710 Programs as well as more recent designs. Results are presented for a number of not-yet optimized fast spectrum engine options.

75 FR 34219 - Revision of Fee Schedules; Fee Recovery for FY 2010

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-16

....8 $6.3 $7.5 Spent Fuel Storage/Reactor Decommissioning..... -- -- 2.7 0.2 0.2 Test and Research... 2009 fee is also shown for comparative purposes. Table V--Rebaselined Annual Fees FY2009 Annual FY 2010... Decommissioning Test and Research Reactors (Non-power 87,600 81,700 Reactors) High Enriched Uranium Fuel Facility...

RELAP-7 Development Updates

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

Zhang, Hongbin; Zhao, Haihua; Gleicher, Frederick Nathan

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is beingmore » developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis. The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.« less

Biological oxidation of hydrogen sulfide in mineral media using a biofilm airlift suspension reactor.

PubMed

Moghanloo, G M Mojarrad; Fatehifar, E; Saedy, S; Aghaeifar, Z; Abbasnezhad, H

2010-11-01

Hydrogen sulfide (H(2)S) removal in mineral media using Thiobacillus thioparus TK-1 in a biofilm airlift suspension reactor (BAS) was investigated to evaluate the relationship between biofilm formation and changes in inlet loading rates. Aqueous sodium sulfide was fed as the substrate into the continuous BAS-reactor. The reactor was operated at a constant temperature of 30 degrees C and a pH of 7, the optimal temperature and pH for biomass growth. The startup of the reactor was performed with basalt carrier material. Optimal treatment performance was obtained at a loading rate of 4.8 mol S(2-) m(-3) h(-1) at a conversion efficiency as high as 100%. The main product of H(2)S oxidation in the BAS-reactor was sulfate because of high oxygen concentrations in the airlift reactor. The maximum sulfide oxidation rate was 6.7 mol S(2-) m(-3) h(-1) at a hydraulic residence time of 3.3 h in the mineral medium. The data showed that the BAS-reactor with this microorganism can be used for sulfide removal from industrial effluent. Copyright 2010 Elsevier Ltd. All rights reserved.

Nuclear Thermal Rocket (Ntr) Propulsion: A Proven Game-Changing Technology for Future Human Exploration Missions

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.; McCurdy, David R.; Packard, Thomas W.

2012-01-01

The NTR represents the next evolutionary step in high performance rocket propulsion. It generates high thrust and has a specific impulse (Isp) of approx.900 seconds (s) or more V twice that of today s best chemical rockets. The technology is also proven. During the previous Rover and NERVA (Nuclear Engine for Rocket Vehicle Applications) nuclear rocket programs, 20 rocket reactors were designed, built and ground tested. These tests demonstrated: (1) a wide range of thrust; (2) high temperature carbide-based nuclear fuel; (3) sustained engine operation; (4) accumulated lifetime; and (5) restart capability V all the requirements needed for a human mission to Mars. Ceramic metal cermet fuel was also pursued, as a backup option. The NTR also has significant growth and evolution potential. Configured as a bimodal system, it can generate electrical power for the spacecraft. Adding an oxygen afterburner nozzle introduces a variable thrust and Isp capability and allows bipropellant operation. In NASA s recent Mars Design Reference Architecture (DRA) 5.0 study, the NTR was selected as the preferred propulsion option because of its proven technology, higher performance, lower launch mass, simple assembly and mission operations. In contrast to other advanced propulsion options, NTP requires no large technology scale-ups. In fact, the smallest engine tested during the Rover program V the 25,000 lbf (25 klbf) Pewee engine is sufficient for human Mars missions when used in a clustered engine arrangement. The Copernicus crewed spacecraft design developed in DRA 5.0 has significant capability and a human exploration strategy is outlined here that uses Copernicus and its key components for precursor near Earth asteroid (NEA) and Mars orbital missions prior to a Mars landing mission. Initially, the basic Copernicus vehicle can enable reusable 1-year round trip human missions to candidate NEAs like 1991 JW and Apophis in the late 2020 s to check out vehicle systems. Afterwards, the Copernicus spacecraft and its 2 key components, now configured as an Earth Return Vehicle / propellant tanker, would be used for a short round trip (approx.18 - 20 months)/short orbital stay (60 days) Mars / Phobos survey mission in 2033 using a split mission approach. The paper also discusses NASA s current Foundational Technology Development activities and its pre-decisional plans for future system-level Technology Demonstrations that include ground testing a small (approx.7.5 klbf) scalable NTR before the decade is out with a flight test shortly thereafter.

The Nuclear Thermal Propulsion Stage (NTPS): A Key Space Asset for Human Exploration and Commercial Missions to the Moon

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.; McCurdy, David R.; Burke, Laura M.

2014-01-01

The nuclear thermal rocket (NTR) has frequently been discussed as a key space asset that can bridge the gap between a sustained human presence on the Moon and the eventual human exploration of Mars. Recently, a human mission to a near Earth asteroid (NEA) has also been included as a "deep space precursor" to an orbital mission of Mars before a landing is attempted. In his "post-Apollo" Integrated Space Program Plan (1970 to 1990), Wernher von Braun, proposed a reusable Nuclear Thermal Propulsion Stage (NTPS) to deliver cargo and crew to the Moon to establish a lunar base initially before sending human missions to Mars. The NTR was selected because it was a proven technology capable of generating both high thrust and high specific impulse (Isp approx. 900 s)-twice that of today's best chemical rockets. During the Rover and NERVA programs, 20 rocket reactors were designed, built and successfully ground tested. These tests demonstrated the (1) thrust levels; (2) high fuel temperatures; (3) sustained operation; (4) accumulated lifetime; and (5) restart capability needed for an affordable in-space transportation system. In NASA's Mars Design Reference Architecture (DRA) 5.0 study, the "Copernicus" crewed NTR Mars transfer vehicle used three 25 klbf "Pewee" engines-the smallest and highest performing engine tested in the Rover program. Smaller lunar transfer vehicles-consisting of a NTPS with three approx. 16.7 klbf "SNRE-class" engines, an in-line propellant tank, plus the payload-can be delivered to LEO using a 70 t to LEO upgraded SLS, and can support reusable cargo delivery and crewed lunar landing missions. The NTPS can play an important role in returning humans to the Moon to stay by providing an affordable in-space transportation system that can allow initial lunar outposts to evolve into settlements capable of supporting commercial activities. Over the next decade collaborative efforts between NASA and private industry could open up new exploration and commercial opportunities for both organizations. With efficient NTP, commercial habitation and crew delivery systems, a "mobile cislunar research station" can transport crews to small NEAs delivered to the E-ML2 point. Also possible are week-long "lunar tourism" missions that can carry passengers into lunar orbit for sightseeing (and plenty of picture taking), then return them to Earth orbit where they would re-enter and land using a small reusable lifting body based on NASA's HL-20 design. Mission descriptions, key vehicle features and operational characteristics are described and presented.

Affordable Development and Demonstration of a Small NTR Engine and Stage: A Preliminary NASA, DOE, and Industry Assessment

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.; Sefcik, Robert J.; Fittje, James E.; McCurdy, David R.; Qualls, Arthur L.; Schnitzler, Bruce G.; Werner, James E.; Weitzberg, Abraham; Joyner, Claude R.

2015-01-01

The Nuclear Thermal Rocket (NTR) represents the next evolutionary step in cryogenic liquid rocket engines. Deriving its energy from fission of uranium-235 atoms contained within fuel elements that comprise the engine's reactor core, the NTR can generate high thrust at a specific impulse of approx. 900 seconds or more - twice that of today's best chemical rockets. In FY'11, as part of the AISP project, NASA proposed a Nuclear Thermal Propulsion (NTP) effort that envisioned two key activities - "Foundational Technology Development" followed by system-level "Technology Demonstrations". Five near-term NTP activities identified for Foundational Technology Development became the basis for the NCPS project started in FY'12 and funded by NASA's AES program. During Phase 1 (FY'12-14), the NCPS project was focused on (1) Recapturing fuel processing techniques and fabricating partial length "heritage" fuel elements for the two candidate fuel forms identified by NASA and the DOE - NERVA graphite "composite" and the uranium dioxide (UO2) in tungsten "cermet". The Phase 1 effort also included: (2) Engine Conceptual Design; (3) Mission Analysis and Requirements Definition; (4) Identification of Affordable Options for Ground Testing; and (5) Formulation of an Affordable and Sustainable NTP Development Strategy. During FY'14, a preliminary plan for DDT&E was outlined by GRC, the DOE and industry for NASA HQ that involved significant system-level demonstration projects that included GTD tests at the NNSS, followed by a FTD mission. To reduce development costs, the GTD and FTD tests use a small, low thrust (approx. 7.5 or 16.5 klbf) engine. Both engines use graphite composite fuel and a "common" fuel element design that is scalable to higher thrust (approx. 25 klbf) engines by increasing the number of elements in a larger diameter core that can produce greater thermal power output. To keep the FTD mission cost down, a simple "1-burn" lunar flyby mission was considered along with maximizing the use of existing and flight proven liquid rocket and stage hardware (e.g., from the RL10-B2 engine and Delta Cryogenic Second Stage) to further ensure affordability. This paper provides a preliminary NASA, DOE and industry assessment of what is required - the key DDT&E activities, development options, and the associated schedule - to affordably build, ground test and fly a small NTR engine and stage within a 10-year timeframe.

A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor.

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

Bretscher, M. M.

1998-10-14

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% {sup 235}U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm{sup 3} and 3.8 gU/cm{sup 3} are required tomore » match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.« less

Evaluation of Recent Upgrades to the NESS (Nuclear Engine System Simulation) Code

NASA Technical Reports Server (NTRS)

Fittje, James E.; Schnitzler, Bruce G.

2008-01-01

The Nuclear Thermal Rocket (NTR) concept is being evaluated as a potential propulsion technology for exploratory expeditions to the moon, Mars, and beyond. The need for exceptional propulsion system performance in these missions has been documented in numerous studies, and was the primary focus of a considerable effort undertaken during the Rover/NERVA program from 1955 to 1973. The NASA Glenn Research Center is leveraging this past NTR investment in their vehicle concepts and mission analysis studies with the aid of the Nuclear Engine System Simulation (NESS) code. This paper presents the additional capabilities and upgrades made to this code in order to perform higher fidelity NTR propulsion system analysis and design, and a comparison of its results to the Small Nuclear Rocket Engine (SNRE) design.

Fast quench reactor and method

DOEpatents

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

1998-05-12

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

A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2013-01-01

This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

The thermal decomposition of the benzyl radical in a heated micro-reactor. II. Pyrolysis of the tropyl radical

NASA Astrophysics Data System (ADS)

Buckingham, Grant T.; Porterfield, Jessica P.; Kostko, Oleg; Troy, Tyler P.; Ahmed, Musahid; Robichaud, David J.; Nimlos, Mark R.; Daily, John W.; Ellison, G. Barney

2016-07-01

Cycloheptatrienyl (tropyl) radical, C7H7, was cleanly produced in the gas-phase, entrained in He or Ne carrier gas, and subjected to a set of flash-pyrolysis micro-reactors. The pyrolysis products resulting from C7H7 were detected and identified by vacuum ultraviolet photoionization mass spectrometry. Complementary product identification was provided by infrared absorption spectroscopy. Pyrolysis pressures in the micro-reactor were roughly 200 Torr and residence times were approximately 100 μs. Thermal cracking of tropyl radical begins at 1100 K and the products from pyrolysis of C7H7 are only acetylene and cyclopentadienyl radicals. Tropyl radicals do not isomerize to benzyl radicals at reactor temperatures up to 1600 K. Heating samples of either cycloheptatriene or norbornadiene never produced tropyl (C7H7) radicals but rather only benzyl (C6H5CH2). The thermal decomposition of benzyl radicals has been reconsidered without participation of tropyl radicals. There are at least three distinct pathways for pyrolysis of benzyl radical: the Benson fragmentation, the methyl-phenyl radical, and the bridgehead norbornadienyl radical. These three pathways account for the majority of the products detected following pyrolysis of all of the isotopomers: C6H5CH2, C6H5CD2, C6D5CH2, and C6H513CH2. Analysis of the temperature dependence for the pyrolysis of the isotopic species (C6H5CD2, C6D5CH2, and C6H513CH2) suggests the Benson fragmentation and the norbornadienyl pathways open at reactor temperatures of 1300 K while the methyl-phenyl radical channel becomes active at slightly higher temperatures (1500 K).

The thermal decomposition of the benzyl radical in a heated micro-reactor. II. Pyrolysis of the tropyl radical

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

Buckingham, Grant T.; Porterfield, Jessica P.; Kostko, Oleg

2016-07-05

Cycloheptatrienyl (tropyl) radical, C7H7, was cleanly produced in the gas-phase, entrained in He or Ne carrier gas, and subjected to a set of flash-pyrolysis micro-reactors. The pyrolysis products resulting from C7H7 were detected and identified by vacuum ultraviolet photoionization mass spectrometry. Complementary product identification was provided by infrared absorption spectroscopy. Pyrolysis pressures in the micro-reactor were roughly 200 Torr and residence times were approximately 100 us. Thermal cracking of tropyl radical begins at 1100 K and the products from pyrolysis of C7H7 are only acetylene and cyclopentadienyl radicals. Tropyl radicals do not isomerize to benzyl radicals at reactor temperatures upmore » to 1600 K. Heating samples of either cycloheptatriene or norbornadiene never produced tropyl (C7H7) radicals but rather only benzyl (C6H5CH2). The thermal decomposition of benzyl radicals has been reconsidered without participation of tropyl radicals. There are at least three distinct pathways for pyrolysis of benzyl radical: the Benson fragmentation, the methyl-phenyl radical, and the bridgehead norbornadienyl radical. These three pathways account for the majority of the products detected following pyrolysis of all of the isotopomers: C6H5CH2, C6H5CD2, C6D5CH2, and C6H5 13CH2. Analysis of the temperature dependence for the pyrolysis of the isotopic species (C6H5CD2, C6D5CH2, and C6H5 13CH2) suggests the Benson fragmentation and the norbornadienyl pathways open at reactor temperatures of 1300 K while the methyl-phenyl radical channel becomes active at slightly higher temperatures (1500 K).« less

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

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

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers dealmore » with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.« less

Impact of radiation dose on nuclear shuttle configuration

NASA Technical Reports Server (NTRS)

Goetz, C. A.; Billings, M. A.

1972-01-01

The impact of nuclear radiation (from the NERVA propulsion system) on the selection of a reference configuration for each of two classes of the reusable nuclear shuttle is considered. One class was characterized by a single propellant tank, the shape of whose bottom was found to have a pronounced effect on crew radiation levels and associated shield weight requirements. A trade study of shield weight versus structural weight indicated that the minimum-weight configuration for this class had a tank bottom in the shape of a frustum of a 10 deg-half-angle cone. A hybrid version of this configuration was found to affect crew radiation levels in substantially the same manner. The other class of RNS consisted of a propulsion module and eight propellant modules. Radiation analyses of various module arrangements led to a design configuration with no external shield requirements.

Characterization of a joint recirculation of concentrated leachate and leachate to landfills with a microaerobic bioreactor for leachate treatment.

PubMed

He, Ruo; Wei, Xiao-Meng; Tian, Bao-Hu; Su, Yao; Lu, Yu-Lan

2015-12-01

With comparison of a traditional landfill, a joint recirculation of concentrated leachate and leachate to landfills with or without a microaerobic reactor for leachate treatment was investigated in this study. The results showed that the joint recirculation of concentrated leachate and leachate with a microaerobic reactor for leachate treatment could not only utilize the microaerobic reactor to buffer the fluctuation of quality and quantity of leachate during landfill stabilization, but also reduce the inhibitory effect of acidic pH and high concentrations of ammonium in recycled liquid on microorganisms and accelerate the degradation of landfilled waste. After 390 days of operation, the discharge of COD and total nitrogen (TN) from the landfill with leachate pretreatment by a microaerobic reactor was 7.4 and 0.9 g, respectively, which accounted for 0.7% and 2.6% of COD, 1.9% and 7.5% of the TN discharge from the landfills without recirculation and directly recirculated with leachate and concentrated leachate, respectively. The degradation of the organic matter and biodegradable matter (BDM) in the landfill reactors could fit well with the first-order kinetics. The highest degradation of the organic matter and BDM was observed in the joint recirculation system with a microaerobic reactor for leachate treatment with the degradation constant of the first-order kinetics of 0.001 and 0.002. Copyright © 2015 Elsevier Ltd. All rights reserved.

Comparison of two modified coal ash ferric-carbon micro-electrolysis ceramic media for pretreatment of tetracycline wastewater.

PubMed

Yang, Kunlun; Jin, Yang; Yue, Qinyan; Zhao, Pin; Gao, Yuan; Wu, Suqing; Gao, Baoyu

2017-05-01

Application of modified sintering ferric-carbon ceramics (SFC) and sintering-free ferric-carbon ceramics (SFFC) based on coal ash and scrap iron for pretreatment of tetracycline (TET) wastewater was investigated in this article. Physical property, morphological character, toxic metal leaching content, and crystal component were studied to explore the application possibility of novel ceramics in micro-electrolysis reactors. The influences of operating conditions including influent pH, hydraulic retention time (HRT), and air-water ratio (A/W) on the removal of tetracycline were studied. The results showed that SFC and SFFC were suitable for application in micro-electrolysis reactors. The optimum conditions of SFC reactor were pH of 3, HRT of 7 h, and A/W of 10. For SFFC reactor, the optimum conditions were pH of 2, HRT of 7 h, and A/W of 15. In general, the TET removal efficiency of SFC reactor was better than that of SFFC reactor. However, the harden resistance of SFFC was better than that of SFC. Furthermore, the biodegradability of TET wastewater was improved greatly after micro-electrolysis pretreatment for both SFC and SFFC reactors.

Coliform culturability in over- versus undersaturated drinking waters.

PubMed

Grandjean, D; Fass, S; Tozza, D; Cavard, J; Lahoussine, V; Saby, S; Guilloteau, H; Block, J-C

2005-05-01

The culturability of Escherichia coli in undersaturated drinking water with respect to CaCO3 (corrosive water) or in oversaturated water (non-corrosive water) was tested in different reactors: glass flasks (batch, "non-reactive" wall); glass reactors (chemostat, "non-reactive" wall) versus a corroded cast iron Propella reactor (chemostat, "reactive" wall) and a 15-year-old distribution system pilot (chemostat, "reactive" wall with 1% corroded cast iron and 99% cement-lined cast iron). The E. coli in E. coli-spiked drinking water was not able to maintain its culturability and colonize the experimental systems. It appears from our results that the optimal pH for maintaining E. coli culturability was around 8.2 or higher. However, in reactors with a reactive wall (corroded cast iron), the decline in E. coli culturability was slower when the pH was adjusted to 7.9 or 7.7 (i.e. a reactor fed with corrosive water; pHpHs). We tentatively deduce that corrosion products coming from chemical reactions driven by corrosive waters on the pipe wall improve E. coli culturability.

ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

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

Lell, R. M.; Schaefer, R. W.; McKnight, R. D.

Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactormore » physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of demonstration-size LMFBRs. As a benchmark, ZPR-6/7 was devoid of many 'real' reactor features, such as simulated control rods and multiple enrichment zones, in its reference form. Those kinds of features were investigated experimentally in variants of the reference ZPR-6/7 or in other critical assemblies in the Demonstration Reactor Benchmark Program.« less

Anaerobic digestion of pressed off leachate from the organic fraction of municipal solid waste

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

Nayono, Satoto E.; Institute of Biology for Engineers and Biotechnology of Wastewater, University of Karlsruhe, Am Fasanengarten, 76131 Karlsruhe; Winter, Josef, E-mail: josef.winter@iba.uka.d

2010-10-15

A highly polluted liquid ('press water') was obtained from the pressing facility for the organic fraction of municipal solid waste in a composting plant. Methane productivity of the squeezed-off leachate was investigated in batch assays. To assess the technical feasibility of 'press water' as a substrate for anaerobic digestion, a laboratory-scale glass column reactor was operated semi-continuously at 37 {sup o}C. A high methane productivity of 270 m{sup -3} CH{sub 4} ton{sup -1} COD{sub added} or 490 m{sup -3} CH{sub 4} ton{sup -1} VS{sub added} was achieved in the batch experiment. The semi-continuously run laboratory-scale reactor was initially operated atmore » an organic loading rate of 10.7 kg COD m{sup -3} d{sup -1}. The loading was increased to finally 27.7 kg COD m{sup -3} d{sup -1}, corresponding to a reduction of the hydraulic retention time from initially 20 to finally 7.7 days. During the digestion, a stable elimination of organic material (measured as COD elimination) of approximately 60% was achieved. Linearly with the increment of the OLR, the volumetric methane production of the reactor increased from 2.6 m{sup 3} m{sub reactor}{sup -3} d{sup -1} to 7.1 m{sup 3} m{sub reactor}{sup -3} d{sup -1}. The results indicated that 'press water' from the organic fraction of municipal solid waste was a suitable substrate for anaerobic digestion which gave a high biogas yield even at very high loading rates.« less

Investigation of a para-ortho hydrogen reactor for application to spacecraft sensor cooling

NASA Technical Reports Server (NTRS)

Nast, T. C.

1983-01-01

The utilization of solid hydrogen in space for sensor and instrument cooling is a very efficient technique for long term cooling or for cooling at high heat rates. The solid hydrogen can provide temperatures as low as 7 to 8 K to instruments. Vapor cooling is utilized to reduce parasitic heat inputs to the 7 to 8 K stage and is effective in providing intermediate cooling for instrument components operating at higher temperatures. The use of solid hydrogen in place of helium may lead to weight reductions as large as a factor of ten and an attendent reduction in system volume. The results of an investigation of a catalytic reactor for use with a solid hydrogen cooling system is presented. Trade studies were performed on several configurations of reactor to meet the requirements of high reactor efficiency with low pressure drop. Results for the selected reactor design are presented for both liquid hydrogen systems operating at near atmospheric pressure and the solid hydrogen cooler operating as low as 1 torr.

Dynamics of heat-pipe reactors

NASA Technical Reports Server (NTRS)

Niederauer, G. F.

1971-01-01

A split-core heat pipe reactor, fueled with either U(233)C or U(235)C in a tungsten cermet and cooled by 7-Li-W heat pipes, was examined for the effects of the heat pipes on reactor while trying to safely absorb large reactivity inputs through inherent shutdown mechanisms. Limits on ramp reactivity inputs due to fuel melting temperature and heat pipe wall heat flux were mapped for the reactor in both startup and at-power operating modes.

Crystal Structure of the Extracellular Cholinesterase-Like Domain from Neuroligin-2

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

Koehnke,J.; Jin, X.; Budreck, E.

Neuroligins (NLs) are catalytically inactive members of a family of cholinesterase-like transmembrane proteins that mediate cell adhesion at neuronal synapses. Postsynaptic neuroligins engage in Ca2+-dependent transsynaptic interactions via their extracellular cholinesterase domain with presynaptic neurexins (NRXs). These interactions may be regulated by two short splice insertions (termed A and B) in the NL cholinesterase domain. Here, we present the 3.3- Angstroms crystal structure of the ectodomain from NL2 containing splice insertion A (NL2A). The overall structure of NL2A resembles that of cholinesterases, but several structural features are unique to the NL proteins. First, structural elements surrounding the esterase active-site regionmore » differ significantly between active esterases and NL2A. On the opposite surface of the NL2A molecule, the positions of the A and B splice insertions identify a candidate NRX interaction site of the NL protein. Finally, sequence comparisons of NL isoforms allow for mapping the location of residues of previously identified mutations in NL3 and NL4 found in patients with autism spectrum disorders. Overall, the NL2 structure promises to provide a valuable model for dissecting NL isoform- and synapse-specific functions.« less

Crystal structure of the extracellular cholinesterase-like domain from neuroligin-2

PubMed Central

Koehnke, Jesko; Jin, Xiangshu; Budreck, Elaine C.; Posy, Shoshana; Scheiffele, Peter; Honig, Barry; Shapiro, Lawrence

2008-01-01

Neuroligins (NLs) are catalytically inactive members of a family of cholinesterase-like transmembrane proteins that mediate cell adhesion at neuronal synapses. Postsynaptic neuroligins engage in Ca2+-dependent transsynaptic interactions via their extracellular cholinesterase domain with presynaptic neurexins (NRXs). These interactions may be regulated by two short splice insertions (termed A and B) in the NL cholinesterase domain. Here, we present the 3.3-Å crystal structure of the ectodomain from NL2 containing splice insertion A (NL2A). The overall structure of NL2A resembles that of cholinesterases, but several structural features are unique to the NL proteins. First, structural elements surrounding the esterase active-site region differ significantly between active esterases and NL2A. On the opposite surface of the NL2A molecule, the positions of the A and B splice insertions identify a candidate NRX interaction site of the NL protein. Finally, sequence comparisons of NL isoforms allow for mapping the location of residues of previously identified mutations in NL3 and NL4 found in patients with autism spectrum disorders. Overall, the NL2 structure promises to provide a valuable model for dissecting NL isoform- and synapse-specific functions. PMID:18250328

Criticality experiments and analysis of molybdenum reflected cylindrical uranyl fluoride water solution reactors

NASA Technical Reports Server (NTRS)

Fieno, D.; Fox, T.; Mueller, R.

1972-01-01

Clean criticality data were obtained from molybdenum-reflected cylindrical uranyl-fluoride-water solution reactors. Using ENDF/B molybdenum cross sections, a nine energy group two-dimensional transport calculation of a reflected reactor configuration predicted criticality to within 7 cents of the experimental value. For these reactors, it was necessary to compute the reflector resonance integral by a detailed transport calculation at the core-reflector interface volume in the energy region of the two dominant resonances of natural molybdenum.

75 FR 11375 - Revision of Fee Schedules; Fee Recovery for FY 2010

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-10

... Spent Fuel Storage/Reactor Decommissioning..... 2.7 0.2 0.2 Test and Research Reactors 0.2 0.0 0.0 Fuel... categories of licenses. The FY 2009 fee is also shown for comparative purposes. Table V--Rebaselined Annual...) Spent Fuel Storage/Reactor 122,000 143,000 Decommissioning Test and Research Reactors (Non-power 87,600...

[Study on the start-up of anaerobic ammonium oxidation process in biological activated carbon reactor].

PubMed

Lai, Wei-Yi; Zhou, Wei-Li; He, Sheng-Bing

2013-08-01

In order to shorten the start-up time of anaerobic ammonium oxidation (ANAMMOX) reactor, biological activated cabon reactor was applied. Three lab scale UASB reactors were seeded with anaerobic sludge, fed with synthetic wastewater containing ammonia and nitrite, and supplemented with granular activated carbon on day 0, 33 and 56, respectively. The nitrogen removal performance of the first reactor, into which GAC was added on day 0, showed no significant improvement in 90 days. After being suspended for about one month, the secondary start-up of this reactor succeeded in another 33 days (totally 123 days). 49 d and 85 d were taken for the other two reactors started up by the addition of GAC on day 33 and 56, respectively. After the reactors were started up, the average removal rates of total nitrogen were 89.8%, 86.7% and 86.7%, respectively. The start-up process could be divided into four stages, namely, the bacterial autolysis phase, the lag phase, the improve phase and the stationary phase, and the best time for adding GAC carrier was right after the start of the lag phase.

Synaptogenesis Is Modulated by Heparan Sulfate in Caenorhabditis elegans

PubMed Central

Lázaro-Peña, María I.; Díaz-Balzac, Carlos A.; Bülow, Hannes E.; Emmons, Scott W.

2018-01-01

The nervous system regulates complex behaviors through a network of neurons interconnected by synapses. How specific synaptic connections are genetically determined is still unclear. Male mating is the most complex behavior in Caenorhabditis elegans. It is composed of sequential steps that are governed by > 3000 chemical connections. Here, we show that heparan sulfates (HS) play a role in the formation and function of the male neural network. HS, sulfated in position 3 by the HS modification enzyme HST-3.1/HS 3-O-sulfotransferase and attached to the HS proteoglycan glypicans LON-2/glypican and GPN-1/glypican, functions cell-autonomously and nonautonomously for response to hermaphrodite contact during mating. Loss of 3-O sulfation resulted in the presynaptic accumulation of RAB-3, a molecule that localizes to synaptic vesicles, and disrupted the formation of synapses in a component of the mating circuits. We also show that the neural cell adhesion protein NRX-1/neurexin promotes and the neural cell adhesion protein NLG-1/neuroligin inhibits the formation of the same set of synapses in a parallel pathway. Thus, neural cell adhesion proteins and extracellular matrix components act together in the formation of synaptic connections. PMID:29559501

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

Brown, Nicholas R.; Powers, Jeffrey J.; Mueller, Don

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy researchmore » and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.« less

Target-fueled nuclear reactor for medical isotope production

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

Coats, Richard L.; Parma, Edward J.

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7more » to 21 days.« less

Si impurity concentration in nominally undoped Al0.7Ga0.3N grown in a planetary MOVPE reactor

NASA Astrophysics Data System (ADS)

Jeschke, J.; Knauer, A.; Weyers, M.

2018-02-01

The unintentional silicon incorporation during the metalorganic vapor phase epitaxy (MOVPE) of nominally undoped Al0.7Ga0.3N in a Planetary Reactor under various growth conditions was investigated. Dependent on growth temperature, pressure and V/III ratio, Si concentrations of below 1 × 1016 up to 4 × 1017 cm-3 were measured. Potential Si sources are discussed and, by comparing samples grown in a SiC coated reactor setup and in a TaC coated setup, the SiC coatings in the reactor are identified as the most likely source for the unintentional Si doping at elevated temperatures above 1080 °C. Under identical growth conditions the background Si concentration can be reduced by up to an order of magnitude when using TaC coatings.

Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

NASA Technical Reports Server (NTRS)

Valentine, Peter G.; Allen, Lee R.; Shapiro, Alan P.

2012-01-01

With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

78 FR 32279 - Advisory Committee On Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-29

... NUCLEAR REGULATORY COMMISSION Advisory Committee On Reactor Safeguards; Notice of Meeting In accordance with the purposes of Sections 29 and 182b of the Atomic Energy Act (42 U.S.C. 2039, 2232b), the Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on June 5-7, 2013, 11545 Rockville Pike...

Cavity temperature and flow characteristics in a gas-core test reactor

NASA Technical Reports Server (NTRS)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

The thermal decomposition of the benzyl radical in a heated micro-reactor. II. Pyrolysis of the tropyl radical

DOE PAGES

Buckingham, Grant T.; Porterfield, Jessica P.; Kostko, Oleg; ...

2016-07-05

Cycloheptatrienyl (tropyl) radical, C 7H 7, was cleanly produced in the gas-phase, entrained in He or Ne carrier gas, and subjected to a set of flash-pyrolysis micro-reactors. In this study, the pyrolysis products resulting from C 7H 7 were detected and identified by vacuum ultraviolet photoionization mass spectrometry. Complementary product identification was provided by infrared absorption spectroscopy. Pyrolysis pressures in the micro-reactor were roughly 200 Torr and residence times were approximately 100 μs. Thermal cracking of tropyl radical begins at 1100 K and the products from pyrolysis of C 7H 7 are only acetylene and cyclopentadienyl radicals. Tropyl radicals domore » not isomerize to benzyl radicals at reactor temperatures up to 1600 K. Heating samples of either cycloheptatriene or norbornadiene never produced tropyl (C 7H 7) radicals but rather only benzyl (C 6H 5CH 2). The thermal decomposition of benzyl radicals has been reconsidered without participation of tropyl radicals. There are at least three distinct pathways for pyrolysis of benzyl radical: the Benson fragmentation, the methyl-phenyl radical, and the bridgehead norbornadienyl radical. These three pathways account for the majority of the products detected following pyrolysis of all of the isotopomers: C 6H 5CH 2, C 6H 5CD 2, C 6D 5CH 2, and C 6H 5 13CH 2. Finally, analysis of the temperature dependence for the pyrolysis of the isotopic species (C 6H 5CD 2, C 6D 5CH 2, and C 6H 5 13CH 2) suggests the Benson fragmentation and the norbornadienyl pathways open at reactor temperatures of 1300 K while the methyl-phenyl radical channel becomes active at slightly higher temperatures (1500 K).« less

The thermal decomposition of the benzyl radical in a heated micro-reactor. II. Pyrolysis of the tropyl radical

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

Buckingham, Grant T.; National Bioenergy Center, National Renewable Energy Laboratory, 15013 Denver West Parkway, Golden Colorado 80401; Porterfield, Jessica P.

2016-07-07

Cycloheptatrienyl (tropyl) radical, C{sub 7}H{sub 7}, was cleanly produced in the gas-phase, entrained in He or Ne carrier gas, and subjected to a set of flash-pyrolysis micro-reactors. The pyrolysis products resulting from C{sub 7}H{sub 7} were detected and identified by vacuum ultraviolet photoionization mass spectrometry. Complementary product identification was provided by infrared absorption spectroscopy. Pyrolysis pressures in the micro-reactor were roughly 200 Torr and residence times were approximately 100 μs. Thermal cracking of tropyl radical begins at 1100 K and the products from pyrolysis of C{sub 7}H{sub 7} are only acetylene and cyclopentadienyl radicals. Tropyl radicals do not isomerize tomore » benzyl radicals at reactor temperatures up to 1600 K. Heating samples of either cycloheptatriene or norbornadiene never produced tropyl (C{sub 7}H{sub 7}) radicals but rather only benzyl (C{sub 6}H{sub 5}CH{sub 2}). The thermal decomposition of benzyl radicals has been reconsidered without participation of tropyl radicals. There are at least three distinct pathways for pyrolysis of benzyl radical: the Benson fragmentation, the methyl-phenyl radical, and the bridgehead norbornadienyl radical. These three pathways account for the majority of the products detected following pyrolysis of all of the isotopomers: C{sub 6}H{sub 5}CH{sub 2}, C{sub 6}H{sub 5}CD{sub 2}, C{sub 6}D{sub 5}CH{sub 2}, and C{sub 6}H{sub 5}{sup 13}CH{sub 2}. Analysis of the temperature dependence for the pyrolysis of the isotopic species (C{sub 6}H{sub 5}CD{sub 2}, C{sub 6}D{sub 5}CH{sub 2}, and C{sub 6}H{sub 5}{sup 13}CH{sub 2}) suggests the Benson fragmentation and the norbornadienyl pathways open at reactor temperatures of 1300 K while the methyl-phenyl radical channel becomes active at slightly higher temperatures (1500 K).« less

Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

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

Maloy, Stuart Andrew

In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

Precomputed state dependent digital control of a nuclear rocket engine

NASA Technical Reports Server (NTRS)

Johnson, M. R.

1972-01-01

A control method applicable to multiple-input multiple-output nonlinear time-invariant systems in which desired behavior can be expressed explicitly as a trajectory in system state space is developed. The precomputed state dependent control method is basically a synthesis technique in which a suboptimal control law is developed off-line, prior to system operation. This law is obtained by conducting searches at a finite number of points in state space, in the vicinity of some desired trajectory, to obtain a set of constant control vectors which tend to return the system to the desired trajectory. These vectors are used to evaluate the unknown coefficients in a control law having an assumed hyperellipsoidal form. The resulting coefficients constitute the heart of the controller and are used in the on-line computation of control vectors. Two examples of PSDC are given prior to the more detailed description of the NERVA control system development.

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

High-Temperature Fluid-Wall Reactor Technology Research, Test and Evaluation Performed at Naval Construction Battalion Center, Gulfport, Mississippi, for the United States Air Force Installation/Restoration Program

DTIC Science & Technology

1988-01-01

the reactor Duties: The Process Engineers rotate with the Lead Operator to monitor the process at the top of the reactor through the site glass...pant cuffs and coverhoods of coveralls, will be attached to gloves, boots and coveralls, using duct tape. * IF AMBIENT WORK STATIONS TEMPERATURE IS...L of the sample fortification solution (Section ýý8) containing 1C 12-2,3,7,8-TCDD at a concentration of 0.5 ng/1,Land C14-2,3,7,8-TCDD at a

Revised Point of Departure Design Options for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Fittje, James E.; Borowski, Stanley K.; Schnitzler, Bruce

2015-01-01

In an effort to further refine potential point of departure nuclear thermal rocket engine designs, four proposed engine designs representing two thrust classes and utilizing two different fuel matrix types are designed and analyzed from both a neutronics and thermodynamic cycle perspective. Two of these nuclear rocket engine designs employ a tungsten and uranium dioxide cermet (ceramic-metal) fuel with a prismatic geometry based on the ANL-200 and the GE-710, while the other two designs utilize uranium-zirconium-carbide in a graphite composite fuel and a prismatic fuel element geometry developed during the Rover/NERVA Programs. Two engines are analyzed for each fuel type, a small criticality limited design and a 111 kN (25 klbf) thrust class engine design, which has been the focus of numerous manned mission studies, including NASA's Design Reference Architecture 5.0. slightly higher T/W ratios, but they required substantially more 235U.

Performance Study of Chromium (VI) Removal in Presence of Phenol in a Continuous Packed Bed Reactor by Escherichia coli Isolated from East Calcutta Wetlands

PubMed Central

Chakraborty, Bhaswati; Indra, Suvendu; Hazra, Ditipriya; Betai, Rupal; Ray, Lalitagauri; Basu, Srabanti

2013-01-01

Organic pollutants, like phenol, along with heavy metals, like chromium, are present in various industrial effluents that pose serious health hazard to humans. The present study looked at removal of chromium (VI) in presence of phenol in a counter-current continuous packed bed reactor packed with E. coli cells immobilized on clay chips. The cells removed 85% of 500 mg/L of chromium (VI) from MS media containing glucose. Glucose was then replaced by 500 mg/L phenol. Temperature and pH of the MS media prior to addition of phenol were 30°C and 7, respectively. Hydraulic retention times of phenol- and chromium (VI)-containing synthetic media and air flow rates were varied to study the removal efficiency of the reactor system. Then temperature conditions of the reactor system were varied from 10°C to 50°C, the optimum being 30°C. The pH of the media was varied from pH 1 to pH 12, and the optimum pH was found to be 7. The maximum removal efficiency of 77.7% was achieved for synthetic media containing phenol and chromium (VI) in the continuous reactor system at optimized conditions, namely, hydraulic retention time at 4.44 hr, air flow rate at 2.5 lpm, temperature at 30°C, and pH at 7. PMID:24073400

Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

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

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2014-10-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC)more » system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.« less

Nuclear Thermal Rocket/Vehicle Design Options for Future NASA Missions to the Moon and Mars

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.; Corban, Robert R.; Mcguire, Melissa L.; Beke, Erik G.

1995-01-01

The nuclear thermal rocket (NTR) provides a unique propulsion capability to planners/designers of future human exploration missions to the Moon and Mars. In addition to its high specific impulse (approximately 850-1000 s) and engine thrust-to-weight ratio (approximately 3-10), the NTR can also be configured as a 'dual mode' system capable of generating electrical power for spacecraft environmental systems, communications, and enhanced stage operations (e.g., refrigeration for long-term liquid hydrogen storage). At present the Nuclear Propulsion Office (NPO) is examining a variety of mission applications for the NTR ranging from an expendable, single-burn, trans-lunar injection (TLI) stage for NASA's First Lunar Outpost (FLO) mission to all propulsive, multiburn, NTR-powered spacecraft supporting a 'split cargo-piloted sprint' Mars mission architecture. Each application results in a particular set of requirements in areas such as the number of engines and their respective thrust levels, restart capability, fuel operating temperature and lifetime, cryofluid storage, and stage size. Two solid core NTR concepts are examined -- one based on NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor (NDR) technology, and a second concept which utilizes a ternary carbide 'twisted ribbon' fuel form developed by the Commonwealth of Independent States (CIS). The NDR and CIS concepts have an established technology database involving significant nuclear testing at or near representative operating conditions. Integrated systems and mission studies indicate that clusters of two to four 15 to 25 klbf NDR or CIS engines are sufficient for most of the lunar and Mars mission scenarios currently under consideration. This paper provides descriptions and performance characteristics for the NDR and CIS concepts, summarizes NASA's First Lunar Outpost and Mars mission scenarios, and describes characteristics for representative cargo and piloted vehicles compatible with a reference 240 t-class heavy lift launch vehicle (HLLV) and smaller 120 t HLLV option. Attractive performance characteristics and high-leverage technologies associated with both the engine and stage are identified, and supporting parametric sensitivity data is provided. The potential for commonality of engine and stage components to satisfy a broad range of lunar and Mars missions is also discussed.

Personal notes [of D.S. Lewis, 7 September 1956 to 31 December 1959

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

Lewis, D.S.

1956-09-07

This report is a copy of the personal log of D.S. Lewis of the Irradiation Processing Dept. of Reactor Operations at Hanford and covers the period from 7 September 1956 through 31 December 1959. Data are presented on the following: (1) basic reactor operating data, including daily operating data, outage resumes, injuries and incidents, charging and tube replacement rates, panellit gage (flowmeter) trip failures, and thermocouple failures, and (2) basic reactor information on the water plant, electrical distribution, VSR`s, HCR`s, Ball 3X, Safety circuits, gas system, effluent system, process tube cross-section, and production scheduling.

Muon trackers for imaging a nuclear reactor

NASA Astrophysics Data System (ADS)

Kume, N.; Miyadera, H.; Morris, C. L.; Bacon, J.; Borozdin, K. N.; Durham, J. M.; Fuzita, K.; Guardincerri, E.; Izumi, M.; Nakayama, K.; Saltus, M.; Sugita, T.; Takakura, K.; Yoshioka, K.

2016-09-01

A detector system for assessing damage to the cores of the Fukushima Daiichi nuclear reactors by using cosmic-ray muon tomography was developed. The system consists of a pair of drift-tube tracking detectors of 7.2× 7.2-m2 area. Each muon tracker consists of 6 x-layer and 6 y-layer drift-tube detectors. Each tracker is capable of measuring muon tracks with 12 mrad angular resolutions, and is capable of operating under 50-μ Sv/h radiation environment by removing gamma induced background with a novel time-coincidence logic. An estimated resolution to observe nuclear fuel debris at Fukushima Daiichi is 0.3 m when the core is imaged from outside the reactor building.

RELAP-7 Closure Correlations

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

Zou, Ling; Berry, R. A.; Martineau, R. C.

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework, MOOSE (Multi-Physics Object Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s and TRACE’s capabilities and extends their analysis capabilities for all reactor system simulation scenarios. The RELAP-7 codemore » utilizes the well-posed 7-equation two-phase flow model for compressible two-phase flow. Closure models used in the TRACE code has been reviewed and selected to reflect the progress made during the past decades and provide a basis for the colure correlations implemented in the RELAP-7 code. This document provides a summary on the closure correlations that are currently implemented in the RELAP-7 code. The closure correlations include sub-grid models that describe interactions between the fluids and the flow channel, and interactions between the two phases.« less

A comparison of the technological effectiveness of dairy wastewater treatment in anaerobic UASB reactor and anaerobic reactor with an innovative design.

PubMed

Jedrzejewska-Cicinska, M; Kozak, K; Krzemieniewski, M

2007-10-01

The present research was an investigation of the influence of an innovative design of reactor filled with polyethylene (PE) granulate on model dairy wastewater treatment efficiency under anaerobic conditions compared to that obtained in a typical UASB reactor. The experiment was conducted at laboratory scale. An innovative reactor was designed with the reaction chamber inclined 30 degrees in relation to the ground with upward waste flow and was filled with PE granular material. Raw model dairy wastewater was fed to two anaerobic reactors of different design at the organic loading rate of 4 kg COD m(-3)d(-1). Throughout the experiment, a higher removal efficiency of organic compounds was observed in the reactor with an innovative design and it was higher by 7.1% on average than in the UASB reactor. The total suspended solids was lower in the wastewater treated in the anaerobic reactor with the innovative design. Applying a PE granulated filling in the chamber of the innovative reactor contributed to an even distribution of sludge biomass in the reactor, reducing washout of anaerobic sludge biomass from the reaction chamber and giving a higher organic compounds removal efficiency.

Compact power reactor

DOEpatents

Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

Key role of temperature monitoring in interpretation of microwave effect on transesterification and esterification reactions for biodiesel production.

PubMed

Mazubert, Alex; Taylor, Cameron; Aubin, Joelle; Poux, Martine

2014-06-01

Microwave effects have been quantified, comparing activation energies and pre-exponential factors to those obtained in a conventionally-heated reactor for biodiesel production from waste cooking oils via transesterification and esterification reactions. Several publications report an enhancement of biodiesel production using microwaves, however recent reviews highlight poor temperature measurements in microwave reactors give misleading reaction performances. Operating conditions have therefore been carefully chosen to investigate non-thermal microwave effects alone. Temperature is monitored by an optical fiber sensor, which is more accurate than infrared sensors. For the transesterification reaction, the activation energy is 37.1kJ/mol (20.1-54.2kJ/mol) in the microwave-heated reactor compared with 31.6kJ/mol (14.6-48.7kJ/mol) in the conventionally-heated reactor. For the esterification reaction, the activation energy is 45.4kJ/mol (31.8-58.9kJ/mol) for the microwave-heated reactor compared with 56.1kJ/mol (55.7-56.4kJ/mol) for conventionally-heated reactor. The results confirm the absence of non-thermal microwave effects for homogenous-catalyzed reactions. Copyright © 2014 Elsevier Ltd. All rights reserved.

A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

DOE PAGES

Mandelli, Diego; Prescott, Steven; Smith, Curtis; ...

2015-05-17

In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code calledmore » NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.« less

75 FR 42469 - Firstenergy Nuclear Operating Company; Request for Licensing Action

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-21

... nuclear plant in Ohio, preventing the reactor from restarting until such time that the NRC determines... Commission's regulations. The request has been referred to the Director of the Office of Nuclear Reactor... of Nuclear Reactor Regulation. [FR Doc. 2010-17834 Filed 7-20-10; 8:45 am] BILLING CODE 7590-01-P ...

Cleanup Verification Package for the 118-F-7, 100-F Miscellaneous Hardware Storage Vault

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

M. J. Appel

2006-11-02

This cleanup verification package documents completion of remedial action for the 118-F-7, 100-F Miscellaneous Hardware Storage Vault. The site consisted of an inactive solid waste storage vault used for temporary storage of slightly contaminated reactor parts that could be recovered and reused for the 100-F Area reactor operations.

Muon trackers for imaging a nuclear reactor

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

Kume, N.; Miyadera, H.; Morris, C. L.

A detector system for assessing damage to the cores of the Fukushima Daiichi nuclear reactors by using cosmic-ray muon tomography was developed. Furthermore, the system consists of a pair of drift-tube tracking detectors of 7.2× 7.2-m 2 area. In each muon tracker there consists 6 x-layer and 6 y-layer drift-tube detectors. Each tracker is capable of measuring muon tracks with 12 mrad angular resolutions, and is capable of operating under 50-μ Sv/h radiation environment by removing gamma induced background with a novel time-coincidence logic. An estimated resolution to observe nuclear fuel debris at Fukushima Daiichi is 0.3 m when themore » core is imaged from outside the reactor building.« less

Muon trackers for imaging a nuclear reactor

DOE PAGES

Kume, N.; Miyadera, H.; Morris, C. L.; ...

2016-09-21

A detector system for assessing damage to the cores of the Fukushima Daiichi nuclear reactors by using cosmic-ray muon tomography was developed. Furthermore, the system consists of a pair of drift-tube tracking detectors of 7.2× 7.2-m 2 area. In each muon tracker there consists 6 x-layer and 6 y-layer drift-tube detectors. Each tracker is capable of measuring muon tracks with 12 mrad angular resolutions, and is capable of operating under 50-μ Sv/h radiation environment by removing gamma induced background with a novel time-coincidence logic. An estimated resolution to observe nuclear fuel debris at Fukushima Daiichi is 0.3 m when themore » core is imaged from outside the reactor building.« less

Fate of LCFA in the co-digestion of cow manure, food waste and discontinuous addition of oil.

PubMed

Neves, L; Oliveira, R; Alves, M M

2009-12-01

Different concentrations of oily waste were added in a discontinuous mode and recurrently to anaerobic continuous stirred tank reactors fed with cow manure and food waste. Four continuous stirred tank reactors were run in parallel. A control reactor (R1) received no additional oil and R2, R3 and R4 received increasing concentrations of oil in two different experimental approaches. First, the lipids composition was forced to change suddenly, in three moments, without changing the total chemical oxygen demand (COD) fed to the reactors. The only long chain fatty acid (LCFA) detected onto the R1 solid matrix was palmitic acid (C16:0). Nevertheless in the solid matrix of R2, R3 and R4C16:0 and stearic acid were detected. For occasional increase in the oil concentration up to 7.7gCOD(oil)/L(reactor) (55% Oil(COD)/Total(COD)) no statistical differences were detected between the reactors, in terms of methane production, effluent soluble COD, effluent volatile fatty acids and total and volatile solids removal. Therefore this experiment allowed to conclude that cow manure-food waste co-digestion presents sufficient buffer capacity to endure solid-associated LCFA concentration up to 20-25gCOD-LCFA/kgTS. In a second experiment higher concentrations of oil were added, raising occasionally the concentration in the reactors to 9, 12, 15 and 18gCOD(oil)/L(reactor). All pulses had a positive effect in methane production, with the exception of the highest oil pulse concentration, that persistently impaired the reactor performance. This experiment demonstrates that threshold values for LCFA and C16:0 accumulation onto the solid matrix, of about 180-220gCOD-LCFA/kgTS and 120-150gCOD-C16:0/kgTS, should not be surpassed in order to prevent persistent reactor failure, as occurs in some full scale co-digestion plants.

Development and Characterization of 6Li-doped Liquid Scintillator Detectors for PROSPECT

NASA Astrophysics Data System (ADS)

Gaison, Jeremy; Prospect Collaboration

2016-09-01

PROSPECT, the Precision Reactor Oscillation and Spectrum experiment, is a phased reactor antineutrino experiment designed to search for eV-scale sterile neutrinos via short-baseline neutrino oscillations and to make a precision measurement of the 235U reactor antineutrino spectrum. A multi-ton, optically segmented detector will be deployed at Oak Ridge National Laboratory's (ORNL) High Flux Isotope Reactor (HFIR) to measure the reactor spectrum at baselines ranging from 7-12m. A two-segment detector prototype with 50 liters of active liquid scintillator target has been built to verify the detector design and to benchmark its performance. In this presentation, we will summarize the performance of this detector prototype and describe the optical and energy calibration of the segmented PROSPECT detectors.

RADIATION DAMAGE IN REACTOR MATERIALS. Proceedings of the Symposium on Radiation Damage in Solids and Reactor Materials Held in Venice, 7-11 May 1962

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

None

1964-10-31

Thirty papers and 3 reviews of papers and panel discussions presented at the Symposium on Radiation Damage in Solids and Reactor Materials are given. Eighteen papers were previously abstracted for NSA. Separate abstracts were prepared for the remaining 15 papers. (M.C.G.)

Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

NASA Astrophysics Data System (ADS)

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

2016-10-01

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

MTR WING, TRA604. FIRST FLOOR PLAN. ENTRY LOBBY, MACHINE SHOP, ...

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

MTR WING, TRA-604. FIRST FLOOR PLAN. ENTRY LOBBY, MACHINE SHOP, INSTRUMENT SHOP, COUNTING ROOM, HEALTH PHYSICS LAB, LABS AND OFFICES, STORAGE, SHIPPING AND RECEIVING. BLAW-KNOX 3150-4-2, 7/1950. INL INDEX NO. 053-604-00-099-100008, REV. 7. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Efficient preparation of enantiopure D-phenylalanine through asymmetric resolution using immobilized phenylalanine ammonia-lyase from Rhodotorula glutinis JN-1 in a recirculating packed-bed reactor.

PubMed

Zhu, Longbao; Zhou, Li; Huang, Nan; Cui, Wenjing; Liu, Zhongmei; Xiao, Ke; Zhou, Zhemin

2014-01-01

An efficient enzymatic process was developed to produce optically pure D-phenylalanine through asymmetric resolution of the racemic DL-phenylalanine using immobilized phenylalanine ammonia-lyase (RgPAL) from Rhodotorula glutinis JN-1. RgPAL was immobilized on a modified mesoporous silica support (MCM-41-NH-GA). The resulting MCM-41-NH-GA-RgPAL showed high activity and stability. The resolution efficiency using MCM-41-NH-GA-RgPAL in a recirculating packed-bed reactor (RPBR) was higher than that in a stirred-tank reactor. Under optimal operational conditions, the volumetric conversion rate of L-phenylalanine and the productivity of D-phenylalanine reached 96.7 mM h⁻¹ and 0.32 g L⁻¹ h⁻¹, respectively. The optical purity (eeD) of D-phenylalanine exceeded 99%. The RPBR ran continuously for 16 batches, the conversion ratio did not decrease. The reactor was scaled up 25-fold, and the productivity of D-phenylalanine (eeD>99%) in the scaled-up reactor reached 7.2 g L⁻¹ h⁻¹. These results suggest that the resolution process is an alternative method to produce highly pure D-phenylalanine.

Efficient Preparation of Enantiopure D-Phenylalanine through Asymmetric Resolution Using Immobilized Phenylalanine Ammonia-Lyase from Rhodotorula glutinis JN-1 in a Recirculating Packed-Bed Reactor

PubMed Central

Huang, Nan; Cui, Wenjing; Liu, Zhongmei; Xiao, Ke; Zhou, Zhemin

2014-01-01

An efficient enzymatic process was developed to produce optically pure D-phenylalanine through asymmetric resolution of the racemic DL-phenylalanine using immobilized phenylalanine ammonia-lyase (RgPAL) from Rhodotorula glutinis JN-1. RgPAL was immobilized on a modified mesoporous silica support (MCM-41-NH-GA). The resulting MCM-41-NH-GA-RgPAL showed high activity and stability. The resolution efficiency using MCM-41-NH-GA-RgPAL in a recirculating packed-bed reactor (RPBR) was higher than that in a stirred-tank reactor. Under optimal operational conditions, the volumetric conversion rate of L-phenylalanine and the productivity of D-phenylalanine reached 96.7 mM h−1 and 0.32 g L−1 h−1, respectively. The optical purity (ee D) of D-phenylalanine exceeded 99%. The RPBR ran continuously for 16 batches, the conversion ratio did not decrease. The reactor was scaled up 25-fold, and the productivity of D-phenylalanine (ee D>99%) in the scaled-up reactor reached 7.2 g L−1 h−1. These results suggest that the resolution process is an alternative method to produce highly pure D-phenylalanine. PMID:25268937

The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

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

De Bruyn, D.; Engelen, J.; Ortega, A.

MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of threemore » years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)« less

MTR, TRA603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTYMETER CHOPPER ...

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

MTR, TRA-603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTY-METER CHOPPER HOUSE. COFFIN TURNING ROLLS. REMOVABLE PANEL OVER CANAL ON EAST SIDE. NEW PLUG STORAGE ACCESS. DOOR SCHEDULE INDICATES STEEL (FOR VAULT), WIRE MESH, AND HOLLOW METAL TYPES. STORAGE AND ISSUE ROOM. SAFETY SHOWERS. DOORWAY TO WING, TRA-604. BLAW-KNOX 3150-803-2, 7/1950. INL INDEX NO. 531-0603-00-098-100561, REV. 10. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

High-calorific biogas production from anaerobic digestion of food waste using a two-phase pressurized biofilm (TPPB) system.

PubMed

Li, Yeqing; Liu, Hong; Yan, Fang; Su, Dongfang; Wang, Yafei; Zhou, Hongjun

2017-01-01

To obtain high calorific biogas via anaerobic digestion without additional upgrading equipment, a two-phase pressurized biofilm system was built up, including a conventional continuously stirred tank reactor and a pressurized biofilm anaerobic reactor (PBAR). Four different pressure levels (0.3, 0.6, 1.0 and 1.7MPa) were applied to the PBAR in sequence, with the organic loading rate maintained at 3.1g-COD/L/d. Biogas production, gas composition, process stability parameters were measured. Results showed that with the pressure increasing from 0.3MPa to 1.7MPa, the pH value decreased from 7.22±0.19 to 6.98±0.05, the COD removal decreased from 93.0±0.9% to 79.7±1.2% and the methane content increased from 80.5±1.5% to 90.8±0.8%. Biogas with higher calorific value of 36.2MJ/m 3 was obtained at a pressure of 1.7MPa. Pressure showed a significant effect on biogas production and gas quality in methanogenesis reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

Multi-Megawatt Space Nuclear Power Generation

DTIC Science & Technology

1993-06-28

electric generation, both for open- and closed-cycle opera- tion. These reactors use the particulate fuel of the type developed for HTGR reactors. What...commercial HTGR power reactors, the particles are held in place and directly cooled. Figure 2.7 shows the two types of fuel particles developed for...of MW(e), for pulsed energy devices. The FBR would use HTGR -type particle fuel , contained in a annular bed be- tween two porous frits. Helium would

Modification of UASB reactor by using CFD simulations for enhanced treatment of municipal sewage.

PubMed

Das, Suprotim; Sarkar, Supriya; Chaudhari, Sanjeev

2018-02-01

Up-flow anaerobic sludge blanket (UASB) has been in use since last few decades for the treatment of organic wastewaters. However, the performance of UASB reactor is quite low for treatment of low strength wastewaters (LSWs) due to less biogas production leading to poor mixing. In the present research work, a modification was done in the design of UASB to improve mixing of reactor liquid which is important to enhance the reactor performance. The modified UASB (MUASB) reactor was designed by providing a slanted baffle along the height of the reactor having an angle of 5.7° with the vertical wall. A two-dimensional computational fluid dynamics (CFD) simulation of three phase gas-liquid-solid flow in MUASB reactor was performed and compared with conventional UASB reactor. The CFD study indicated better mixing in terms of vorticity magnitude in MUASB reactor as compared to conventional UASB, which was reflected in the reactor performance. The performance of MUASB was compared with conventional UASB reactor for the onsite treatment of domestic sewage as LSW. Around 16% higher total chemical oxygen demand removal efficiency was observed in MUASB reactor as compared to conventional UASB during this study. Therefore, this MUASB model demonstrates a qualitative relationship between mixing and performance during the treatment of LSW. From the study, it seems that MUASB holds promise for field applications.

Fukushima Daiichi Muon Imaging

NASA Astrophysics Data System (ADS)

Miyadera, Haruo

2015-10-01

Japanese government announced cold-shutdown condition of the reactors at Fukushima Daiichi by the end of 2011, and mid- and long-term roadmap towards decommissioning has been drawn. However, little is known for the conditions of the cores because access to the reactors has been limited by the high radiation environment. The debris removal from the Unit 1 - 3 is planned to start as early as 2020, but the dismantlement is not easy without any realistic information of the damage to the cores, and the locations and amounts of the fuel debris. Soon after the disaster of Fukushima Daiichi, several teams in the US and Japan proposed to apply muon transmission or scattering imagings to provide information of the Fukushima Daiichi reactors without accessing inside the reactor building. GEANT4 modeling studies of Fukushima Daiichi Unit 1 and 2 showed clear superiority of the muon scattering method over conventional transmission method. The scattering method was demonstrated with a research reactor, Toshiba Nuclear Critical Assembly (NCA), where a fuel assembly was imaged with 3-cm resolution. The muon scattering imaging of Fukushima Daiichi was approved as a national project and is aiming at installing muon trackers to Unit 2. A proposed plan includes installation of muon trackers on the 2nd floor (operation floor) of turbine building, and in front of the reactor building. Two 7mx7m detectors were assembled at Toshiba and tested.

Measurements of the Reactor Antineutrino with Solid State Scintillation Detector

NASA Astrophysics Data System (ADS)

Alekseev, I.; Belov, V.; Brudanin, V.; Danilov, M.; Egorov, V.; Filosofov, D.; Fomina, M.; Hons, Z.; Kazartsev, S.; Kobyakin, A.; Kuznetsov, A.; Machikhiliyan, I.; Medvedev, D.; Nesterov, V.; Olshevsky, A.; Pogorelov, N.; Ponomarev, D.; Rozova, I.; Rumyantseva, N.; Rusinov, V.; Salamatin, A.; Samigullin, E.; Shevchik, Ye.; Shirchenko, M.; Shitov, Yu.; Skrobova, N.; Starostin, A.; Svirida, D.; Tarkovsky, E.; Tikhomirov, I.; Vlášek, J.; Zhitnikov, I.; Zinatulina, D.

Measurements of reactor antineutrino play an important role in the efforts at the frontier of the modern physics. The DANSS collaboration presents preliminary results of a one year run with a cubic meter solid state detector placed below 3.1 GW industrial light water reactor. The experiment is sensitive to sterile neutrino in the most interesting region of mixing parameter space. 2500 scintillation strips of the sensitive volume of the detector have multilayer passive shielding of copper, lead and borated polyethylene and active muon veto. Detector position below the reactor gives an advantage of overburden about 50 m of water equivalent providing factor of six in cosmic muon suppression and eliminating fast neutrons.The detector is placed on a vertically movable platform which allows to change the distance to the reactor core center in the range 10.7-12.7 m within a few minutes. The strips are read out individually by SiPMs and in groups of 50 by PMTs. 5000 inverse beta-decay events per day are collected in the fiducial volume, which is 78% of the whole detector, at the position closest to the reactor. Overburden, active veto and good segmentation of the detector result in an excellent signal to background ratio. The talk is dedicated to the data analysis and preliminary results. The experiment status is also presented.

An Early Sensitive Period Induces Long-Lasting Plasticity in the Honeybee Nervous System

PubMed Central

Grosso, Juan P.; Barneto, Jesica A.; Velarde, Rodrigo A.; Pagano, Eduardo A.; Zavala, Jorge A.; Farina, Walter M.

2018-01-01

The effect of early experiences on the brain during a sensitive period exerts a long-lasting influence on the mature individual. Despite behavioral and neural plasticity caused by early experiences having been reported in the honeybee Apis mellifera, the presence of a sensitive period in which associative experiences lead to pronounced modifications in the adult nervous system is still unclear. Laboratory-reared bees were fed with scented food within specific temporal windows and were assessed for memory retention, in the regulation of gene expression related to the synaptic formation and in the olfactory perception of their antennae at 17 days of age. Bees were able to retain a food-odor association acquired 5–8 days after emergence, but not before, and showed better retention than those exposed to an odor at 9–12 days. In the brain, the odor-rewarded experiences that occurred at 5–8 days of age boosted the expression levels of the cell adhesion proteins neurexin 1 (Nrx1) and neuroligin 2 (Nlg2) involved in synaptic strength. At the antennae, the experiences increased the electrical response to a novel odor but not to the one experienced. Therefore, a sensitive period that induces long-lasting behavioral, functional and structural changes is found in adult honeybees. PMID:29449804

An Early Sensitive Period Induces Long-Lasting Plasticity in the Honeybee Nervous System.

PubMed

Grosso, Juan P; Barneto, Jesica A; Velarde, Rodrigo A; Pagano, Eduardo A; Zavala, Jorge A; Farina, Walter M

2018-01-01

The effect of early experiences on the brain during a sensitive period exerts a long-lasting influence on the mature individual. Despite behavioral and neural plasticity caused by early experiences having been reported in the honeybee Apis mellifera , the presence of a sensitive period in which associative experiences lead to pronounced modifications in the adult nervous system is still unclear. Laboratory-reared bees were fed with scented food within specific temporal windows and were assessed for memory retention, in the regulation of gene expression related to the synaptic formation and in the olfactory perception of their antennae at 17 days of age. Bees were able to retain a food-odor association acquired 5-8 days after emergence, but not before, and showed better retention than those exposed to an odor at 9-12 days. In the brain, the odor-rewarded experiences that occurred at 5-8 days of age boosted the expression levels of the cell adhesion proteins neurexin 1 ( Nrx1 ) and neuroligin 2 ( Nlg2 ) involved in synaptic strength. At the antennae, the experiences increased the electrical response to a novel odor but not to the one experienced. Therefore, a sensitive period that induces long-lasting behavioral, functional and structural changes is found in adult honeybees.

Irradiation Microstructure of Austenitic Steels and Cast Steels Irradiated in the BOR-60 Reactor at 320°C

NASA Astrophysics Data System (ADS)

Yang, Yong; Chen, Yiren; Huang, Yina; Allen, Todd; Rao, Appajosula

Reactor internal components are subjected to neutron irradiation in light water reactors, and with the aging of nuclear power plants around the world, irradiation-induced material degradations are of concern for reactor internals. Irradiation-induced defects resulting from displacement damage are critical for understanding degradation in structural materials. In the present work, microstructural changes due to irradiation in austenitic stainless steels and cast steels were characterized using transmission electron microscopy. The specimens were irradiated in the BOR-60 reactor, a fast breeder reactor, up to 40 dpa at 320°C. The dose rate was approximately 9.4x10-7 dpa/s. Void swelling and irradiation defects were analyzed for these specimens. A high density of faulted loops dominated the irradiated-altered microstructures. Along with previous TEM results, a dose dependence of the defect structure was established at 320°C.

Online monitoring of the Osiris reactor with the Nucifer neutrino detector

NASA Astrophysics Data System (ADS)

Boireau, G.; Bouvet, L.; Collin, A. P.; Coulloux, G.; Cribier, M.; Deschamp, H.; Durand, V.; Fechner, M.; Fischer, V.; Gaffiot, J.; Gérard Castaing, N.; Granelli, R.; Kato, Y.; Lasserre, T.; Latron, L.; Legou, P.; Letourneau, A.; Lhuillier, D.; Mention, G.; Mueller, Th. A.; Nghiem, T.-A.; Pedrol, N.; Pelzer, J.; Pequignot, M.; Piret, Y.; Prono, G.; Scola, L.; Starzinski, P.; Vivier, M.; Dumonteil, E.; Mancusi, D.; Varignon, C.; Buck, C.; Lindner, M.; Bazoma, J.; Bouvier, S.; Bui, V. M.; Communeau, V.; Cucoanes, A.; Fallot, M.; Gautier, M.; Giot, L.; Guilloux, G.; Lenoir, M.; Martino, J.; Mercier, G.; Milleto, T.; Peuvrel, N.; Porta, A.; Le Quéré, N.; Renard, C.; Rigalleau, L. M.; Roy, D.; Vilajosana, T.; Yermia, F.; Nucifer Collaboration

2016-06-01

Originally designed as a new nuclear reactor monitoring device, the Nucifer detector has successfully detected its first neutrinos. We provide the second-shortest baseline measurement of the reactor neutrino flux. The detection of electron antineutrinos emitted in the decay chains of the fission products, combined with reactor core simulations, provides a new tool to assess both the thermal power and the fissile content of the whole nuclear core and could be used by the International Agency for Atomic Energy to enhance the safeguards of civil nuclear reactors. Deployed at only 7.2 m away from the compact Osiris research reactor core (70 MW) operating at the Saclay research center of the French Alternative Energies and Atomic Energy Commission, the experiment also exhibits a well-suited configuration to search for a new short baseline oscillation. We report the first results of the Nucifer experiment, describing the performances of the ˜0.85 m3 detector remotely operating at a shallow depth equivalent to ˜12 m of water and under intense background radiation conditions. Based on 145 (106) days of data with the reactor on (off), leading to the detection of an estimated 40760 ν¯ e , the mean number of detected antineutrinos is 281 ±7 (stat )±18 (syst )ν¯ e/day , in agreement with the prediction of 277 ±23 ν¯ e/day . Because of the large background, no conclusive results on the existence of light sterile neutrinos could be derived, however. As a first societal application we quantify how antineutrinos could be used for the Plutonium Management and Disposition Agreement.

Spatial Burnout in Water Reactors with Nonuniform Startup Distributions of Uranium and Boron

NASA Technical Reports Server (NTRS)

Fox, Thomas A.; Bogart, Donald

1955-01-01

Spatial burnout calculations have been made of two types of water moderated cylindrical reactor using boron as a burnable poison to increase reactor life. Specific reactors studied were a version of the Submarine Advanced Reactor (sAR) and a supercritical water reactor (SCW) . Burnout characteristics such as reactivity excursion, neutron-flux and heat-generation distributions, and uranium and boron distributions have been determined for core lives corresponding to a burnup of approximately 7 kilograms of fully enriched uranium. All reactivity calculations have been based on the actual nonuniform distribution of absorbers existing during intervals of core life. Spatial burnout of uranium and boron and spatial build-up of fission products and equilibrium xenon have been- considered. Calculations were performed on the NACA nuclear reactor simulator using two-group diff'usion theory. The following reactor burnout characteristics have been demonstrated: 1. A significantly lower excursion in reactivity during core life may be obtained by nonuniform rather than uniform startup distribution of uranium. Results for SCW with uranium distributed to provide constant radial heat generation and a core life corresponding to a uranium burnup of 7 kilograms indicated a maximum excursion in reactivity of 2.5 percent. This compared to a maximum excursion of 4.2 percent obtained for the same core life when w'anium was uniformly distributed at startup. Boron was incorporated uniformly in these cores at startup. 2. It is possible to approach constant radial heat generation during the life of a cylindrical core by means of startup nonuniform radial and axial distributions of uranium and boron. Results for SCW with nonuniform radial distribution of uranium to provide constant radial heat generation at startup and with boron for longevity indicate relatively small departures from the initially constant radial heat generation distribution during core life. Results for SAR with a sinusoidal distribution rather than uniform axial distributions of boron indicate significant improvements in axial heat generation distribution during the greater part of core life. 3. Uranium investments for cylindrical reactors with nonuniform radial uranium distributions which provide constant radial heat generation per unit core volume are somewhat higher than for reactors with uniform uranium concentration at startup. On the other hand, uranium investments for reactors with axial boron distributions which approach constant axial heat generation are somewhat smaller than for reactors with uniform boron distributions at startup.

Low Energy Neutrino Physics at the Kuo-Sheng Reactor Laboratory in Taiwan

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

Lin, S.-T.

2006-11-17

A laboratory has been constructed by the TEXONO Collaboration at the Kuo-Sheng Reactor Power Plant in Taiwan to study low energy neutrino physics. A limit on the neutrino magnetic moment of {mu}{nu}({nu}-bare) < 7.2 x 10-11 {mu}B at 90% confidence level has been achieved from measurements with a high-purity germanium detector, as well as the electron neutrinos ({nu}{sub e}) produced from nuclear power reactors has been studied. Other research program at Kuo-Sheng are surveyed.

Two Stage Anaerobic Reactor Design and Treatment To Produce Biogas From Mixed Liquor of Vegetable Waste

NASA Astrophysics Data System (ADS)

Budiastuti, H.; Ghozali, M.; Wicaksono, H. K.; Hadiansyah, R.

2018-01-01

Municipal solid waste has become a common challenged problem to be solved for developing countries including Indonesia. Municipal solid waste generating is always bigger than its treatment to reduce affect of environmental pollution. This research tries to contribute to provide an alternative solution to treat municipal solid waste to produce biogas. Vegetable waste was obtained from Gedebage Market, Bandung and starter as a source of anaerobic microorganisms was cow dung obtained from a cow farm in Lembang. A two stage anaerobic reactor was designed and built to treat the vegetable waste in a batch run. The capacity of each reactor is 20 liters but its active volume in each reactor is 15 liters. Reactor 1 (R1) was fed up with mixture of filtered blended vegetable waste and water at ratio of 1:1 whereas Reactor 2 (R2) was filled with filtered mixed liquor of cow dung and water at ratio of 1:1. Both mixtures were left overnight before use. Into R1 it was added EM-4 at concentration of 10%. pH in R1 was maintained at 5 - 6.5 whereas pH in R1 was maintained at 6.5 - 7.5. Temperature of reactors was not maintained to imitate the real environmental temperature. Parameters taken during experiment were pH, temperature, COD, MLVSS, and composition of biogas. The performance of reactor built was shown from COD efficiencies reduction obtained of about 60% both in R1 and R2, pH average in R1 of 4.5 ± 1 and R2 of 7 ± 0.6, average temperature in both reactors of 25 ± 2°C. About 1L gas produced was obtained during the last 6 days of experiment in which CH4 obtained was 8.951 ppm and CO2 of 1.087 ppm. The maximum increase of MLVSS in R1 reached 156% and R2 reached 89%.

Analysis of a boron-carbide-drum-controlled critical reactor experiment

NASA Technical Reports Server (NTRS)

Mayo, W. T.

1972-01-01

In order to validate methods and cross sections used in the neutronic design of compact fast-spectrum reactors for generating electric power in space, an analysis of a boron-carbide-drum-controlled critical reactor was made. For this reactor the transport analysis gave generally satisfactory results. The calculated multiplication factor for the most detailed calculation was only 0.7-percent Delta k too high. Calculated reactivity worth of the control drums was $11.61 compared to measurements of $11.58 by the inverse kinetics methods and $11.98 by the inverse counting method. Calculated radial and axial power distributions were in good agreement with experiment.

Global risk of radioactive fallout after major nuclear reactor accidents

NASA Astrophysics Data System (ADS)

Lelieveld, J.; Kunkel, D.; Lawrence, M. G.

2012-05-01

Major reactor accidents of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the cumulative, global risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents (the most severe ones on the International Nuclear Event Scale, INES 7), using particulate 137Cs and gaseous 131I as proxies for the fallout. Our results indicate that previously the occurrence of INES 7 major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a major reactor accident of any nuclear power plant worldwide, more than 90% of emitted 137Cs would be transported beyond 50 km and about 50% beyond 1000 km distance before being deposited. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human exposure due to deposition are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in West Europe and South Asia, where a major reactor accident can subject around 30 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

Organic matter degradation in a greywater recycling system using a multistage moving bed biofilm reactor (MBBR).

PubMed

Saidi, Assia; Masmoudi, Khaoula; Nolde, Erwin; El Amrani, Btissam; Amraoui, Fouad

2017-12-01

Greywater is an important non-conventional water resource which can be treated and recycled in buildings. A decentralized greywater recycling system for 223 inhabitants started operating in 2006 in Berlin, Germany. High load greywater undergoes advanced treatment in a multistage moving bed biofilm reactor (MBBR) followed by sand filtration and UV disinfection. The treated water is used safely as service water for toilet flushing. Monitoring of the organic matter degradation was pursued to describe the degradation processes in each stage and optimize the system. Results showed that organic matter reduction was achieved for the most part in the first three reactors, whereas the highest reduction rate was observed in the third reactor in terms of COD (chemical oxygen demand), dissolved organic carbon and BOD 7 (biological oxygen demand). The results also showed that the average loading rate entering the system was 3.7 kg COD/d, while the removal rate was 3.4 kg COD/d in a total bioreactor volume of 11.7 m³. In terms of BOD, the loading rate was 2.8 kg BOD/d and it was almost totally removed. This system requires little space (0.15 m²/person) and maintenance work of less than one hour per month and it shows operational stability under peak loads.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

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

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. Themore » objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.« less

Generation of the V4.2m5 and AMPX and MPACT 51 and 252-Group Libraries with ENDF/B-VII.0 and VII.1

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

Kim, Kang Seog

The evaluated nuclear data file (ENDF)/B-7.0 v4.1m3 MPACT 47-group library has been used as a main library for the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronics simulator in simulating pressurized water reactor (PWR) problems. Recent analysis for the high void boiling water reactor (BWR) fuels and burnt fuels indicates that the 47-group library introduces relatively large reactivity bias. Since the 47- group structure does not match with the SCALE 6.2 252-group boundaries, the CASL Virtual Environment for Reactor Applications Core Simulator (VERA-CS) MPACT library must be maintained independently, which causes quality assurance concerns. In order to addressmore » this issue, a new 51-group structure has been proposed based on the MPACT 47- g and SCALE 252-g structures. In addition, the new CASL library will include a 19-group structure for gamma production and interaction cross section data based on the SCALE 19- group structure. New AMPX and MPACT 51-group libraries have been developed with the ENDF/B-7.0 and 7.1 evaluated nuclear data. The 19-group gamma data also have been generated for future use, but they are only available on the AMPX 51-g library. In addition, ENDF/B-7.0 and 7.1 MPACT 252-g libraries have been generated for verification purposes. Various benchmark calculations have been performed to verify and validate the newly developed libraries.« less

Analog to digital converter system for temperature monitoring -- B, C, D, DR, F, and H reactors

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

Ballowe, J.W.

1961-03-23

This document discusses a proposal that certain presently installed reactor process water outlet temperature data logging equipment in subject reactors to be replaced with new functionally simplified equipment of a more adequate design. The primary purpose of the proposed installation is to replace existing equipment which is obsolete and in three reactors is worn out to the point where the equipment is out of service frequently for periods of time up to 8 hours or more. The new equipment will provide reliable process tube temperature information for use in the functions of reactor control and product accountability. Based upon anticipatedmore » incremental production gains resulting from use of the new equipment, the amortization period for the project is calculated at 2.7 years.« less

NERVA materials development

NASA Technical Reports Server (NTRS)

Mandell, B.

1970-01-01

Materials development topics include: development of analysis techniques to adjust heterogeneous data; determination of thermal conductivity for AISI 347 stainless steel and elastic moduli and Poisson's ratio for Inconel 718 and Ti 5Al-2.5Sn; embrittlement effects of 1400 psi gaseous hydrogen for alloy 718 and Ti 5Al-2.5Sn; cryogenic radiation damage of Ti 5Al-2.5Sn; and evaluation of prepreg, impregnation, and fabric materials for optimum fibrous graphite properties. Component support topics include: tensile design allowable development of Ti 5Al-2.5Sn for turbopump applications; evaluation of fatigue, fracture toughness, and stress corrosion properties of AA 7039-T63 for pressure vessel applications; development of AISI 347 sheet tensile and creep properties for nozzle applications; evaluation of orbital weld techniques for aluminum line fabrication; material selection of shield materials; development of high load friction and wear properties of hard chrome/gold plate combinations; and evaluation of weld processes for NASS duct coolant channel fabrication.

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

NASA Astrophysics Data System (ADS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.

Modeling of a Reaction-Distillation-Recycle System to Produce Dimethyl Ether through Methanol Dehydration

NASA Astrophysics Data System (ADS)

Muharam, Y.; Zulkarnain, L. M.; Wirya, A. S.

2018-03-01

The increase in the dimethyl ether yield through methanol dehydration due to a recycle integration to a reaction-distillation system was studied in this research. A one-dimensional phenomenological model of a methanol dehydration reactor and a shortcut model of distillation columns were used to achieve the aim. Simulation results show that 10.7 moles/s of dimethyl ether is produced in a reaction-distillation system with the reactor length being 4 m, the reactor inlet pressure being 18 atm, the reactor inlet temperature being 533 K, the reactor inlet velocity being 0.408 m/s, and the distillation pressure being 8 atm. The methanol conversion is 90% and the dimethyl ether yield is 48%. The integration of the recycle stream to the system increases the dimethyl ether yield by 8%.

Organic loading rate effect on the acidogenesis of cheese whey: a comparison between UASB and SBR reactors.

PubMed

Calero, R; Iglesias-Iglesias, R; Kennes, C; Veiga, M C

2017-09-16

Volatile fatty acids (VFA) production and degree of acidification (DA) were investigated in the anaerobic treatment of cheese whey by comparison of two processes: a continuous process using a laboratory upflow anaerobic sludge blanket (UASB) reactor and a discontinuous process using a sequencing batch reactor (SBR). The main purpose of this work was to study the organic loading rate (OLR) effect on the yield of VFA in two kinds of reactors. The predominant products in the acidogenic process in both reactors were: acetate, propionate, butyrate and valerate. The maximum DA obtained was 98% in an SBR at OLR of 2.7 g COD L -1 d -1 , and 97% in the UASB at OLR at 15.1 g COD L -1 d -1 . The results revealed that the UASB reactor was more efficient at a medium OLR with a higher VFA yield, while with the SBR reactor, the maximum acidification was obtained at a lower OLR with changes in the VFA profile at different OLRs applied.

METHOD OF CONTROLLING CORROSION IN A NEUTRONIC REACTOR

DOEpatents

Kidder, C.P.; Sloan, C.K.

1959-10-01

A method is described for reducing or removing corrosion and iron deposits on aluminum surfaces from coolant water comprising adding to the coolant alkali metal dichromate in a concentration of between 1.8 and 2.2 ppm, adjusting the pH to between 7.3 and 7.8 by adding CaCO/sub 3/ or other similar material, and adding a silicious material such as diatomaceous earth of a particle size of 5 to 15 microns to effect a suspension of between 2 and 300 ppm and circulating it through the reactor.

Enzyme-Cascade Analysis of the Rio Tinto Subsurface Environment: A Biosensor Experiment

NASA Technical Reports Server (NTRS)

McKay, David S.; Lynch, Kennda; Wainwright, Norman; Child, Alice; Williams, Kendra; McKay, David; Amils, Ricardo; Gonzalez, Elena; Stoker, Carol

2004-01-01

The Portable Test System (PTS), designed & developed by Charles Rivers Laboratories, Inc. (Charleston, SC) is a portable instrument that was designed to perform analysis of enzymatic assays related to rapid assessment of microbial contamination (Wainwright, 2003). The enzymatic cascade of Limulus Amebocyte Lysate (LAL) is known to be one of the most sensitive techniques available for microbial detection, enabling the PTS to be evaluated as a potential life detection instrument for in situ Astrobiology missions. In the summer of 2003 the system was tested as a part of the Mars Astrobiology Research and Technology Experiment (MARTE) ground truth science campaign in the Rio Tinto Analogue environment near Nerva, Spain. The preliminary results show that the PTS analysis correlates well with the contamination control tests and the more traditional lab-based biological assays performed during the MARTE field mission. Further work will be conducted on this research during a second field campaign in 2004 and a technology demonstration of a prototype instrument that includes autonomous sample preparation will occur in 2005.

Micropollutant removal from black water and grey water sludge in a UASB-GAC reactor.

PubMed

Butkovskyi, A; Sevenou, L; Meulepas, R J W; Hernandez Leal, L; Zeeman, G; Rijnaarts, H H M

2018-02-01

The effect of granular activated carbon (GAC) addition on the removal of diclofenac, ibuprofen, metoprolol, galaxolide and triclosan in a up-flow anaerobic sludge blanket (UASB) reactor was studied. Prior to the reactor studies, batch experiments indicated that addition of activated carbon to UASB sludge can decrease micropollutant concentrations in both liquid phase and sludge. In continuous experiments, two UASB reactors were operated for 260 days at an HRT of 20 days, using a mixture of source separated black water and sludge from aerobic grey water treatment as influent. GAC (5.7 g per liter of reactor volume) was added to one of the reactors on day 138. No significant difference in COD removal and biogas production between reactors with and without GAC addition was observed. In the presence of GAC, fewer micropollutants were washed out with the effluent and a lower accumulation of micropollutants in sludge and particulate organic matter occurred, which is an advantage in micropollutant emission reduction from wastewater. However, the removal of micropollutants by adding GAC to a UASB reactor would require more activated carbon compared to effluent post-treatment. Additional research is needed to estimate the effect of bioregeneration on the lifetime of activated carbon in a UASB-GAC reactor.

A study on the use of the BioBall® as a biofilm carrier in a sequencing batch reactor.

PubMed

Masłoń, Adam; Tomaszek, Janusz A

2015-11-01

Described in this study are experiments conducted to evaluate the removal of organics and nutrients from synthetic wastewater by a moving bed sequencing batch biofilm reactor using BioBall® carriers as biofilm media. The work involving a 15L-laboratory scale MBSBBR (moving bed sequencing batch biofilm reactor) model showed that the wastewater treatment system was based on biochemical processes taking place with activated sludge and biofilm microorganisms developing on the surface of the BioBall® carriers. Classical nitrification and denitrification and the typical enhanced biological phosphorus removal process were achieved in the reactor analyzed, which operated with a volumetric organic loading of 0.84-0.978gCODL(-1)d(-1). The average removal efficiencies for COD, total nitrogen and total phosphorus were found to be 97.7±0.5%, 87.8±2.6% and 94.3±1.3%, respectively. Nitrification efficiency reached levels in the range 96.5-99.7%. Copyright © 2015 Elsevier Ltd. All rights reserved.

Thermal Hydraulics Design and Analysis Methodology for a Solid-Core Nuclear Thermal Rocket Engine Thrust Chamber

NASA Technical Reports Server (NTRS)

Wang, Ten-See; Canabal, Francisco; Chen, Yen-Sen; Cheng, Gary; Ito, Yasushi

2013-01-01

Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions. This chapter describes a thermal hydraulics design and analysis methodology developed at the NASA Marshall Space Flight Center, in support of the nuclear thermal propulsion development effort. The objective of this campaign is to bridge the design methods in the Rover/NERVA era, with a modern computational fluid dynamics and heat transfer methodology, to predict thermal, fluid, and hydrogen environments of a hypothetical solid-core, nuclear thermal engine the Small Engine, designed in the 1960s. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics and heat transfer platform, while formulations of flow and heat transfer through porous and solid media were implemented to describe those of hydrogen flow channels inside the solid24 core. Design analyses of a single flow element and the entire solid-core thrust chamber of the Small Engine were performed and the results are presented herein

Premium quality 5A1-2.5 Sn ELI titanium production

NASA Technical Reports Server (NTRS)

Dessau, P. P.; Harris, C. L.

1972-01-01

Preliminary design and reliability analysis conducted on the turbopump for the NERVA 75,000 full flow cycle engine, indicated that the turbopump bearings were the most critical turbopump parts in meeting the 10 hour life at the required turbopump reliability of .99978. The analysis revealed that significant reductions (approximately a factor of 3.25) in bearing loads would be achieved by fabricating the rotating parts from titanium in lieu of A286 or 718. This is basically due to the difference in density of the materials and the resulting mass effect on the location of the first and second stick mode critical speeds. For the selected rotor configuration, the lighter material has a first critical speed at approximately 36,000 rpm, while that of the heavier material has a first critical at approximately 27,000 rpm. As the operating range of the turbopump is from 0 to 30,000 rpm, the heavier material would have a stick mode critical in the operating range.

Effect of organic matter strength on anammox for modified greenhouse turtle breeding wastewater treatment.

PubMed

Chen, Chongjun; Huang, Xiaoxiao; Lei, Chenxiao; Zhang, Tian C; Wu, Weixiang

2013-11-01

Anaerobic ammonium-N removal from modified greenhouse turtle breeding wastewater with different chemical oxygen demand (COD) strengths (194.0-577.8 mg L(-1)) at relatively fixed C/N ratios (≈ 2) was investigated using a lab-scale up-flow anaerobic sludge blanket (UASB) anammox reactor. During the entire experiment, the total nitrogen (TN) removal efficiency was about 85% or higher, while the average COD removal efficiency was around 56.5 ± 7.9%. Based on the nitrogen and carbon balance, the nitrogen removal contribution was 79.6 ± 4.2% for anammox, 12.7 ± 3.0% for denitrification+denitritation and 7.7 ± 4.9% for other mechanisms. Denaturing gradient gel electrophoresis (DGGE) analyses revealed that Planctomycete, Proteobacteria and Chloroflexi bacteria were coexisted in the reactor. Anammox was always dominant when the reactor was fed with different COD concentrations, which indicated the stability of the anammox process with the coexistence of the denitrification process in treating greenhouse turtle breeding wastewater. Copyright © 2013 Elsevier Ltd. All rights reserved.

Synfuels from fusion: using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

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

Werner, R.W.

1982-11-01

This study is concerned with the following area: (1) the tandem mirror reactor and its physics; (2) energy balance; (3) the lithium oxide canister blanket system; (4) high-temperature blanket; (5) energy transport system-reactor to process; (6) thermochemical hydrogen processes; (7) interfacing the GA cycle; (8) matching power and temperature demands; (9) preliminary cost estimates; (10) synfuels beyond hydrogen; and (11) thermodynamics of the H/sub 2/SO/sub 4/-H/sub 2/O system. (MOW)

The effect of mixing on fermentation of primary solids, glycerol, and biodiesel waste.

PubMed

Ghasemi, Marzieh; Randall, Andrew A

2018-03-01

In this study, the effect of mixing on volatile fatty acid (VFA) production and composition was investigated through running five identical bench-scale reactors that were filled with primary solid and dosed with either pure glycerol or biodiesel waste. Experimental results revealed that there was an inverse correlation between the mixing intensity and the VFA production. The total VFA production in the un-mixed reactor was 9,787 ± 3,601 mg COD/L, whereas in the reactor mixed at 100 rpm this dropped to 3,927 ± 1,175 mg COD/L, while both types of reactor were dosed with pure glycerol at the beginning of each cycle to reach the initial concentration of 1,000 mg/L (1,217 mg COD/L). Propionic acid was the dominant VFA in all the reactors except the reactor mixed at 30 rpm. It is hypothesized that low mixing facilitated hydrogen transfer between obligate hydrogen producing acetogens (OHPA) and hydrogen consuming acidogens in these non-methanogenic reactors. Also, in a narrower range of mixing (0 or 7 rpm), the total VFA production in biodiesel waste-fed reactors was considerably higher than that of pure glycerol-fed reactors.

Pilot-Scale Selenium Bioremediation of San Joaquin Drainage Water with Thauera selenatis

PubMed Central

Cantafio, A. W.; Hagen, K. D.; Lewis, G. E.; Bledsoe, T. L.; Nunan, K. M.; Macy, J. M.

1996-01-01

This report describes a simple method for the bioremediation of selenium from agricultural drainage water. A medium-packed pilot-scale biological reactor system, inoculated with the selenate-respiring bacterium Thauera selenatis, was constructed at the Panoche Water District, San Joaquin Valley, Calif. The reactor was used to treat drainage water (7.6 liters/min) containing both selenium and nitrate. Acetate (5 mM) was the carbon source-electron donor reactor feed. Selenium oxyanion concentrations (selenate plus selenite) in the drainage water were reduced by 98%, to an average of 12 (plusmn) 9 (mu)g/liter. Frequently (47% of the sampling days), reactor effluent concentrations of less than 5 (mu)g/liter were achieved. Denitrification was also observed in this system; nitrate and nitrite concentrations in the drainage water were reduced to 0.1 and 0.01 mM, respectively (98% reduction). Analysis of the reactor effluent showed that 91 to 96% of the total selenium recovered was elemental selenium; 97.9% of this elemental selenium could be removed with Nalmet 8072, a new, commercially available precipitant-coagulant. Widespread use of this system (in the Grasslands Water District) could reduce the amount of selenium deposited in the San Joaquin River from 7,000 to 140 lb (ca. 3,000 to 60 kg)/year. PMID:16535401

Kinetic study on the effect of temperature on biogas production using a lab scale batch reactor.

PubMed

Deepanraj, B; Sivasubramanian, V; Jayaraj, S

2015-11-01

In the present study, biogas production from food waste through anaerobic digestion was carried out in a 2l laboratory-scale batch reactor operating at different temperatures with a hydraulic retention time of 30 days. The reactors were operated with a solid concentration of 7.5% of total solids and pH 7. The food wastes used in this experiment were subjected to characterization studies before and after digestion. Modified Gompertz model and Logistic model were used for kinetic study of biogas production. The kinetic parameters, biogas yield potential of the substrate (B), the maximum biogas production rate (Rb) and the duration of lag phase (λ), coefficient of determination (R(2)) and root mean square error (RMSE) were estimated in each case. The effect of temperature on biogas production was evaluated experimentally and compared with the results of kinetic study. The results demonstrated that the reactor with operating temperature of 50°C achieved maximum cumulative biogas production of 7556ml with better biodegradation efficiency. Copyright © 2015 Elsevier Inc. All rights reserved.

ETR CRITICAL FACILITY, TRA654. SCIENTISTS STAND AT EDGE OF TANK ...

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

ETR CRITICAL FACILITY, TRA-654. SCIENTISTS STAND AT EDGE OF TANK AND LIFT REMOVABLE BRIDGE ABOVE THE REACTOR. CONTROL RODS AND FUEL RODS ARE BELOW ENOUGH WATER TO SHIELD WORKERS ABOVE. NOTE CRANE RAILS ALONG WALLS, PUMICE BLOCK WALLS. INL NEGATIVE NO. 57-3690. R.G. Larsen, Photographer, 7/29/1957 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Real-time LMR control parameter generation using advanced adaptive synthesis

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

King, R.W.; Mott, J.E.

1990-01-01

The reactor delta T'', the difference between the average core inlet and outlet temperatures, for the liquid-sodium-cooled Experimental Breeder Reactor 2 is empirically synthesized in real time from, a multitude of examples of past reactor operation. The real-time empirical synthesis is based on reactor operation. The real-time empirical synthesis is based on system state analysis (SSA) technology embodied in software on the EBR 2 data acquisition computer. Before the real-time system is put into operation, a selection of reactor plant measurements is made which is predictable over long periods encompassing plant shutdowns, core reconfigurations, core load changes, and plant startups.more » A serial data link to a personal computer containing SSA software allows the rapid verification of the predictability of these plant measurements via graphical means. After the selection is made, the real-time synthesis provides a fault-tolerant estimate of the reactor delta T accurate to {plus}/{minus}1{percent}. 5 refs., 7 figs.« less

Dynamic biogas upgrading based on the Sabatier process: thermodynamic and dynamic process simulation.

PubMed

Jürgensen, Lars; Ehimen, Ehiaze Augustine; Born, Jens; Holm-Nielsen, Jens Bo

2015-02-01

This study aimed to investigate the feasibility of substitute natural gas (SNG) generation using biogas from anaerobic digestion and hydrogen from renewable energy systems. Using thermodynamic equilibrium analysis, kinetic reactor modeling and transient simulation, an integrated approach for the operation of a biogas-based Sabatier process was put forward, which was then verified using a lab scale heterogenous methanation reactor. The process simulation using a kinetic reactor model demonstrated the feasibility of the production of SNG at gas grid standards using a single reactor setup. The Wobbe index, CO2 content and calorific value were found to be controllable by the H2/CO2 ratio fed the methanation reactor. An optimal H2/CO2 ratio of 3.45-3.7 was seen to result in a product gas with high calorific value and Wobbe index. The dynamic reactor simulation verified that the process start-up was feasible within several minutes to facilitate surplus electricity use from renewable energy systems. Copyright © 2014 Elsevier Ltd. All rights reserved.

Removal of sulphates acidity and iron from acid mine drainage in a bench scale biochemical treatment system.

PubMed

Prasad, D; Henry, J G

2009-02-01

The focus of this study was to develop a simple biochemical system to treat acid mine drainage for its safe disposal. Recovery and reuse of the metals removed were not considered. A three-step process for the treatment of acid mine drainage (AMD), proposed earlier, separates sulphate reducing activity from metal precipitation units and from a pH control system. Following our earlier work on the first step (biological reactor), this paper examines the second step (i.e. chemical reactor). The objectives of this study were: (1) to determine the increase in pH and the reduction of iron in the chemical reactor for different proportions of simulated AMD, and (2) to assess the capability of the chemical reactor. A series of experiments was conducted to study the effects of addition of alkaline sulphidogenic liquor (ASL) derived from a batch sulphidogenic biological reactor (operating with activated sludge and a COD/SO4 ratio of 1.6) on the simulated AMD characteristics. At 60-minute contact time, addition of 30% ASL (pH of 7.60-7.76) to the chemical reactor with 70% AMD (pH of 1.65-2.02), increased the pH of the AMD to 6.57 and alkalinity from 0 to 485 mg l(-1) as CaCO3, respectively and precipitated about 97% of the iron present in the simulated AMD. Others have demonstrated that metals in mine drainage can be precipitated by bacterial sulphate reduction. In this study, iron, a common and major component of mine drainage was used as a surrogate for metals in general. The results indicate the feasibility of treating AMD by an engineered sulphidogenic anaerobic reactor followed by a chemical reactor and that our three-step biochemical process has important advantages over other conventional AMD treatment systems.

Bioelectrochemical enhancement of methane production from highly concentrated food waste in a combined anaerobic digester and microbial electrolysis cell.

PubMed

Park, Jungyu; Lee, Beom; Tian, Donjie; Jun, Hangbae

2018-01-01

A microbial electrolysis cell (MEC) is a promising technology for enhancing biogas production from an anaerobic digestion (AD) reactor. In this study, the effects of the MEC on the rate of methane production from food waste were examined by comparing an AD reactor with an AD reactor combined with a MEC (AD+MEC). The use of the MEC accelerated methane production and stabilization via rapid organic oxidation and rapid methanogenesis. Over the total experimental period, the methane production rate and stabilization time of the AD+MEC reactor were approximately 1.7 and 4.0 times faster than those of the AD reactor. Interestingly however, at the final steady state, the methane yields of both the reactors were similar to the theoretical maximum methane yield. Based on these results, the MEC did not increase the methane yield over the theoretical value, but accelerated methane production and stabilization by bioelectrochemical reactions. Copyright © 2017 Elsevier Ltd. All rights reserved.

Effect of temperature on selenium removal from wastewater by UASB reactors.

PubMed

Dessì, Paolo; Jain, Rohan; Singh, Satyendra; Seder-Colomina, Marina; van Hullebusch, Eric D; Rene, Eldon R; Ahammad, Shaikh Ziauddin; Carucci, Alessandra; Lens, Piet N L

2016-05-01

The effect of temperature on selenium (Se) removal by upflow anaerobic sludge blanket (UASB) reactors treating selenate and nitrate containing wastewater was investigated by comparing the performance of a thermophilic (55 °C) versus a mesophilic (30 °C) UASB reactor. When only selenate (50 μM) was fed to the UASB reactors (pH 7.3; hydraulic retention time 8 h) with excess electron donor (lactate at 1.38 mM corresponding to an organic loading rate of 0.5 g COD L(-1) d(-1)), the thermophilic UASB reactor achieved a higher total Se removal efficiency (94.4 ± 2.4%) than the mesophilic UASB reactor (82.0 ± 3.8%). When 5000 μM nitrate was further added to the influent, total Se removal was again better under thermophilic (70.1 ± 6.6%) when compared to mesophilic (43.6 ± 8.8%) conditions. The higher total effluent Se concentration in the mesophilic UASB reactor was due to the higher concentrations of biogenic elemental Se nanoparticles (BioSeNPs). The shape of the BioSeNPs observed in both UASB reactors was different: nanospheres and nanorods, respectively, in the mesophilic and thermophilic UASB reactors. Microbial community analysis showed the presence of selenate respirers as well as denitrifying microorganisms. Copyright © 2016 Elsevier Ltd. All rights reserved.

Anaerobic/aerobic treatment of a petrochemical wastewater from two aromatic transformation processes by fluidized bed reactors.

PubMed

Estrada-Arriaga, Edson B; Ramirez-Camperos, Esperanza; Moeller-Chavez, Gabriela E; García-Sanchez, Liliana

2012-01-01

An integrated fluidized bed reactor (FBR) has been employed as the treatment for petrochemical industry wastewaters with high organic matter and aromatic compounds, under anaerobic and aerobic conditions. The system was operated at hydraulic residence time (HRT) of 2.7 and 2.2 h in the anaerobic and aerobic reactor, respectively. The degree of fluidization in the beds was 30%. This system showed a high performance on the removal of organic matter and aromatic compounds. At different organic loading rates (OLR), the chemical oxygen demand (COD) removal in the anaerobic reactor was close to 85% and removals of the COD up to 94% were obtained in the aerobic reactor. High removals of benzene, toluene, ethylbenzene, xylenes, styrene, 1,2,4-trimethylbenzene, 1,3,5-trimethylbenzene and naphthalene were achieved in this study.

Modeling residence-time distribution in horizontal screw hydrolysis reactors

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

Sievers, David A.; Stickel, Jonathan J.

The dilute-acid thermochemical hydrolysis step used in the production of liquid fuels from lignocellulosic biomass requires precise residence-time control to achieve high monomeric sugar yields. Difficulty has been encountered reproducing residence times and yields when small batch reaction conditions are scaled up to larger pilot-scale horizontal auger-tube type continuous reactors. A commonly used naive model estimated residence times of 6.2-16.7 min, but measured mean times were actually 1.4-2.2 the estimates. Here, this study investigated how reactor residence-time distribution (RTD) is affected by reactor characteristics and operational conditions, and developed a method to accurately predict the RTD based on key parameters.more » Screw speed, reactor physical dimensions, throughput rate, and process material density were identified as major factors affecting both the mean and standard deviation of RTDs. The general shape of RTDs was consistent with a constant value determined for skewness. The Peclet number quantified reactor plug-flow performance, which ranged between 20 and 357.« less

Modeling residence-time distribution in horizontal screw hydrolysis reactors

DOE PAGES

Sievers, David A.; Stickel, Jonathan J.

2017-10-12

The dilute-acid thermochemical hydrolysis step used in the production of liquid fuels from lignocellulosic biomass requires precise residence-time control to achieve high monomeric sugar yields. Difficulty has been encountered reproducing residence times and yields when small batch reaction conditions are scaled up to larger pilot-scale horizontal auger-tube type continuous reactors. A commonly used naive model estimated residence times of 6.2-16.7 min, but measured mean times were actually 1.4-2.2 the estimates. Here, this study investigated how reactor residence-time distribution (RTD) is affected by reactor characteristics and operational conditions, and developed a method to accurately predict the RTD based on key parameters.more » Screw speed, reactor physical dimensions, throughput rate, and process material density were identified as major factors affecting both the mean and standard deviation of RTDs. The general shape of RTDs was consistent with a constant value determined for skewness. The Peclet number quantified reactor plug-flow performance, which ranged between 20 and 357.« less

Steam reforming of heptane in a fluidized bed membrane reactor

NASA Astrophysics Data System (ADS)

Rakib, Mohammad A.; Grace, John R.; Lim, C. Jim; Elnashaie, Said S. E. H.

n-Heptane served as a model compound to study steam reforming of naphtha as an alternative feedstock to natural gas for production of pure hydrogen in a fluidized bed membrane reactor. Selective removal of hydrogen using Pd 77Ag 23 membrane panels shifted the equilibrium-limited reactions to greater conversion of the hydrocarbons and lower yields of methane, an intermediate product. Experiments were conducted with no membranes, with one membrane panel, and with six panels along the height of the reactor to understand the performance improvement due to hydrogen removal in a reactor where catalyst particles were fluidized. Results indicate that a fluidized bed membrane reactor (FBMR) can provide a compact reformer for pure hydrogen production from a liquid hydrocarbon feedstock at moderate temperatures (475-550 °C). Under the experimental conditions investigated, the maximum achieved yield of pure hydrogen was 14.7 moles of pure hydrogen per mole of heptane fed.

Isotopic Transmutations in Irradiated Beryllium and Their Implications on MARIA Reactor Operation

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

Andrzejewski, Krzysztof J.; Kulikowska, Teresa A

2004-04-15

Beryllium irradiated by neutrons with energies above 0.7 MeV undergoes (n,{alpha}) and (n,2n) reactions. The Be(n,{alpha}) reaction results in subsequent buildup of {sup 6}Li and {sup 3}He isotopes with large thermal neutron absorption cross sections causing poisoning of irradiated beryllium. The amount of the poison isotopes depends on the neutron flux level and spectrum. The high-flux MARIA reactor operated in Poland since 1975 consists of a beryllium matrix with fuel channels in cutouts of beryllium blocks. As the experimental determination of {sup 6}Li, {sup 3}H, and {sup 3}He content in the operational reactor is impossible, a systematic computational study ofmore » the effect of {sup 3}He and {sup 6}Li presence in beryllium blocks on MARIA reactor reactivity and power density distribution has been undertaken. The analysis of equations governing the transmutation has been done for neutron flux parameters typical for MARIA beryllium blocks. Study of the mutual influence of reactor operational parameters and the buildup of {sup 6}Li, {sup 3}H, and {sup 3}He in beryllium blocks has shown the necessity of a detailed spatial solution of transmutation equations in the reactor, taking into account the whole history of its operation. Therefore, fuel management calculations using the REBUS code with included chains for Be(n,{alpha})-initiated reactions have been done for the whole reactor lifetime. The calculated poisoning of beryllium blocks has been verified against the critical experiment of 1993. Finally, the current {sup 6}Li, {sup 3}H, and {sup 3}He contents, averaged for each beryllium block, have been calculated. The reactivity drop caused by this poisoning is {approx}7%.« less

Changes in the Structure and Function of Microbial Communities in Drinking Water Treatment Bioreactors upon Addition of Phosphorus▿ †

PubMed Central

Li, Xu; Upadhyaya, Giridhar; Yuen, Wangki; Brown, Jess; Morgenroth, Eberhard; Raskin, Lutgarde

2010-01-01

Phosphorus was added as a nutrient to bench-scale and pilot-scale biologically active carbon (BAC) reactors operated for perchlorate and nitrate removal from contaminated groundwater. The two bioreactors responded similarly to phosphorus addition in terms of microbial community function (i.e., reactor performance), while drastically different responses in microbial community structure were detected. Improvement in reactor performance with respect to perchlorate and nitrate removal started within a few days after phosphorus addition for both reactors. Microbial community structures were evaluated using molecular techniques targeting 16S rRNA genes. Clone library results showed that the relative abundance of perchlorate-reducing bacteria (PRB) Dechloromonas and Azospira in the bench-scale reactor increased from 15.2% and 0.6% to 54.2% and 11.7% after phosphorus addition, respectively. Real-time quantitative PCR (qPCR) experiments revealed that these increases started within a few days after phosphorus addition. In contrast, after phosphorus addition, the relative abundance of Dechloromonas in the pilot-scale reactor decreased from 7.1 to 0.6%, while Zoogloea increased from 17.9 to 52.0%. The results of this study demonstrated that similar operating conditions for bench-scale and pilot-scale reactors resulted in similar contaminant removal performances, despite dramatically different responses from microbial communities. These findings suggest that it is important to evaluate the microbial community compositions inside bioreactors used for drinking water treatment, as they determine the microbial composition in the effluent and impact downstream treatment requirements for drinking water production. This information could be particularly relevant to drinking water safety, if pathogens or disinfectant-resistant bacteria are detected in the bioreactors. PMID:20889793

75 FR 51499 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-20

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal Hydraulics Phenomena The ACRS Subcommittee on Thermal Hydraulics Phenomena will hold a meeting on September 7, 2010, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The entire meeting...

PBF Reactor Building (PER620). Detail of arrangement of highdensity blocks ...

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

PBF Reactor Building (PER-620). Detail of arrangement of high-density blocks and other basement shielding. Date: February 1966. Ebasco Services 1205 PER/PBF 620-A-7. INEEL index no. 761-0620-00-205-123070 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Optimization of the moving-bed biofilm sequencing batch reactor (MBSBR) to control aeration time by kinetic computational modeling: Simulated sugar-industry wastewater treatment.

PubMed

Faridnasr, Maryam; Ghanbari, Bastam; Sassani, Ardavan

2016-05-01

A novel approach was applied for optimization of a moving-bed biofilm sequencing batch reactor (MBSBR) to treat sugar-industry wastewater (BOD5=500-2500 and COD=750-3750 mg/L) at 2-4 h of cycle time (CT). Although the experimental data showed that MBSBR reached high BOD5 and COD removal performances, it failed to achieve the standard limits at the mentioned CTs. Thus, optimization of the reactor was rendered by kinetic computational modeling and using statistical error indicator normalized root mean square error (NRMSE). The results of NRMSE revealed that Stover-Kincannon (error=6.40%) and Grau (error=6.15%) models provide better fits to the experimental data and may be used for CT optimization in the reactor. The models predicted required CTs of 4.5, 6.5, 7 and 7.5 h for effluent standardization of 500, 1000, 1500 and 2500 mg/L influent BOD5 concentrations, respectively. Similar pattern of the experimental data also confirmed these findings. Copyright © 2016 Elsevier Ltd. All rights reserved.

Design of the Sandia-Israel 20-kW reflux heat-pipe solar receiver/reactor

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

Diver, R.B.; Ginn, W.C.

1987-09-01

This report describes the design and fabrication of a 20-kW sodium reflux heat-pipe solar receiver/reactor for CO/sub 2/ reforming of methane. This project is part of a bilateral agreement between the United States and Israel. Under the terms of the agreement the solar receiver/reactor has been designed and built by Sandia National Laboratories for testing in the 7-meter solar furnace facility at the Weizmann Institute of Science in Rehovot, Israel. 16 refs., 11 figs., 2 tabs.

Thermal energy storage material thermophysical property measurement and heat transfer impact

NASA Technical Reports Server (NTRS)

Tye, R. P.; Bourne, J. G.; Destarlais, A. O.

1976-01-01

The thermophysical properties of salts having potential for thermal energy storage to provide peaking energy in conventional electric utility power plants were investigated. The power plants studied were the pressurized water reactor, boiling water reactor, supercritical steam reactor, and high temperature gas reactor. The salts considered were LiNO3, 63LiOH/37 LiCl eutectic, LiOH, and Na2B4O7. The thermal conductivity, specific heat (including latent heat of fusion), and density of each salt were measured for a temperature range of at least + or - 100 K of the measured melting point. Measurements were made with both reagent and commercial grades of each salt.

Application of a fluidized bed reactor charged with aragonite for control of alkalinity, pH and carbon dioxide in marine recirculating aquaculture systems

USGS Publications Warehouse

Paul S Wills, PhD; Pfeiffer, Timothy; Baptiste, Richard; Watten, Barnaby J.

2016-01-01

Control of alkalinity, dissolved carbon dioxide (dCO2), and pH are critical in marine recirculating aquaculture systems (RAS) in order to maintain health and maximize growth. A small-scale prototype aragonite sand filled fluidized bed reactor was tested under varying conditions of alkalinity and dCO2 to develop and model the response of dCO2 across the reactor. A large-scale reactor was then incorporated into an operating marine recirculating aquaculture system to observe the reactor as the system moved toward equilibrium. The relationship between alkalinity dCO2, and pH across the reactor are described by multiple regression equations. The change in dCO2 across the small-scale reactor indicated a strong likelihood that an equilibrium alkalinity would be maintained by using a fluidized bed aragonite reactor. The large-scale reactor verified this observation and established equilibrium at an alkalinity of approximately 135 mg/L as CaCO3, dCO2 of 9 mg/L, and a pH of 7.0 within 4 days that was stable during a 14 day test period. The fluidized bed aragonite reactor has the potential to simplify alkalinity and pH control, and aid in dCO2 control in RAS design and operation. Aragonite sand, purchased in bulk, is less expensive than sodium bicarbonate and could reduce overall operating production costs.

Co-digestion performance of organic fraction of municipal solid waste with leachate: Preliminary studies.

PubMed

Guven, Huseyin; Akca, Mehmet Sadik; Iren, Erol; Keles, Fatih; Ozturk, Izzet; Altinbas, Mahmut

2018-01-01

The main aim of the study was to evaluate the co-digestion performance of OFMSW with different wastes. Leachate, reverse osmosis (RO) concentrate collected from a leachate treatment facility and dewatered sewage sludge taken from a wastewater treatment plant (WWTP) were used for co-digestion in this paper. An extra effort was made to observe the effect of leachate inclusion in the co-digestion. In the study, the mono-digestion of OFMSW, leachate, RO concentrate and sewage sludge as well as digestion of 7 different waste mixtures were carried out for this objective. The experiments were carried out for approximately 50days under mesophilic conditions. The highest methane yield was 785L CH 4 /kg VS added in the reactor, which had only OFMSW. While the methane yield derived from OFMSW was found higher than previous studies, methane yield of leachate was found to be 110L CH 4 /kg VS added , which was lower than findings in the literature. The mono-substrate of OFMSW was followed by the reactor of having waste mixture of leachate+sewage sludge+OFMSW+water (C7) with 391L CH 4 /kg VS added , which was the only combination included water. In order to understand the effect of leachate and water inclusions on co-digestion, two separate waste combinations; leachate+sewage sludge+OFMSW+water (C7) and leachate+sewage sludge+OFMSW (C1) were prepared that had different amounts of leachate but same amounts of other wastes. The methane yield of leachate+sewage sludge+OFMSW+water (C7) indicated that addition of some water instead of leachate could stimulate biogas production. Methane yield of this reactor was found to be 71% higher than the waste combination of leachate+sewage sludge+OFMSW (C1). It could be thought that the high amount of non-biodegradable matters in leachate could be responsible for lower methane yield in leachate+sewage sludge+OFMSW (C1) reactor. Methane yields of the reactors showed that co-digestion of OFMSW and leachate could be a solution not only for treatment of leachate and but also increasing the biogas potential of leachate. Leachate addition could also adjust optimum total solids (TS) content in anaerobic digestion. It was also understood that RO concentrate did not affect the methane yield in a negative way. The similar characterization of leachate and RO concentrate in this study could offer the utilization of RO concentrate instead of leachate. The findings showed that volatile solids (VS) removals were changed from 32% to 61% in the reactors. While the reactor of leachate+RO concentrate+OFMSW (C6) had the highest VS removal, the reactor of the sole substrate leachate had the lowest VS removal. Copyright © 2017 Elsevier Ltd. All rights reserved.

Antibiotic Fermentation Broth Treatment by a pilot upflow anaerobic sludge bed reactor and kinetic modeling.

PubMed

Coskun, T; Kabuk, H A; Varinca, K B; Debik, E; Durak, I; Kavurt, C

2012-10-01

In this study, an upflow anaerobic sludge blanket (UASB) mesophilic reactor was used to remove antibiotic fermentation broth wastewater. The hydraulic retention time was held constant at 13.3 days. The volumetric organic loading value increased from 0.33 to 7.43 kg(COD)m(-3)d(-1) using antibiotic fermentation broth wastewater gradually diluted with various ratios of domestic wastewater. A COD removal efficiency of 95.7% was obtained with a maximum yield of 3,700 L d(-1) methane gas production. The results of the study were interpreted using the modified Stover-Kincannon, first-order, substrate mass balance and Van der Meer and Heertjes kinetic models. The obtained kinetic coefficients showed that antibiotic fermentation broth wastewater can be successfully treated using a UASB reactor while taking COD removal and methane production into account. Copyright © 2012 Elsevier Ltd. All rights reserved.

Susceptibility constants of airborne bacteria to dielectric barrier discharge for antibacterial performance evaluation.

PubMed

Park, Chul Woo; Hwang, Jungho

2013-01-15

Dielectric barrier discharge (DBD) is a promising method to remove contaminant bioaerosols. The collection efficiency of a DBD reactor is an important factor for determining a reactor's removal efficiency. Without considering collection, simply defining the inactivation efficiency based on colony counting numbers for DBD as on and off may lead to overestimation of the inactivation efficiency of the DBD reactor. One-pass removal tests of bioaerosols were carried out to deduce the inactivation efficiency of the DBD reactor using both aerosol- and colony-counting methods. Our DBD reactor showed good performance for removing test bioaerosols for an applied voltage of 7.5 kV and a residence time of 0.24s, with η(CFU), η(Number), and η(Inactivation) values of 94%, 64%, and 83%, respectively. Additionally, we introduce the susceptibility constant of bioaerosols to DBD as a quantitative parameter for the performance evaluation of a DBD reactor. The modified susceptibility constant, which is the ratio of the susceptibility constant to the volume of the plasma reactor, has been successfully demonstrated for the performance evaluation of different sized DBD reactors under different DBD operating conditions. Our methodology will be used for design optimization, performance evaluation, and prediction of power consumption of DBD for industrial applications. Copyright © 2012 Elsevier B.V. All rights reserved.

Deciphering the science behind electrocoagulation to remove suspended clay particles from water.

PubMed

Holt, P K; Barton, G W; Mitchell, C A

2004-01-01

Electrocoagulation removes pollutant material from water by a combination of coagulant delivered from a sacrificial aluminium anode and hydrogen bubbles evolved at an inert cathode. Rates of clay particle flotation and settling were experimentally determined in a 7 L batch reactor over a range of currents (0.25-2.0 A) and pollutant loadings (0.1-1.7 g/L). Sedimentation and flotation are the dominant removal mechanism at low and high currents, respectively. This shift in separation mode can be explained by analysing the reactor in terms of a published dissolved air flotation model.

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

PubMed

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

2012-08-01

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

PROCESS WATER BUILDING, TRA605. AERIAL TAKEN WHILE SEVERAL PIPE TRENCHES ...

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

PROCESS WATER BUILDING, TRA-605. AERIAL TAKEN WHILE SEVERAL PIPE TRENCHES REMAINED OPEN. CAMERA FACES EASTERLY. NOTE DUAL PIPES BETWEEN REACTOR BUILDING AND NORTH SIDE OF PROCESS WATER BUILDING. PIPING NEAR WORKING RESERVOIR HEADS FOR RETENTION RESERVOIR. PIPE FROM DEMINERALIZER ENTERS MTR FROM NORTH. SEE ALSO TRENCH FOR COOLANT AIR DUCT AT SOUTH SIDE OF MTR AND LEADING TO FAN HOUSE AND STACK. INL NEGATIVE NO. 2966-A. Unknown Photographer, 7/31/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

High-sensitivity fast neutron detector KNK-2-7M

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

Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Dovbysh, L. Ye.; Ovchinnikov, M. A.

2015-12-15

The construction of the fast neutron detector KNK-2-7M is briefly described. The results of the study of the detector in the pulse-counting mode are given for the fissions of {sup 237}Np nuclei in the radiator of the neutron-sensitive section and in the current mode with the separation of sectional currents of functional sections. The possibilities of determining the effective number of {sup 237}Np nuclei in the radiator of the neutronsensitive section are considered. The diagnostic possibilities of the detector in the counting mode are shown by example of the analysis of the reference data from the neutron-field characteristics in themore » working hall of the BR-K1 reactor. The diagnostic possibilities of the detector in the current operating mode are shown by example of the results of measuring the {sup 237}Np-fission intensity in the BR-K1 reactor power start-ups implemented in the mode of fission-pulse generation on delayed neutrons at the detector arrangement inside the reactor core cavity under conditions of a wide variation of the reactor radiation field.« less

Processing and fabrication of mixed uranium/refractory metal carbide fuels with liquid-phase sintering

NASA Astrophysics Data System (ADS)

Knight, Travis W.; Anghaie, Samim

2002-11-01

Optimization of powder processing techniques were sought for the fabrication of single-phase, solid-solution mixed uranium/refractory metal carbide nuclear fuels - namely (U, Zr, Nb)C. These advanced, ultra-high temperature nuclear fuels have great potential for improved performance over graphite matrix, dispersed fuels tested in the Rover/NERVA program of the 1960s and early 1970s. Hypostoichiometric fuel samples with carbon-to-metal ratios of 0.98, uranium metal mole fractions of 5% and 10%, and porosities less than 5% were fabricated. These qualities should provide for the longest life and highest performance capability for these fuels. Study and optimization of processing methods were necessary to provide the quality assurance of samples for meaningful testing and assessment of performance for nuclear thermal propulsion applications. The processing parameters and benefits of enhanced sintering by uranium carbide liquid-phase sintering were established for the rapid and effective consolidation and formation of a solid-solution mixed carbide nuclear fuel.

Vibration computer programs E13101, E13102, E13104, and E13112 and application to the NERVA program. Project 187: Methodology documentation

NASA Technical Reports Server (NTRS)

Mironenko, G.

1972-01-01

Programs for the analyses of the free or forced, undamped vibrations of one or two elastically-coupled lumped parameter teams are presented. Bearing nonlinearities, casing and rotor distributed mass and elasticity, rotor imbalance, forcing functions, gyroscopic moments, rotary inertia, and shear and flexural deformations are all included in the system dynamics analysis. All bearings have nonlinear load displacement characteristics, the solution is achieved by iteration. Rotor imbalances allowed by such considerations as pilot tolerances and runouts as well as bearing clearances (allowing concail or cylindrical whirl) determine the forcing function magnitudes. The computer programs first obtain a solution wherein the bearings are treated as linear springs of given spring rates. Then, based upon the computed bearing reactions, new spring rates are predicted and another solution of the modified system is made. The iteration is continued until the changes to bearing spring rates and bearing reactions become negligibly small.

ETRMTR MECHANICAL SERVICES BUILDING, TRA653. CAMERA FACING NORTHWEST AS BUILDING ...

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

ETR-MTR MECHANICAL SERVICES BUILDING, TRA-653. CAMERA FACING NORTHWEST AS BUILDING WAS NEARLY COMPLETE. INL NEGATIVE NO. 57-3653. K. Mansfield, Photographer, 7/22/1957 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

The rate of decay of fresh fission products from a nuclear reactor

NASA Astrophysics Data System (ADS)

Dolan, David J.

Determining the rate of decay of fresh fission products from a nuclear reactor is complex because of the number of isotopes involved, different types of decay, half-lives of the isotopes, and some isotopes decay into other radioactive isotopes. Traditionally, a simplified rule of 7s and 10s is used to determine the dose rate from nuclear weapons and can be to estimate the dose rate from fresh fission products of a nuclear reactor. An experiment was designed to determine the dose rate with respect to time from fresh fission products of a nuclear reactor. The experiment exposed 0.5 grams of unenriched Uranium to a fast and thermal neutron flux from a TRIGA Research Reactor (Lakewood, CO) for ten minutes. The dose rate from the fission products was measured by four Mirion DMC 2000XB electronic personal dosimeters over a period of six days. The resulting dose rate following a rule of 10s: the dose rate of fresh fission products from a nuclear reactor decreases by a factor of 10 for every 10 units of time.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-23

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Digital Instrumentation and Control Systems; Notice of Meeting The ACRS Subcommittee on Digital Instrumentation and Control Systems (DI&C) will hold a meeting on September 7, 2011, Room T-2B1, 11545 Rockville...

76 FR 32240 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Digital...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-03

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Digital Instrumentation and Control Systems; Notice of Meeting The ACRS Subcommittee on Digital Instrumentation and Control Systems (DI&C) will hold a meeting on June 7, 2011, Room T-2B1, 11545 Rockville Pike...

Search for eV sterile neutrinos at a nuclear reactor — the Stereo project

NASA Astrophysics Data System (ADS)

Haser, J.; Stereo Collaboration

2016-05-01

The re-analyses of the reference spectra of reactor antineutrinos together with a revised neutrino interaction cross section enlarged the absolute normalization of the predicted neutrino flux. The tension between previous reactor measurements and the new prediction is significant at 2.7 σ and is known as “reactor antineutrino anomaly”. In combination with other anomalies encountered in neutrino oscillation measurements, this observation revived speculations about the existence of a sterile neutrino in the eV mass range. Mixing of this light sterile neutrino with the active flavours would lead to a modification of the detected antineutrino flux. An oscillation pattern in energy and space could be resolved by a detector at a distance of few meters from a reactor core: the neutrino detector of the Stereo project will be located at about 10 m distance from the ILL research reactor in Grenoble, France. Lengthwise separated in six target cells filled with 2 m3 Gd-loaded liquid scintillator in total, the experiment will search for a position-dependent distortion in the energy spectrum.

Defining microbial community composition and seasonal variation in a sewage treatment plant in India using a down-flow hanging sponge reactor.

PubMed

Nomoto, Naoki; Hatamoto, Masashi; Hirakata, Yuga; Ali, Muntjeer; Jayaswal, Komal; Iguchi, Akinori; Okubo, Tsutomu; Takahashi, Masanobu; Kubota, Kengo; Tagawa, Tadashi; Uemura, Shigeki; Yamaguchi, Takashi; Harada, Hideki

2018-05-01

The characteristics of the microbial community in a practical-scale down-flow hanging sponge (DHS) reactor, high in organic matter and sulfate ion concentration, and the seasonal variation of the microbial community composition were investigated. Microorganisms related to sulfur oxidation and reduction (2-27%), as well as Leucobacter (7.50%), were abundant in the reactor. Anaerobic bacteria (27-38% in the first layer) were also in abundance and were found to contribute to the removal of organic matter from the sewage in the reactor. By comparing the Simpson index, the abundance-based coverage estimator (ACE) index, and the species composition of the microbial community across seasons (summer/dry, summer/rainy, autumn/dry, and winter/dry), the microbial community was found to change in composition only during the winter season. In addition to the estimation of seasonal variation, the difference in the microbial community composition along the axes of the DHS reactor was investigated for the first time. Although the abundance of each bacterial species differed along both axes of the reactor, the change of the community composition in the reactor was found to be greater along the vertical axis than the horizontal axis of the DHS reactor.

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

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

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-19

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritiummore » allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.« less

Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

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

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.

1994-04-01

The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ``flooded cavity``, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array canmore » deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications.« less

Design and fabrication of a fixed-bed batch type pyrolysis reactor for pilot scale pyrolytic oil production in Bangladesh

NASA Astrophysics Data System (ADS)

Aziz, Mohammad Abdul; Al-khulaidi, Rami Ali; Rashid, MM; Islam, M. R.; Rashid, MAN

2017-03-01

In this research, a development and performance test of a fixed-bed batch type pyrolysis reactor for pilot scale pyrolysis oil production was successfully completed. The characteristics of the pyrolysis oil were compared to other experimental results. A solid horizontal condenser, a burner for furnace heating and a reactor shield were designed. Due to the pilot scale pyrolytic oil production encountered numerous problems during the plant’s operation. This fixed-bed batch type pyrolysis reactor method will demonstrate the energy saving concept of solid waste tire by creating energy stability. From this experiment, product yields (wt. %) for liquid or pyrolytic oil were 49%, char 38.3 % and pyrolytic gas 12.7% with an operation running time of 185 minutes.

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

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

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

2010-11-01

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

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...

Innovative Approach to Validation of Ultraviolet (UV) Reactors for Disinfection in Drinking Water Systems

EPA Science Inventory

Slide presentation at Conference: ASCE 7th Civil Engineering Conference in the Asian Region. USEPA in partnership with the Cadmus Group, Carollo Engineers, and other State & Industry collaborators, are evaluating new approaches for validating UV reactors to meet groundwater & sur...

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

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

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang

2006-10-01

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progressmore » toward element testing will be reviewed.« less

Design Rules for High Temperature Microchemical Systems

DTIC Science & Technology

2006-10-25

as expected for a CSTR , while the conversion in the channel reactor is as expected for a PFR. Flow visualization using smoke to image the flow...that according to the standard Taylor-Aris analysis all reactors should show CSTR behavior in the limit of rapid diffusion of all of the reactants...0.4 0.5 0.6 0.7 0.8 C on ve rs io n CSTR PFR Data Posted Reactor PFR CSTR 0 0.1 0.2 0.3 0.4 0.5 0.6 Residence Time, Sec 0 0.2 0.4 0.6 0.8 1 C on ve

Enhanced degradation of p-chlorophenol in a novel pulsed high voltage discharge reactor.

PubMed

Bian, Wenjuan; Ying, Xiangli; Shi, Junwen

2009-03-15

The yields of active specie such as ozone, hydrogen peroxide and hydroxyl radical were all enhanced in a novel discharge reactor. In the reactor, the original formation rate of hydroxyl radical was 2.27 x 10(-7) mol L(-1)s(-1), which was about three times than that in the contrast reactor. Ozone was formed in gas-phase and was transferred into the liquid. The characteristic of mass transfer was better in the novel reactor than that in the contrast reactor, which caused much higher ozone concentration in liquid. The dissociation of hydrogen peroxide was more evident in the former, which promoted the formations of hydroxyl radical. The p-chlorophenol (4-CP) degradation was also enhanced. Most of the ozone transferred into the liquid and hydrogen peroxide generated by discharge could be utilized by the degradation process of 4-CP. About 97% 4-CP was removed in 36 min discharge in the novel reactor. Organic acids such as formic, acetic, oxalic, propanoic and maleic acid were generated and free chloride ions were released in the degradation process. With the formation of organic acid, the pH was decreased and the conductivity was increased.

Treatment of phenolic wastewater in an anaerobic fixed bed reactor (AFBR) - recovery after shock loading.

PubMed

Bajaj, Mini; Gallert, Claudia; Winter, Josef

2009-03-15

An anaerobic fixed bed reactor (AFBR) was run for 550 days with a mixed microbial flora to stabilize synthetic wastewater that contained glucose and phenol as main carbon sources. The influent phenol concentration was gradually increased from 2 to 40 mmol/l within 221 days. The microbial flora was able to adapt to this high phenol concentration with an average of 94% phenol removal. Microbial adaptation at such a high phenol concentration is not reported elsewhere. The maximum phenol removal observed before the phenol shock load was 39.47 mmol/l or 3.7 g phenol/l at a hydraulic retention time (HRT) of 2.5 days and an organic loading rate (OLR) of 5.3 g/l.d which amounts to a phenol removal rate of ca. 15.8 mmol phenol/l.d. The chemical oxygen demand (COD) removal before exposing the reactor to a shock load corresponded with phenol removal. A shock load was induced in the reactor by increasing the phenol concentration from 40 to 50 mmol/l in the influent. The maximum phenol removal rate observed after shock load was 18 mmol/l.d at 5.7 g COD/l.d. But this was not a stable rate and a consistent drop in COD and phenol removal was observed for 1 week, followed by a sharp decline and production of fatty acids. Recovery of the reactor was possible only when no feed was provided to the reactor for 1 month and the phenol concentration was increased gradually. When glucose was omitted from the influent, unknown intermediates of anaerobic phenol metabolism were observed for some time.

WATER PROCESS SYSTEM FLOW DIAGRAM FOR MTR, TRA603. SUMMARY OF ...

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

WATER PROCESS SYSTEM FLOW DIAGRAM FOR MTR, TRA-603. SUMMARY OF COOLANT FLOW FROM WORKING RESERVOIR TO INTERIOR OF REACTOR'S THERMAL SHIELD. NAMES TANK SECTIONS. PIPE AND DRAIN-LINE SIZES. SHOWS DIRECTION OF AIR FLOW THROUGH PEBBLE AND GRAPHITE BLOCK ZONE. NEUTRON CURTAIN AND THERMAL COLUMN DOOR. BLAW-KNOX 3150-92-7, 3/1950. INL INDEX NO. 531-0603-51-098-100036, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Hydrogen and water reactor safety: proceedings

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

Not Available

1982-01-01

Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

Basic requirements for a 1000-MW(electric) class tokamak fusion-fission hybrid reactor and its blanket concept

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

Hatayama, Ariyoshi; Ogasawara, Masatada; Yamauchi, Michinori

1994-08-01

Plasma size and other basic performance parameters for 1000-MW(electric) power production are calculated with the blanket energy multiplication factor, the M value, as a parameter. The calculational model is base don the International Thermonuclear Experimental Reactor (ITER) physics design guidelines and includes overall plant power flow. Plasma size decreases as the M value increases. However, the improvement in the plasma compactness and other basic performance parameters, such as the total plant power efficiency, becomes saturated above the M = 5 to 7 range. THus, a value in the M = 5 to 7 range is a reasonable choice for 1000-MW(electric)more » hybrids. Typical plasma parameters for 1000-MW(electric) hybrids with a value of M = 7 are a major radius of R = 5.2 m, minor radius of a = 1.7 m, plasma current of I{sub p} = 15 MA, and toroidal field on the axis of B{sub o} = 5 T. The concept of a thermal fission blanket that uses light water as a coolant is selected as an attractive candidate for electricity-producing hybrids. An optimization study is carried out for this blanket concept. The result shows that a compact, simple structure with a uniform fuel composition for the fissile region is sufficient to obtain optimal conditions for suppressing the thermal power increase caused by fuel burnup. The maximum increase in the thermal power is +3.2%. The M value estimated from the neutronics calculations is {approximately}7.0, which is confirmed to be compatible with the plasma requirement. These studies show that it is possible to use a tokamak fusion core with design requirements similar to those of ITER for a 1000-MW(electric) power reactor that uses existing thermal reactor technology for the blanket. 30 refs., 22 figs., 4 tabs.« less

Proteotyping of laboratory-scale biogas plants reveals multiple steady-states in community composition.

PubMed

Kohrs, F; Heyer, R; Bissinger, T; Kottler, R; Schallert, K; Püttker, S; Behne, A; Rapp, E; Benndorf, D; Reichl, U

2017-08-01

Complex microbial communities are the functional core of anaerobic digestion processes taking place in biogas plants (BGP). So far, however, a comprehensive characterization of the microbiomes involved in methane formation is technically challenging. As an alternative, enriched communities from laboratory-scale experiments can be investigated that have a reduced number of organisms and are easier to characterize by state of the art mass spectrometric-based (MS) metaproteomic workflows. Six parallel laboratory digesters were inoculated with sludge from a full-scale BGP to study the development of enriched microbial communities under defined conditions. During the first three month of cultivation, all reactors (R1-R6) were functionally comparable regarding biogas productions (375-625 NL L reactor volume -1 d -1 ), methane yields (50-60%), pH values (7.1-7.3), and volatile fatty acids (VFA, <5 mM). Nevertheless, a clear impact of the temperature (R3, R4) and ammonia (R5, R6) shifts was observed for the respective reactors. In both reactors operated under thermophilic regime, acetic and propionic acid (10-20 mM) began to accumulate. While R4 recovered quickly from acidification, the levels of VFA remained to be high in R3 resulting in low pH values of 6.5-6.9. The digesters R5 and R6 operated under the high ammonia regime (>1 gNH 3 L -1 ) showed an increase to pH 7.5-8.0, accumulation of acetate (>10 mM), and decreasing biogas production (<125 NL L reactor volume -1 d -1 ). Tandem MS (MS/MS)-based proteotyping allowed the identification of taxonomic abundances and biological processes. Although all reactors showed similar performances, proteotyping and terminal restriction fragment length polymorphisms (T-RFLP) fingerprinting revealed significant differences in the composition of individual microbial communities, indicating multiple steady-states. Furthermore, cellulolytic enzymes and cellulosomal proteins of Clostridium thermocellum were identified to be specific markers for the thermophilic reactors (R3, R4). Metaproteins found in R3 indicated hydrogenothrophic methanogenesis, whereas metaproteins of acetoclastic methanogenesis were identified in R4. This suggests not only an individual evolution of microbial communities even for the case that BGPs are started at the same initial conditions under well controlled environmental conditions, but also a high compositional variance of microbiomes under extreme conditions. Copyright © 2017 Elsevier Ltd. All rights reserved.

78 FR 68867 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on US-APWR...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-15

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on US-APWR; Notice of Meeting The ACRS Subcommittee on US-APWR will hold a meeting on November 20... Sections 3.7 and 3.8) of the Safety Evaluation Report (SER) associated with the US-APWR design...

78 FR 20959 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on US-APWR

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-08

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on US-APWR The ACRS Subcommittee on US-APWR will hold a meeting on April 25- 26, 2013, Room T-2B1... 7, ``Instrumentation and Control,'' of the Safety Evaluation Report (SER) associated with the US...

The use of moving bed bio-reactor to laundry wastewater treatment

NASA Astrophysics Data System (ADS)

Bering, Sławomira; Mazur, Jacek; Tarnowski, Krzysztof; Janus, Magdalena; Mozia, Sylwia; Waldemar Morawski, Antoni

2017-11-01

Large laboratory scale biological treatment test of industrial real wastewater, generated in industrial big laundry, has been conducted in the period of May 2016-August 2016. The research aimed at selection of laundry wastewater treatment technology included tests of two-stage Moving Bed Bio Reactor (MBBR), with two reactors filled with carriers Kaldnes K5 (specific area - 800 m2/m3), have been realized in aerobic condition. Operating on site, in the laundry, reactors have been fed real wastewater from laundry retention tank. To the laundry wastewater, contained mainly surfactants and impurities originating from washed fabrics, a solution of urea to supplement nitrogen content and a solution of acid to correct pH have been added. Daily flow of raw wastewater Qd was equal to 0.6-0.8 m3/d. The values of determined wastewater quality indicators showed that substantial decrease of pollutants content have been reached: BOD5 by 94.7-98.1%, COD by 86.9-93.5%, the sum of anionic and nonionic surfactants by 98.7-99.8%. The quality of the purified wastewater, after star-up period, meets the legal requirements regarding the standards for wastewater discharged to the environment.

Utilization of useless pesticides in a plasma reactor

NASA Astrophysics Data System (ADS)

Lozhechnik, A. V.; Mossé, A. L.; Savchin, V. V.; Skomorokhov, D. S.; Khvedchin, I. V.

2011-09-01

Investigations on destruction of isophene C14H18O7N2 and the butyl ether of 2,4-dichlorophenoxyacetic acid (Cl2C6H3OCH2COOCH2CH(CH3)2) are performed. The plasma treatment of toxic waste is implemented in a plasma reactor with a three-jet mixing chamber. Air is used as the plasma-forming gas.

Simulation of German PKL refill/reflood experiment K9A using RELAP4/MOD7. [PWR

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

Hsu, M.T.; Davis, C.B.; Behling, S.R.

This paper describes a RELAP4/MOD7 simulation of West Germany's Kraftwerk Union (KWU) Primary Coolant Loop (PKL) refill/reflood experiment K9A. RELAP4/MOD7, a best-estimate computer program for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This study was the first major simulation using RELAP4/MOD7 since its release by the Idaho National Engineering Laboratory (INEL). The PKL facility is a reduced scale (1:134) representation of a typical West German four-loop 1300 MW pressurized water reactor (PWR). A prototypical scale of the total volume to power ratio wasmore » maintained. The test facility was designed specifically for an experiment simulating the refill/reflood phase of a Loss-of-Coolant Accident (LOCA).« less

Oxidation of aluminum alloy cladding for research and test reactor fuel

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

2008-08-01

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

Electrochemical treatment of tannery effluent using a battery integrated DC-DC converter and solar PV power supply--an approach towards environment and energy management.

PubMed

Iyappan, K; Basha, C Ahmed; Saravanathamizhan, R; Vedaraman, N; Tahiyah Nou Shene, C A; Begum, S Nathira

2014-01-01

Electrochemical oxidation of tannery effluent was carried out in batch, batch recirculation and continuous reactor configurations under different conditions using a battery-integrated DC-DC converter and solar PV power supply. The effect of current density, electrolysis time and fluid flow rate on chemical oxygen demand (COD) removal and energy consumption has been evaluated. The results of batch reactor show that a COD reduction of 80.85% to 96.67% could be obtained. The results showed that after 7 h of operation at a current density of 2.5 A dm(-2) and flow rate of 100 L h(-1) in batch recirculation reactor, the removal of COD is 82.14% and the specific energy consumption was found to be 5.871 kWh (kg COD)(-1) for tannery effluent. In addition, the performance of single pass flow reactors (single and multiple reactors) system of various configurations are analyzed.

Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation

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

Ross, Kyle W.; Gauntt, Randall O.; Cardoni, Jeffrey N.

2013-11-01

Data, a brief description of key boundary conditions, and results of Sandia National Laboratories’ ongoing MELCOR analysis of the Fukushima Unit 2 accident are given for the reactor core isolation cooling (RCIC) system. Important assumptions and related boundary conditions in the current analysis additional to or different than what was assumed/imposed in the work of SAND2012-6173 are identified. This work is for the U.S. Department of Energy’s Nuclear Energy University Programs fiscal year 2014 Reactor Safety Technologies Research and Development Program RC-7: RCIC Performance under Severe Accident Conditions.

Calculation and Experiment of Adding Top Beryllium Shims for Iran MNSR

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

Ebadati, Javad; Rezvanifard, Mehdi; Shahabi, Iraj

2006-07-01

Miniature Neutron Source Reactor which is called MNSR were put into operation on June 1994 in Esfahan Nuclear Technology Center (ENTC). At that time the excess reactivity at the cold condition was 3.85 mk. After 7 years of operation and fuel consumption the reactivity was reduced to 2.90 mk. To compensate this reduction and upgrade the reactor, Beryllium Shim were used at the top of the core. This paper discusses the steps for this accurate and sensitive task. Finally a layer of 1.5 mm Beryllium were added to restore the reactor life. (authors)

Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

NASA Astrophysics Data System (ADS)

Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

2016-04-01

This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.

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.

The PROSPECT physics program

NASA Astrophysics Data System (ADS)

Ashenfelter, J.; Balantekin, A. B.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bignell, L.; Bowden, N. S.; Bowes, A.; Brodsky, J. P.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Commeford, K.; Conant, A. J.; Davee, D.; Dean, D.; Deichert, G.; Diwan, M. V.; Dolinski, M. J.; Dolph, J.; DuVernois, M.; Erikson, A. S.; Febbraro, M. T.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Goddard, B. W.; Green, M.; Hackett, B. T.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Insler, J.; Jaffe, D. E.; Jones, D.; Langford, T. J.; Littlejohn, B. R.; Martinez Caicedo, D. A.; Matta, J. T.; McKeown, R. D.; Mendenhall, M. P.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Neilson, R.; Nikkel, J. A.; Norcini, D.; Pushin, D.; Qian, X.; Romero, E.; Rosero, R.; Seilhan, B. S.; Sharma, R.; Sheets, S.; Surukuchi, P. T.; Trinh, C.; Varner, R. L.; Viren, B.; Wang, W.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yeh, M.; Yen, Y.-R.; Zangakis, G. Z.; Zhang, C.; Zhang, X.; PROSPECT Collaboration

2016-11-01

The precision reactor oscillation and spectrum experiment, PROSPECT, is designed to make a precise measurement of the antineutrino spectrum from a highly-enriched uranium reactor and probe eV-scale sterile neutrinos by searching for neutrino oscillations over a distance of several meters. PROSPECT is conceived as a 2-phase experiment utilizing segmented 6Li-doped liquid scintillator detectors for both efficient detection of reactor antineutrinos through the inverse beta decay reaction and excellent background discrimination. PROSPECT Phase I consists of a movable 3 ton antineutrino detector at distances of 7-12 m from the reactor core. It will probe the best-fit point of the {ν }e disappearance experiments at 4σ in 1 year and the favored region of the sterile neutrino parameter space at \\gt 3σ in 3 years. With a second antineutrino detector at 15-19 m from the reactor, Phase II of PROSPECT can probe the entire allowed parameter space below 10 eV2 at 5σ in 3 additional years. The measurement of the reactor antineutrino spectrum and the search for short-baseline oscillations with PROSPECT will test the origin of the spectral deviations observed in recent {θ }13 experiments, search for sterile neutrinos, and conclusively address the hypothesis of sterile neutrinos as an explanation of the reactor anomaly.

How much does the MSW effect contribute to the reactor antineutrino anomaly?

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

Valdiviesso, G. A.

2015-05-15

It has been pointed out that there is a 5.7 ± 2.3 discrepancy between the predicted and the observed reactor antineutrino flux in very short baseline experiments. Several causes for this anomaly have been discussed, including a possible non-standard forth sterile neutrino. In order to quantify how much non-standard this anomaly really is, the standard MSW effect is reviewed. Knowing that reactor antineutrinos are produced in a dense medium (the nuclear fuel) and is usually detected in a less dense one (water, or scintillator), non-adiabatic effects are expected to happen, creating a difference between the creation and detection mixing angles.

Affordable Development and Demonstration of a Small NTR Engine and Stage: How Small is Big Enough?

NASA Technical Reports Server (NTRS)

Borowski, S. K.; Sefcik, R. J.; Fittje, J. E.; McCurdy, D. R.; Qualls, A. L.; Schnitzler, B. G.; Werner, J.; Weitzberg, A.; Joyner, C. R.

2015-01-01

In FY11, NASA formulated a plan for Nuclear Thermal Propulsion (NTP) development that included Foundational Technology Development followed by system-level Technology Demonstrations The ongoing NTP project, funded by NASAs Advanced Exploration Systems (AES) program, is focused on Foundational Technology Development and includes 5 key task activities:(1) Fuel element fabrication and non-nuclear validation testing of heritage fuel options;(2) Engine conceptual design;(3) Mission analysis and engine requirements definition;(4) Identification of affordable options for ground testing; and(5) Formulation of an affordable and sustainable NTP development program Performance parameters for Point of Departure designs for a small criticality-limited and full size 25 klbf-class engine were developed during FYs 13-14 using heritage fuel element designs for both RoverNERVA Graphite Composite (GC) and Ceramic Metal (Cermet) fuel forms To focus the fuel development effort and maximize use of its resources, the AES program decided, in FY14, that a leader-follower down selection between GC and cermet fuel was required An Independent Review Panel (IRP) was convened by NASA and tasked with reviewing the available fuel data and making a recommendation to NASA. In February 2015, the IRP recommended and the AES program endorsed GC as the leader fuel In FY14, a preliminary development schedule DDTE plan was produced by GRC, DOE industry for the AES program. Assumptions, considerations and key task activities are presented here Two small (7.5 and 16.5 klbf) engine sizes were considered for ground and flight technology demonstration within a 10-year timeframe; their ability to support future human exploration missions was also examined and a recommendation on a preferred size is provided.

Posttest analysis of international standard problem 10 using RELAP4/MOD7. [PWR

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

Hsu, M.; Davis, C.B.; Peterson, A.C. Jr.

RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of amore » hypothetical loss-of-coolant accident (LOCA).« less

Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Jaradat, Safwan Qasim Mohammad

Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

DOEpatents

Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

1999-03-23

A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

Sequential dark and photo fermentation hydrogen production from hydrolyzed corn stover: A pilot test using 11 m3 reactor.

PubMed

Zhang, Quanguo; Zhang, Zhiping; Wang, Yi; Lee, Duu-Jong; Li, Gang; Zhou, Xuehua; Jiang, Danping; Xu, Bo; Lu, Chaoyang; Li, Yameng; Ge, Xumeng

2018-04-01

Pilot tests of sequential dark and photo fermentation H 2 production were for the first time conducted in a 11 m 3 reactor (3 m 3 for dark and 8 m 3 for photo compartments). A combined solar and light-emitting diode illumination system and a thermal controlling system was installed and tested. With dark fermentation unit maintained at pH 4.5 and 35 °C and photo fermentation unit at pH 7.0 and 30 °C, the overall biogas production rate using hydrolyzed corn stover as substrate reached 87.8 ± 3.8 m 3 /d with 68% H 2 content, contributed by dark unit at 7.5 m 3 -H 2 /m 3 -d and by photo unit at 4.7 m 3 /m 3 -d. Large variation was noted for H 2 production rate in different compartments of the tested units, revealing the adverse effects of poor mixing, washout, and other inhomogeneity associated with large reactor operations. Copyright © 2018 Elsevier Ltd. All rights reserved.

ETR, TRA642. FLOOR PLAN UNDER BALCONY ON CONSOLE FLOOR. MOTORGENERATOR ...

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

ETR, TRA-642. FLOOR PLAN UNDER BALCONY ON CONSOLE FLOOR. MOTOR-GENERATOR SETS AND OTHER ELECTRICAL EQUIPMENT. PHILLIPS PETROLEUM COMPANY ETR-D-1781, 7/1960. INL INDEX NO. 532-0642-00-706-020384, REV. 1. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

SOUTH WING, MTR661. INTERIOR DETAIL INSIDE LAB ROOM 131. CAMERA ...

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

SOUTH WING, MTR-661. INTERIOR DETAIL INSIDE LAB ROOM 131. CAMERA FACING NORTHEAST. NOTE CONCRETE BLOCK WALLS. SAFETY SHOWER AND EYE WASHER AT REAR WALL. INL NEGATIVE NO. HD46-7-2. Mike Crane, Photographer, 2/2005. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

78 FR 71673 - Agency Information Collection Activities: Submission for the Office of Management and Budget (OMB...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-29

..., power reactors under construction, research and test reactors, agreement states, non-agreement states, as well as users of byproduct material (e.g. departments of health, medical centers, steel mills, well loggers, and radiographers.) 7. An estimate of the number of annual responses: 339. [[Page 71674...

Use of Nitrogen Trifluoride To Purify Molten Salt Reactor Coolant and Heat Transfer Fluoride Salts

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

Scheele, Randall D.; Casella, Andrew M.; McNamara, Bruce K.

2017-05-02

Abstract: The molten salt cooled nuclear reactor is included as one of the Generation IV reactor types. One of the challenges with the implementation of this reactor is purifying and maintaining the purity of the various molten fluoride salts that will be used as coolants. The method used for Oak Ridge National Laboratory’s molten salt experimental test reactor was to treat the coolant with a mixture of H2 and HF at 600°C. In this article we evaluate thermal NF3 treatment for purifying molten fluoride salt coolant candidates based on NF3’s 1) past use to purify fluoride salts, 2) other industrialmore » uses, 3) commercial availability, 4) operational, chemical, and health hazards, 5) environmental effects and environmental risk management methods, 6) corrosive properties, and 7) thermodynamic potential to eliminate impurities that could arise due to exposure to water and oxygen. Our evaluation indicates that nitrogen trifluoride is a viable and safer alternative to the previous method.« less

Mini-MITEE: Ultra Small, Ultra Light NTP Engines for Robotic Science and Manned Exploration Missions

NASA Astrophysics Data System (ADS)

Powell, James; Maise, George; Paniagua, John

2006-01-01

A compact, ultra lightweight Nuclear Thermal Propulsion (NTP) engine design is described with the capability to carry out a wide range of unique and important robotic science missions that are not possible using chemical or Nuclear Electric Propulsion (NEP). The MITEE (MInature ReacTor EnginE) reactor uses hydrogeneous moderator, such as solid lithium-7 hydride, and high temperature cermet tungsten/UO2 nuclear fuel. The reactor is configured as a modular pressure tube assembly, with each pressure tube containing an outer annual shell of moderator with an inner annular region of W/UO2 cermet fuel sheets. H2 propellant flows radially inwards through the moderator and fuel regions, exiting at ~3000 K into a central channel that leads to a nozzle at the end of the pressure tube. Power density in the fuel region is 10 to 20 megawatts per liter, depending on design, producing a thrust output on the order of 15,000 Newtons and an Isp of ~1000 seconds. 3D Monte Carlo neutronic analyses are described for MITEE reactors utilizing various fissile fuel options (U-235, U-233, and Am242m) and moderators (7LiH and BeH2). Reactor mass ranges from a maximum of 100 kg for the 7LiH/U-235 option to a minimum of 28 kg for the BeH2/Am-242 m option. Pure thrust only and bi-modal (thrust plus electric power generation) MITEE designs are described. Potential unique robotic science missions enabled by the MITEE engine are described, including landing on Europa and exploring the ice sheet interior with return of samples to Earth, hopping to and exploring multiple sites on Mars, unlimited ramjet flight in the atmospheres of Jupiter, Saturn, Uranus, and Neptune and landing on, and sample return from Pluto.

Mini-MITEE: Ultra Small, Ultra Light NTP Engines for Robotic Science and Manned Exploration Missions

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

Powell, James; Maise, George; Paniagua, John

2006-01-20

A compact, ultra lightweight Nuclear Thermal Propulsion (NTP) engine design is described with the capability to carry out a wide range of unique and important robotic science missions that are not possible using chemical or Nuclear Electric Propulsion (NEP). The MITEE (MInature ReacTor EnginE) reactor uses hydrogeneous moderator, such as solid lithium-7 hydride, and high temperature cermet tungsten/UO2 nuclear fuel. The reactor is configured as a modular pressure tube assembly, with each pressure tube containing an outer annual shell of moderator with an inner annular region of W/UO2 cermet fuel sheets. H2 propellant flows radially inwards through the moderator andmore » fuel regions, exiting at {approx}3000 K into a central channel that leads to a nozzle at the end of the pressure tube. Power density in the fuel region is 10 to 20 megawatts per liter, depending on design, producing a thrust output on the order of 15,000 Newtons and an Isp of {approx}1000 seconds. 3D Monte Carlo neutronic analyses are described for MITEE reactors utilizing various fissile fuel options (U-235, U-233, and Am242m) and moderators (7LiH and BeH2). Reactor mass ranges from a maximum of 100 kg for the 7LiH/U-235 option to a minimum of 28 kg for the BeH2/Am-242 m option. Pure thrust only and bi-modal (thrust plus electric power generation) MITEE designs are described. Potential unique robotic science missions enabled by the MITEE engine are described, including landing on Europa and exploring the ice sheet interior with return of samples to Earth, hopping to and exploring multiple sites on Mars, unlimited ramjet flight in the atmospheres of Jupiter, Saturn, Uranus, and Neptune and landing on, and sample return from Pluto.« less

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

Research on sludge-fly ash ceramic particles (SFCP) for synthetic and municipal wastewater treatment in biological aerated filter (BAF).

PubMed

Zhao, Yaqin; Yue, Qinyan; Li, Renbo; Yue, Min; Han, Shuxin; Gao, Baoyu; Li, Qian; Yu, Hui

2009-11-01

Sludge-fly ash ceramic particles (SFCP) and clay ceramic particles (CCP) were employed in two lab-scale up-flow biological aerated filters (BAF) for wastewater treatment to investigate the availability of SFCP used as biofilm support compared with CCP. For synthetic wastewater, under the selected hydraulic retention times (HRT) of 1.5, 0.75 and 0.37 h, respectively, the removal efficiencies of chemical oxygen demand (COD(Cr)) and ammonium nitrogen (NH(4)(+)-N) in SFCP reactor were all higher than those of CCP reactor all through the media height. Moreover, better capabilities responding to loading shock and faster recovery after short intermittence were observed in the SFCP reactor compared with the CCP reactor. For municipal wastewater treatment, which was carried out under HRT of 0.75 h, air-liquid ratio of 7.5 and backwashing period of 48 h, the SFCP reactor also performed better than the CCP reactor, especially for the removal of NH(4)(+)-N.

Characterization of bacterial and archaeal communities in air-cathode microbial fuel cells, open circuit and sealed-off reactors.

PubMed

Shehab, Noura; Li, Dong; Amy, Gary L; Logan, Bruce E; Saikaly, Pascal E

2013-11-01

A large percentage of organic fuel consumed in a microbial fuel cell (MFC) is lost as a result of oxygen transfer through the cathode. In order to understand how this oxygen transfer affects the microbial community structure, reactors were operated in duplicate using three configurations: closed circuit (CC; with current generation), open circuit (OC; no current generation), and sealed off cathodes (SO; no current, with a solid plate placed across the cathode). Most (98 %) of the chemical oxygen demand (COD) was removed during power production in the CC reactor (maximum of 640 ± 10 mW/m(2)), with a low percent of substrate converted to current (coulombic efficiency of 26.5 ± 2.1 %). Sealing the cathode reduced COD removal to 7 %, but with an open cathode, there was nearly as much COD removal by the OC reactor (94.5 %) as the CC reactor. Oxygen transfer into the reactor substantially affected the composition of the microbial communities. Based on analysis of the biofilms using 16S rRNA gene pyrosequencing, microbes most similar to Geobacter were predominant on the anodes in the CC MFC (72 % of sequences), but the most abundant bacteria were Azoarcus (42 to 47 %) in the OC reactor, and Dechloromonas (17 %) in the SO reactor. Hydrogenotrophic methanogens were most predominant, with sequences most similar to Methanobacterium in the CC and SO reactor, and Methanocorpusculum in the OC reactors. These results show that oxygen leakage through the cathode substantially alters the bacterial anode communities, and that hydrogenotrophic methanogens predominate despite high concentrations of acetate. The predominant methanogens in the CC reactor most closely resembled those in the SO reactor, demonstrating that oxygen leakage alters methanogenic as well as general bacterial communities.

Advanced transportation system studies. Alternate propulsion subsystem concepts: Propulsion database

NASA Technical Reports Server (NTRS)

Levack, Daniel

1993-01-01

The Advanced Transportation System Studies alternate propulsion subsystem concepts propulsion database interim report is presented. The objective of the database development task is to produce a propulsion database which is easy to use and modify while also being comprehensive in the level of detail available. The database is to be available on the Macintosh computer system. The task is to extend across all three years of the contract. Consequently, a significant fraction of the effort in this first year of the task was devoted to the development of the database structure to ensure a robust base for the following years' efforts. Nonetheless, significant point design propulsion system descriptions and parametric models were also produced. Each of the two propulsion databases, parametric propulsion database and propulsion system database, are described. The descriptions include a user's guide to each code, write-ups for models used, and sample output. The parametric database has models for LOX/H2 and LOX/RP liquid engines, solid rocket boosters using three different propellants, a hybrid rocket booster, and a NERVA derived nuclear thermal rocket engine.

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.).

Fuels Combustion Research.

DTIC Science & Technology

1985-12-09

benzene and toluene have been greatly improved and show even better corre- spondence with flow reactor results. The studies of alkylated aromatics have...of the flow reactor ; the use of a unique high temperature sampling system; and an automated gas chromatographic apparatus , and the presence of the gas...In these studies a clear analogy between the reactions of the alkylated aromatics and those of the, corresponding alkanes wqs observed [7,8,93. This

The phage‐driven microbial loop in petroleum bioremediation

PubMed Central

Rosenberg, Eugene; Bittan‐Banin, Gili; Sharon, Gil; Shon, Avital; Hershko, Galit; Levy, Itzik; Ron, Eliora Z.

2010-01-01

Summary During the drilling process and transport of crude oil, water mixes with the petroleum. At oil terminals, the water settles to the bottom of storage tanks. This drainage water is contaminated with emulsified oil and water‐soluble hydrocarbons and must be treated before it can be released into the environment. In this study, we tested the efficiency of a continuous flow, two‐stage bioreactor for treating drainage water from an Israeli oil terminal. The bioreactor removed all of the ammonia, 93% of the sulfide and converted 90% of the total organic carbon (TOC) into carbon dioxide. SYBR Gold staining indicated that reactor 1 contained 1.7 × 108 bacteria and 3.7 × 108 phages per millilitre, and reactor 2 contained 1.3 × 108 bacteria and 1.7 × 109 phages per millilitre. The unexpectedly high mineralization of TOC and high concentration of phage in reactor 2 support the concept of a phage‐driven microbial loop in the bioremediation of the drainage water. In general, application of this concept in bioremediation of contaminated water has the potential to increase the efficiency of processes. PMID:21255344

Effect of pentachlorophenol and chemical oxygen demand mass concentrations in influent on operational behaviors of upflow anaerobic sludge blanket (UASB) reactor.

PubMed

Shen, Dong-Sheng; He, Ruo; Liu, Xin-Wen; Long, Yan

2006-08-25

Upflow anaerobic sludge blanket (UASB) reactor that was seeded with anaerobic sludge acclimated to chlorophenols was used to investigate the feasibility of anaerobic biotreatment of synthetic wastewater containing pentachlorophenol (PCP) with additional sucrose as carbon source. Two sets of UASB reactors were operated at one time. But the seeded sludge for the two reactors was different and Reactor I was seeded with the sludge that was acclimated to PCP completely for half a year, and Reactor II was seeded with the mixed sludge that was acclimated for half a year to PCP, 4-CP, 3-CP or 2-CP, respectively. The degradation of PCP and the operation fee treating the wastewater are affected by the concentration of MEDS (microorganism easily degradable substrate). So the confirmation of the suitable ratio of [COD] and [PCP] was the key factor of treating the wastewater containing PCP economically and efficiently. During the experiment, the synthetic wastewater with 180.0 mg L(-1) PCP and 1250-10000 mg L(-1) COD could be treated steadily in the experimental Reactor I. The removal efficiency of PCP was more than 99.5% and the removal efficiency of COD was up to 90%. [PCP] (concentration of PCP) in effluent was less than 0.5 mg L(-1). [PCP] in influent could affect proper [COD] (concentration of COD) range in influent that was required for maintenance of steady running of the experimental reactor with a hydraulic retention time (HRT) from 20 to 22 h. [PCP] in influent would directly affect the necessary [COD] in influent when the UASB reactor ran normally and treated the wastewater containing PCP. When [PCP] was 100.4, 151.6 and 180.8 mg L(-1) in influent, respectively, [COD] in influent had to be controlled about 1250-7500, 2500-5000 and 5000 mg L(-1) to maintain the UASB reactor steady running normally and contemporarily ensure that [COD] and [PCP] in effluent were less than 300 and 0.5 mg L(-1), respectively. With the increase of [PCP] in influent, the range of variation of [COD] in influent endured by the UASB reactor was decreasing. The ratios of [COD] and [PCP] in influent could affect removal efficiency of PCP and COD, the concentration of total volatile fatty acids (VFA) in effluent, biogas quantity and methane content in biogas. [PCP] in influent was linearly or semi-logarithmically correlated to [COD] in effluent when [COD] in influent was 5750+/-250 mg L(-1), and so was the relationship between [COD] in influent and [PCP] in effluent when [PCP] in influent was 100.4 or 151.6 mg L(-1), less than the maximum permissible [PCP]. The sources of seeded sludge, the way of sludge acclimation and the characteristics of anaerobic sludge could all affect the UASB reactor capacity treating PCP. When [PCP] were less than 180.8 mg L(-1) for Reactor I and 151.6 mg L(-1) for Reactor II, the variation of [PCP] in influent had little effect on the UASB reactor volume gas production rate and substrate gas production rate. And [VFA] and pH value in effluent were affected a little. Volume biogas production rate and substrate biogas production rate of the UASB reactor were only affected by [COD] and loading rate in influent. But when [PCP] was more than 151.6 mg L(-1) for Reactor II, the biogas production fell quickly and was over 3 days later. [VFA] in effluent from Reactor II increased up to 2198.1 mg L(-1) quickly and the pH value fell to less than 7. Reactor II could not run normally. The component of VFA accumulated quickly was mainly acetate (above 50%). With [PCP] increased from 7.9 to 180.8 mg L(-1) gradually in influent, the methane content in biogas from Reactor II decreased from 70% to 60%, but the reactor could still run normally. Then as for Reactor II, the content of methane have fallen from 75% to 45% or so quickly. And Reactor II could not run steadily. So the conclusion could be drown that too high [PCP] in influent for UASB reactor mainly inhibited the activity of methane-producing bacteria cultures utilizing the acetate.

Final summary report for 1989 inservice inspection (ISI) of SRS (Savannah River Site) 100-P Reactor tank

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

Morrison, J.M.; Loibl, M.W.

1989-12-15

The integrity of the SRS reactor tanks is a key factor affecting their suitability for continued service since, unlike the external piping system and components, the tanks are virtually irreplaceable. Cracking in various areas of the process water piping systems has occurred beginning in 1960 as a result of several degradation mechanisms, chiefly intergranular stress corrosion cracking (IGSCC) and chloride-induced transgranular cracking. IGSCC, currently the primary degradation mechanism, also occurred in the knuckle'' region (tank wall-to-bottom tube sheet transition piece) unique to C Reactor and was eventually responsible for that reactor being deactivated in 1985. A program of visual examinationsmore » of the SRS reactor tanks was initiated in 1968, which used a specially designed immersible periscope. Under that program the condition of the accessible tank welds and associated heat affected zones (HAZ) was evaluated on a five-year frequency. Prior to 1986, the scope of these inspections comprised approximately 20 percent of the accessible weld area. In late 1986 and early 1987 the scope of the inspections was expanded and a 100 percent visual inspection of accessible welds was performed of the P-, L-, and K-Reactor tanks. Supplemental dye penetrant examinations were performed in L Reactor on selected areas which showed visual indications. No evidence of cracking was detected in any of these inspections of the P-, L-, and K-Reactor tanks. 17 refs., 7 figs.« less

Precision measurement of neutrino oscillation parameters with KamLAND.

PubMed

Abe, S; Ebihara, T; Enomoto, S; Furuno, K; Gando, Y; Ichimura, K; Ikeda, H; Inoue, K; Kibe, Y; Kishimoto, Y; Koga, M; Kozlov, A; Minekawa, Y; Mitsui, T; Nakajima, K; Nakajima, K; Nakamura, K; Nakamura, M; Owada, K; Shimizu, I; Shimizu, Y; Shirai, J; Suekane, F; Suzuki, A; Takemoto, Y; Tamae, K; Terashima, A; Watanabe, H; Yonezawa, E; Yoshida, S; Busenitz, J; Classen, T; Grant, C; Keefer, G; Leonard, D S; McKee, D; Piepke, A; Decowski, M P; Detwiler, J A; Freedman, S J; Fujikawa, B K; Gray, F; Guardincerri, E; Hsu, L; Kadel, R; Lendvai, C; Luk, K-B; Murayama, H; O'Donnell, T; Steiner, H M; Winslow, L A; Dwyer, D A; Jillings, C; Mauger, C; McKeown, R D; Vogel, P; Zhang, C; Berger, B E; Lane, C E; Maricic, J; Miletic, T; Batygov, M; Learned, J G; Matsuno, S; Pakvasa, S; Foster, J; Horton-Smith, G A; Tang, A; Dazeley, S; Downum, K E; Gratta, G; Tolich, K; Bugg, W; Efremenko, Y; Kamyshkov, Y; Perevozchikov, O; Karwowski, H J; Markoff, D M; Tornow, W; Heeger, K M; Piquemal, F; Ricol, J-S

2008-06-06

The KamLAND experiment has determined a precise value for the neutrino oscillation parameter Deltam21(2) and stringent constraints on theta12. The exposure to nuclear reactor antineutrinos is increased almost fourfold over previous results to 2.44 x 10(32) proton yr due to longer livetime and an enlarged fiducial volume. An undistorted reactor nu[over]e energy spectrum is now rejected at >5sigma. Analysis of the reactor spectrum above the inverse beta decay energy threshold, and including geoneutrinos, gives a best fit at Deltam21(2)=7.58(-0.13)(+0.14)(stat) -0.15+0.15(syst) x 10(-5) eV2 and tan2theta12=0.56(-0.07)+0.10(stat) -0.06+0.10(syst). Local Deltachi2 minima at higher and lower Deltam21(2) are disfavored at >4sigma. Combining with solar neutrino data, we obtain Deltam21(2)=7.59(-0.21)+0.21 x 10(-5) eV2 and tan2theta12=0.47(-0.05)+0.06.

Pu-Zr alloy for high-temperature foil-type fuel

DOEpatents

McCuaig, Franklin D.

1977-01-01

A nuclear reactor fuel alloy consists essentially of from slightly greater than 7 to about 4 w/o zirconium, balance plutonium, and is characterized in that the alloy is castable and is rollable to thin foils. A preferred embodiment of about 7 w/o zirconium, balance plutonium, has a melting point substantially above the melting point of plutonium, is rollable to foils as thin as 0.0005 inch thick, and is compatible with cladding material when repeatedly cycled to temperatures above 650.degree. C. Neutron reflux densities across a reactor core can be determined with a high-temperature activation-measurement foil which consists of a fuel alloy foil core sandwiched and sealed between two cladding material jackets, the fuel alloy foil core being a 7 w/o zirconium, plutonium foil which is from 0.005 to 0.0005 inch thick.

Pu-ZR Alloy high-temperature activation-measurement foil

DOEpatents

McCuaig, Franklin D.

1977-08-02

A nuclear reactor fuel alloy consists essentially of from slightly greater than 7 to about 4 w/o zirconium, balance plutonium, and is characterized in that the alloy is castable and is rollable to thin foils. A preferred embodiment of about 7 w/o zirconium, balance plutonium, has a melting point substantially above the melting point of plutonium, is rollable to foils as thin as 0.0005 inch thick, and is compatible with cladding material when repeatedly cycled to temperatures above 650.degree. C. Neutron flux densities across a reactor core can be determined with a high-temperature activation-measurement foil which consists of a fuel alloy foil core sandwiched and sealed between two cladding material jackets, the fuel alloy foil core being a 7 w/o zirconium, plutonium foil which is from 0.005 to 0.0005 inch thick.

Fast-spectrum space-power-reactor concepts using boron control devices

NASA Technical Reports Server (NTRS)

Mayo, W.

1973-01-01

Several fast-spectrum space power reactor concepts that use boron carbide control devices were examined to determine the neutronic feasibility of the designs. The designs considered were (1) a 199-fuel-pin, 12-poison-reflector-control-drum reactor; (2) a 232-fuel-pin reactor with 12 reflector drums and three in-core control rods; (3) a 337-fuel-pin design with 12 incore control rods; and a 181-fuel-pin design with six drums closely coupled to the core to increase reactivity per drum. Adequate reactivity control and excess reactivity could be obtained for each concept, and the goals of 50,000 hours at 2.17 thermal megawatts with a lithium-7 coolant outlet temperature of 1222 K could be met without exceeding the 1-percent-clad-creep criterion. Heating rates in the boron carbide were calculated, but a heat transfer analysis was not done.

Lunar in-core thermionic nuclear reactor power system conceptual design

NASA Technical Reports Server (NTRS)

Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.

1991-01-01

This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Exploration Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

Estimated inventory of radionuclides in former Soviet Union naval reactors dumped in the Kara Sea

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

Mount, M.E.; Sheaffer, M.K.; Abbott, D.T.

1993-07-01

Radionuclide inventories have been estimated for the reactor cores, reactor components, and primary system corrosion products in the former Soviet Union naval reactors dumped at the Abrosimov Inlet, Tsivolka Inlet, Stepovoy Inlet, Techeniye Inlet, and Novaya Zemlya Depression sites in the Kara Sea between 1965 and 1988. For the time of disposal, the inventories are estimated at 69 to 111 kCi of actinides plus daughters and 3,053 to 7,472 kCi of fission products in the reactor cores, 917 to 1,127 kCi of activation products in the reactor components, and 1.4 to 1.6 kCi of activation products in the primary systemmore » corrosion products. At the present time, the inventories are estimated to have decreased to 23 to 38 kCi of actinides plus daughters and 674 to 708 kCi of fission products in the reactor cores, 124 to 126 kCi of activation products in the reactor components, and 0.16 to 0.17 kCi of activation products in the primary system corrosion products. Twenty years from now, the inventories are projected to be 11 to 18 kCi of actinides plus daughters and 415 to 437 kCi of fission products in the reactor cores, 63.5 to 64 kCi of activation products in the reactor components, and 0.014 to 0.015 kCi of activation products in the primary system corrosion products. All actinide activities are estimated to be within a factor of two.« less

Comparison of two-stage thermophilic (68 degrees C/55 degrees C) anaerobic digestion with one-stage thermophilic (55 degrees C) digestion of cattle manure.

PubMed

Nielsen, H B; Mladenovska, Z; Westermann, P; Ahring, B K

2004-05-05

A two-stage 68 degrees C/55 degrees C anaerobic degradation process for treatment of cattle manure was studied. In batch experiments, an increase of the specific methane yield, ranging from 24% to 56%, was obtained when cattle manure and its fractions (fibers and liquid) were pretreated at 68 degrees C for periods of 36, 108, and 168 h, and subsequently digested at 55 degrees C. In a lab-scale experiment, the performance of a two-stage reactor system, consisting of a digester operating at 68 degrees C with a hydraulic retention time (HRT) of 3 days, connected to a 55 degrees C reactor with 12-day HRT, was compared with a conventional single-stage reactor running at 55 degrees C with 15-days HRT. When an organic loading of 3 g volatile solids (VS) per liter per day was applied, the two-stage setup had a 6% to 8% higher specific methane yield and a 9% more effective VS-removal than the conventional single-stage reactor. The 68 degrees C reactor generated 7% to 9% of the total amount of methane of the two-stage system and maintained a volatile fatty acids (VFA) concentration of 4.0 to 4.4 g acetate per liter. Population size and activity of aceticlastic methanogens, syntrophic bacteria, and hydrolytic/fermentative bacteria were significantly lower in the 68 degrees C reactor than in the 55 degrees C reactors. The density levels of methanogens utilizing H2/CO2 or formate were, however, in the same range for all reactors, although the degradation of these substrates was significantly lower in the 68 degrees C reactor than in the 55 degrees C reactors. Temporal temperature gradient electrophoresis profiles (TTGE) of the 68 degrees C reactor demonstrated a stable bacterial community along with a less divergent community of archaeal species. Copyright 2004 Wiley Periodicals, Inc.

New semi-pilot-scale reactor to study the photocatalytic inactivation of phages contained in aerosol.

PubMed

Briggiler Marcó, Mariángeles; Negro, Antonio Carlos; Alfano, Orlando Mario; Quiberoni, Andrea Del Luján

2017-04-12

The aims of this work were to design and build a photocatalytic reactor (UV-A/TiO 2 ) to study the inactivation of phages contained in bioaerosols, which constitute the main dissemination via phages in industrial environments. The reactor is a close system with recirculation that consists of a stainless steel camera (cubic form, side of 60 cm) in which air containing the phage particles circulates and an acrylic compartment with six borosilicate plates covered with TiO 2 . The reactor is externally illuminated by 20 UV-A lamps. Both compartments are connected by a fan to facilitate the sample circulation. Samples are injected into the camera using two piston nebulizers working in series whereas several methodologies for sampling (impinger/syringe, sampling on photocatalytic plates, and impact of air on slide) were assayed. The reactor setup was carried out using phage B1 (Lactobacillus plantarum), and assays demonstrated a decrease of phage counts of 2.7 log orders after 1 h of photocatalytic treatment. Photonic efficiencies of inactivation were assessed by phage sampling on the photocatalytic plates or by impact of air on a glass slide at the photocatalytic reactor exit. Efficiencies of the same order of magnitude were observed using both sampling methods. This study demonstrated that the designed photocatalytic reactor is effective to inactivate phage B1 (Lb. plantarum) contained in bioaerosols.

The present situations and perspectives on utilization of research reactors in Thailand

NASA Astrophysics Data System (ADS)

Chongkum, Somporn

2002-01-01

The Thai Research Reactor 1/Modification 1, a TRIGA Mark III reactor, went critical on November 7, 1977. It has been playing a central role in the development of both Office of Atomic Energy for Peace (OAEP) and nuclear application in Thailand. It has a maximum power of 2 MW (thermal) at steady state and a pulsing capacity of 2000 MW. The highest thermal neutron flux at a central thimber is 1×10 13 n/cm 2/s, which is extensively utilized for radioisotope production, neutron activation analysis and neutron beam experiments, i.e. neutron scattering, prompt gamma analysis and neutron radiography. Following the nuclear technological development, the OAEP is in the process of establishing the Ongkharak Nuclear Research Center (ONRC). The center is being built in Nakhon Nayok province, 60 km northeast of Bangkok. The centerpiece of the ONRC is a multipurpose 10 MW TRIGA research reactor. Facilities are included for the production of radioisotopes for medicine, industry and agriculture, neutron transmutation doping of silicon, and neutron capture therapy. The neutron beam facilities will also be utilized for applied research and technology development as well as training in reactor operations, performance of experiments and reactor physics. This paper describes a recent program of utilization as well as a new research reactor for enlarging the perspectives of its utilization in the future.

Isotopic composition and neutronics of the Okelobondo natural reactor

NASA Astrophysics Data System (ADS)

Palenik, Christopher Samuel

The Oklo-Okelobondo and Bangombe uranium deposits, in Gabon, Africa host Earth's only known natural nuclear fission reactors. These 2 billion year old reactors represent a unique opportunity to study used nuclear fuel over geologic periods of time. The reactors in these deposits have been studied as a means by which to constrain the source term of fission product concentrations produced during reactor operation. The source term depends on the neutronic parameters, which include reactor operation duration, neutron flux and the neutron energy spectrum. Reactor operation has been modeled using a point-source computer simulation (Oak Ridge Isotope Generation and Depletion, ORIGEN, code) for a light water reactor. Model results have been constrained using secondary ionization mass spectroscopy (SIMS) isotopic measurements of the fission products Nd and Te, as well as U in uraninite from samples collected in the Okelobondo reactor zone. Based upon the constraints on the operating conditions, the pre-reactor concentrations of Nd (150 ppm +/- 75 ppm) and Te (<1 ppm) in uraninite were estimated. Related to the burnup measured in Okelobondo samples (0.7 to 13.8 GWd/MTU), the final fission product inventories of Nd (90 to 1200 ppm) and Te (10 to 110 ppm) were calculated. By the same means, the ranges of all other fission products and actinides produced during reactor operation were calculated as a function of burnup. These results provide a source term against which the present elemental and decay abundances at the fission reactor can be compared. Furthermore, they provide new insights into the extent to which a "fossil" nuclear reactor can be characterized on the basis of its isotopic signatures. In addition, results from the study of two other natural systems related to the radionuclide and fission product transport are included. A detailed mineralogical characterization of the uranyl mineralogy at the Bangombe uranium deposit in Gabon, Africa was completed to improve geochemical models of the solubility-limiting phase. A study of the competing effects of radiation damage and annealing in a U-bearing crystal of zircon shows that low temperature annealing in actinide-bearing phases is significant in the annealing of radiation damage.

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

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

DOE PAGES

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; ...

2016-07-15

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

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

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

ETRMTR MECHANICAL SERVICES BUILDING, TRA653, INTERIOR. CAMERA IS INSIDE MEN'S ...

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

ETR-MTR MECHANICAL SERVICES BUILDING, TRA-653, INTERIOR. CAMERA IS INSIDE MEN'S LAVATORY AND SHOWER FACING SOUTHEAST. SHOWER AND TOILET STALLS ARE IN PLACE. ROUND COMMUNAL SINK AT LEFT. INL NEGATIVE NO. 57-3652. K. Mansfield, Photographer, 7/22/1957 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR WING, TRA604. BASEMENT FLOOR PLAN. FIREPROOF RECORD ROOM BELOW ...

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

MTR WING, TRA-604. BASEMENT FLOOR PLAN. FIRE-PROOF RECORD ROOM BELOW COUNTING ROOM. HEATING AND COOLING EQUIPMENT. UNSPECIFIED EXPANSION AREA ALONG WEST WALL. BLAW-KNOX 3150-4-1, 7/1950. INL INDEX NO. 531-0604-00-098-100007, REV. 1. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. CAMERA LOOKING EAST AND TO WEST ...

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

PROCESS WATER BUILDING, TRA-605. CAMERA LOOKING EAST AND TO WEST WALL NOW ENCLOSING FLASH EVAPORATORS. PIPES IN FOREGROUND WILL CARRY DEMINERALIZED COOLING WATER TO AND FROM THE MTR. INL NEGATIVE NO. 2937. Unknown Photographer, 7/30/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Investigation on the quality of bio-oil produced through fast pyrolysis of biomass-polymer waste mixture

NASA Astrophysics Data System (ADS)

Jourabchi, S. A.; Ng, H. K.; Gan, S.; Yap, Z. Y.

2016-06-01

A high-impact poly-styrene (HIPS) was mixed with dried and ground coconut shell (CS) at equal weight percentage. Fast pyrolysis was carried out on the mixture in a fixed bed reactor over a temperature range of 573 K to 1073 K, and a nitrogen (N2) linear velocity range of 7.8x10-5 m/s to 6.7x10-2 m/s to produce bio-oil. Heat transfer and fluid dynamics of the pyrolysis process inside the reactor was visualised by using Computational Fluid Dynamics (CFD). The CFD modelling was validated by experimental results and they both indicated that at temperature of 923 K and N2 linear velocity of 7.8x10-5 m/s, the maximum bio-oil yield of 52.02 wt% is achieved.

The PROSPECT physics program

DOE PAGES

Ashenfelter, J.; Balantekin, A. B.; Band, H. R.; ...

2016-10-17

The precision reactor oscillation and spectrum experiment, PROSPECT, is designed to make a precise measurement of the antineutrino spectrum from a highly-enriched uranium reactor and probe eV-scale sterile neutrinos by searching for neutrino oscillations over a distance of several meters. The subject of this paper, PROSPECT, is conceived as a 2-phase experiment utilizing segmented 6Li-doped liquid scintillator detectors for both efficient detection of reactor antineutrinos through the inverse beta decay reaction and excellent background discrimination. PROSPECT Phase I consists of a movable 3 ton antineutrino detector at distances of 7–12 m from the reactor core. It will probe the best-fitmore » point of the ν e disappearance experiments at 4σ in 1 year and the favored region of the sterile neutrino parameter space at > 3σ in 3 years. With a second antineutrino detector at 15–19 m from the reactor, Phase II of PROSPECT can probe the entire allowed parameter space below 10 eV 2 at 5σ in 3 additional years. Finally, the measurement of the reactor antineutrino spectrum and the search for short-baseline oscillations with PROSPECT will test the origin of the spectral deviations observed in recent θ 13 experiments, search for sterile neutrinos, and conclusively address the hypothesis of sterile neutrinos as an explanation of the reactor anomaly.« less

Molecular Ecology of Bacterial Populations in Environmental Hazardous Chemical Control

DTIC Science & Technology

1991-11-30

Reactor Figure 1. A schematic drawing of the bioreactor system for on-line studies of naphthalene degradation and light production by bioluminescent...the bioluminescent monitoring section. The reactor system consisted of a L. H. Fermentation Series 500 continuous flow bioreactor with a 1 L glass... studied the expression of the upper pathway operon of NAH7. Light induction in response to naphthalene in the strain HK44 was comparable in both

Formation of aerobic granular sludge during the treatment of petrochemical wastewater.

PubMed

Caluwé, Michel; Dobbeleers, Thomas; D'aes, Jolien; Miele, Solange; Akkermans, Veerle; Daens, Dominique; Geuens, Luc; Kiekens, Filip; Blust, Ronny; Dries, Jan

2017-08-01

In this study, petrochemical wastewater from the port of Antwerp was used for the development of aerobic granular sludge. Two different reactor setups were used, (1) a completely aerated sequencing batch reactor (SBR ae ) with a feast/famine regime and (2) a sequencing batch reactor operated with an anaerobic feast/aerobic famine strategy (SBR an ). The seed sludge showed poor settling characteristics with a sludge volume index (SVI) of 285mL.gMLSS -1 and a median particle size by volume of 86.0µm±1.9µm. In both reactors, granulation was reached after 30days with a SVI of 71mL.gMLSS -1 and median granule size of 264.7µm in SBR an and a SVI of 56mL.gMLSS -1 and median granule size of 307.4µm in SBR ae . The chemical oxygen demand (COD) and dissolved organic carbon (DOC) removal was similar in both reactors and above 95%. The anaerobic DOC uptake increased from 0.13% to 43.2% in 60days in SBR an . Copyright © 2017 Elsevier Ltd. All rights reserved.

10 CFR 72.7 - Specific exemptions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Specific exemptions. 72.7 Section 72.7 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.7...

10 CFR 72.7 - Specific exemptions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Specific exemptions. 72.7 Section 72.7 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.7...

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

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

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

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

Graslund, C.; Hellstrand, E.

Sweden benefits in many ways from the reactor safety research performed in other countries. Its own activity complements this effort, but a certain fraction is oriented toward safety issues that are intimately related to the special design of the ASEA-ATOM boiling-water reactor. Through the availability of the decommissioned Marviken reactor plant, Sweden has been able to play a leading role in integral containment experiments with international participation. Joint efforts with other countries are now devoted to defining new large-scale experiments to be performed in the unique Marviken facility. The largest portion of the safety research program in Sweden is performedmore » by Studsvik Energiteknik AB, but various universities, consultant firms, and research institutes are also involved. In addition, a substantial amount of work is done by the reactor vendor ASEA-ATOM. The overall annual budget is at present between $7 and $8 million, with three governmental authorities as the main financing bodies.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

DOE PAGES

Collette, R.; King, J.; Buesch, C.; ...

2016-04-01

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

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

Collette, R.; King, J.; Buesch, C.

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

The effect of catalyst length and downstream reactor distance on catalytic combustor performance

NASA Technical Reports Server (NTRS)

Anderson, D.

1980-01-01

A study was made to determine the effects on catalytic combustor performance which resulted from independently varying the length of a catalytic reactor and the length available for gas-phase reactions downstream of the catalyst. Monolithic combustion catalysts from three manufacturers were tested in a combustion test rig with no. 2 diesel fuel. Catalytic reactor lengths of 2.5 and 5.4 cm, and downstream gas-phase reaction distances of 7.3, 12.4, 17.5, and 22.5 cm were evaluated. Measurements of carbon monoxide, unburned hydrocarbons, nitrogen oxides, and pressure drop were made. The catalytic-reactor pressure drop was less than 1 percent of the upstream total pressure for all test configurations and test conditions. Nitrogen oxides and unburned hydrocarbons emissions were less than 0.25 g NO2/kg fuel and 0.6 g HC/kg fuel, respectively. The minimum operating temperature (defined as the adiabatic combustion temperature required to obtain carbon monoxide emissions below a reference level of 13.6 g CO/kg fuel) ranged from 1230 K to 1500 K for the various conditions and configurations tested. The minimum operating temperature decreased with increasing total (catalytic-reactor-plus-downstream-gas-phase-reactor-zone) residence time but was independent of the relative times spent in each region when the catalytic-reactor residence time was greater than or equal to 1.4 ms.

External events analysis for the Savannah River Site K reactor

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

Brandyberry, M.D.; Wingo, H.E.

1990-01-01

The probabilistic external events analysis performed for the Savannah River Site K-reactor PRA considered many different events which are generally perceived to be external'' to the reactor and its systems, such as fires, floods, seismic events, and transportation accidents (as well as many others). Events which have been shown to be significant contributors to risk include seismic events, tornados, a crane failure scenario, fires and dam failures. The total contribution to the core melt frequency from external initiators has been found to be 2.2 {times} 10{sup {minus}4} per year, from which seismic events are the major contributor (1.2 {times} 10{supmore » {minus}4} per year). Fire initiated events contribute 1.4 {times} 10{sup {minus}7} per year, tornados 5.8 {times} 10{sup {minus}7} per year, dam failures 1.5 {times} 10{sup {minus}6} per year and the crane failure scenario less than 10{sup {minus}4} per year to the core melt frequency. 8 refs., 3 figs., 5 tabs.« less

Startup and oxygen concentration effects in a continuous granular mixed flow autotrophic nitrogen removal reactor.

PubMed

Varas, Rodrigo; Guzmán-Fierro, Víctor; Giustinianovich, Elisa; Behar, Jack; Fernández, Katherina; Roeckel, Marlene

2015-08-01

The startup and performance of the completely autotrophic nitrogen removal over nitrite (CANON) process was tested in a continuously fed granular bubble column reactor (BCR) with two different aeration strategies: controlling the oxygen volumetric flow and oxygen concentration. During the startup with the control of oxygen volumetric flow, the air volume was adjusted to 60mL/h and the CANON reactor had volumetric N loadings ranging from 7.35 to 100.90mgN/Ld with 36-71% total nitrogen removal and high instability. In the second stage, the reactor was operated at oxygen concentrations of 0.6, 0.4 and 0.2mg/L. The best condition was 0.2 mgO2/L with a total nitrogen removal of 75.36% with a CANON reactor activity of 0.1149gN/gVVSd and high stability. The feasibility and effectiveness of CANON processes with oxygen control was demonstrated, showing an alternative design tool for efficiently removing nitrogen species. Copyright © 2015 Elsevier Ltd. All rights reserved.

Anaerobic treatment of sludge from a nitrification-denitrification landfill leachate plant

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

Maranon, E.; Castrillon, L.; Fernandez, Y.

2006-07-01

The viability of anaerobic digestion of sludge from a MSW landfill leachate treatment plant, with COD values ranging between 15,000 and 19,400 mg O{sub 2} dm{sup -3}, in an upflow anaerobic sludge blanket reactor was studied. The reactor employed had a useful capacity of 9 l, operating at mesophilic temperature. Start-up of the reactor was carried out in different steps, beginning with diluted sludge and progressively increasing the amount of sludge fed into the reactor. The study was carried out over a period of 7 months. Different amounts of methanol were added to the feed, ranging between 6.75 and 1more » cm{sup 3} dm{sup -3} of feed in order to favour the growth of methanogenic flora. The achieved biodegradation of the sludge using an upflow anaerobic sludge blanket Reactor was very high for an HRT of 9 days, obtaining decreases in COD of 84-87% by the end of the process. Purging of the digested sludge represented {approx}16% of the volume of the treated sludge.« less

Biodegradation of pharmaceuticals in hospital wastewater by a hybrid biofilm and activated sludge system (Hybas).

PubMed

Escolà Casas, Mònica; Chhetri, Ravi Kumar; Ooi, Gordon; Hansen, Kamilla M S; Litty, Klaus; Christensson, Magnus; Kragelund, Caroline; Andersen, Henrik R; Bester, Kai

2015-10-15

Hospital wastewater contributes a significant input of pharmaceuticals into municipal wastewater. The combination of suspended activated sludge and biofilm processes, as stand-alone or as hybrid process (hybrid biofilm and activated sludge system (Hybas™)) has been suggested as a possible solution for hospital wastewater treatment. To investigate the potential of such a hybrid system for the removal of pharmaceuticals in hospital wastewater a pilot plant consisting of a series of one activated sludge reactor, two Hybas™ reactors and one moving bed biofilm reactor (MBBR) has been established and adapted during 10 months of continuous operation. After this adaption phase batch and continuous experiments were performed for the determination of degradation of pharmaceuticals. Removal of organic matter and nitrification mainly occurred in the first reactor. Most pharmaceuticals were removed significantly. The removal of pharmaceuticals (including X-ray contrast media, β-blockers, analgesics and antibiotics) was fitted to a single first-order kinetics degradation function, giving degradation rate constants from 0 to 1.49 h(-1), from 0 to 7.78 × 10(-1)h(-1), from 0 to 7.86 × 10(-1)h(-1) and from 0 to 1.07 × 10(-1)h(-1) for first, second, third and fourth reactors respectively. Generally, the highest removal rate constants were found in the first and third reactors while the lowest were found in the second one. When the removal rate constants were normalized to biomass amount, the last reactor (biofilm only) appeared to have the most effective biomass in respect to removing pharmaceuticals. In the batch experiment, out of 26 compounds, 16 were assessed to degrade more than 20% of the respective pharmaceutical within the Hybas™ train. In the continuous flow experiments, the measured removals were similar to those estimated from the batch experiments, but the concentrations of a few pharmaceuticals appeared to increase during the first treatment step. Such increase could be attributed to de-conjugation or formation from other metabolites. Copyright © 2015 Elsevier B.V. All rights reserved.

Sequencing batch reactor biofilm system for treatment of milk industry wastewater.

PubMed

Sirianuntapiboon, Suntud; Jeeyachok, Narumon; Larplai, Rarintorn

2005-07-01

A sequencing batch reactor biofilm (MSBR) system was modified from the conventional sequencing batch reactor (SBR) system by installing 2.7 m2 surface area of plastic media on the bottom of the reactor to increase the system efficiency and bio-sludge quality by increasing the bio-sludge in the system. The COD, BOD5, total kjeldahl nitrogen (TKN) and oil & grease removal efficiencies of the MSBR system, under a high organic loading of 1340 g BOD5/m3 d, were 89.3+/-0.1, 83.0+/-0.2, 59.4+/-0.8, and 82.4+/-0.4%, respectively, while they were only 87.0+/-0.2, 79.9+/-0.3, 48.7+/-1.7 and 79.3+/-10%, respectively, in the conventional SBR system. The amount of excess bio-sludge in the MSBR system was about 3 times lower than that in the conventional SBR system. The sludge volume index (SVI) of the MSBR system was lower than 100 ml/g under an organic loading of up to 1340 g BOD5/m3 d. However, the MSBR under an organic loading of 680 g BOD5/m3 d gave the highest COD, BOD5, TKN and oil & grease removal efficiencies of 97.9+/-0.0, 97.9+/-0.1, 79.3+/-1.0 and 94.8+/-0.5%, respectively, without any excess bio-sludge waste. The SVI of suspended bio-sludge in the MSBR system was only 44+/-3.4 ml/g under an organic loading of 680 g BOD5/m3 d.

Optimization of arsenic removal water treatment system through characterization of terminal electron accepting processes.

PubMed

Upadhyaya, Giridhar; Clancy, Tara M; Brown, Jess; Hayes, Kim F; Raskin, Lutgarde

2012-11-06

Terminal electron accepting process (TEAP) zones developed when a simulated groundwater containing dissolved oxygen (DO), nitrate, arsenate, and sulfate was treated in a fixed-bed bioreactor system consisting of two reactors (reactors A and B) in series. When the reactors were operated with an empty bed contact time (EBCT) of 20 min each, DO-, nitrate-, sulfate-, and arsenate-reducing TEAP zones were located within reactor A. As a consequence, sulfate reduction and subsequent arsenic removal through arsenic sulfide precipitation and/or arsenic adsorption on or coprecipitation with iron sulfides occurred in reactor A. This resulted in the removal of arsenic-laden solids during backwashing of reactor A. To minimize this by shifting the sulfate-reducing zone to reactor B, the EBCT of reactor A was sequentially lowered from 20 min to 15, 10, and 7 min. While 50 mg/L (0.81 mM) nitrate was completely removed at all EBCTs, more than 90% of 300 μg/L (4 μM) arsenic was removed with the total EBCT as low as 27 min. Sulfate- and arsenate-reducing bacteria were identified throughout the system through clone libraries and quantitative PCR targeting the 16S rRNA, dissimilatory (bi)sulfite reductase (dsrAB), and dissimilatory arsenate reductase (arrA) genes. Results of reverse transcriptase (RT) qPCR of partial dsrAB (i.e., dsrA) and arrA transcripts corresponded with system performance. The RT qPCR results indicated colocation of sulfate- and arsenate-reducing activities, in the presence of iron(II), suggesting their importance in arsenic removal.

Analysis of JKT01 Neutron Flux Detector Measurements In RSG-GAS Reactor Using LabVIEW

NASA Astrophysics Data System (ADS)

Rokhmadi; Nur Rachman, Agus; Sujarwono; Taryo, Taswanda; Sunaryo, Geni Rina

2018-02-01

The RSG-GAS Reactor, one of the Indonesia research reactors and located in Serpong, is owned by the National Nuclear Energy Agency (BATAN). The RSG-GAS reactor has operated since 1987 and some instrumentation and control systems are considered to be degraded and ageing. It is therefore, necessary to evaluate the safety of all instrumentation and controls and one of the component systems to be evaluated is the performance of JKT01 neutron flux detector. Neutron Flux Detector JKT01 basically detects neutron fluxes in the reactor core and converts it into electrical signals. The electrical signal is then forwarded to the amplifier (Amplifier) to become the input of the reactor protection system. One output of it is transferred to the Main Control Room (RKU) showing on the analog meter as an indicator used by the reactor operator. To simulate all of this matter, a program to simulate the output of the JKT01 Neutron Flux Detector using LabVIEW was developed. The simulated data is estimated using a lot of equations also formulated in LabVIEW. The calculation results are also displayed on the interface using LabVIEW available in the PC. By using this simulation program, it is successful to perform anomaly detection experiments on the JKT01 detector of RSG-GAS Reactor. The simulation results showed that the anomaly JKT01 neutron flux using electrical-current-base are respectively, 1.5×,1.7× and 2.0×.

Ecological and toxicological aspects of the partial meltdown of the Chernobyl nuclear power plant reactor

USGS Publications Warehouse

Eisler, Ronald; Hoffman, David J.; Rattner, Barnett A.; Burton, G. Allen; Cairns, John

1995-01-01

the partial meltdown of the 1000-MW reactor at Chernobyl, Ukraine, on April 26, 1986, released large amounts of radiocesium and other radionuclides into the environment, causing widespread radioactive contamination of Europe and the former Soviet Union.1-7 At least 3,000,000 trillion becquerels (TBq) were released from the fuel during the accident (Table 24.1), dwarfing, by orders of magnitude, radiation released from other highly publicized reactor accidents at Windscale (U.K.) and three-Mile Island (U.S.)3,8 The Chernobyl accident happened while a test was being conducted during a normal scheduled shutdown and is attributed mainly to human error.3

Influence of mass transfer on the ozonation of wastewater from the glass fiber industry.

PubMed

Byun, S; Cho, S H; Yoon, J; Geissen, S U; Vogelpohl, A; Kim, S M

2004-01-01

The mass transfer rate (kLa) is one of the most important parameters in the ozonation of wastewater, because it frequently constitutes the rate-determining step. This study investigated the influence of kLa on the ozonation of glass fiber wastewater using a high-performance jet loop reactor (HJLR), which is well known for its high mass transfer property, and compared the results of this investigation with those obtained using the bubble column reactor. It was found that the higher kLa achieved by increasing the energy input did not lead to higher ozonation efficiency, since the reaction involving the OH radical was greatly hindered at the low pH produced as a result of ozonation. By maintaining the pH at a value greater than 8.0, the higher kLa in the HJLR reactor contributed to increasing not only the TOC removal of wastewater, but also the ozone consumption efficiency, as expressed by the specific ozone consumption. The specific ozone consumption in the HJLR reactor (7.1 g ozone/ g TOC) was 20% better than that in the bubble column reactor.

FAST NEUTRONIC REACTOR

DOEpatents

Snell, A.H.

1957-12-01

This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.

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

Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

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

A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

2010-06-01

2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has beenmore » restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.« less

Characterization of efficient aerobic denitrifiers isolated from two different sequencing batch reactors by 16S-rRNA analysis.

PubMed

Wang, Ping; Li, Xiuting; Xiang, Mufei; Zhai, Qian

2007-06-01

By adopting two sequencing batch reactors (SBRs) A and B, nitrate as the substrate, and the intermittent aeration mode, activated sludge was domesticated to enrich aerobic denitrifiers. The pHs of reactor A were approximately 6.3 at DOs 2.2-6.1 mg/l for a carbon source of 720 mg/l COD; the pHs of reactor B were 6.8-7.8 at DOs 2.2-3.0 mg/l for a carbon source of 1500 mg/l COD. Both reactors maintained an influent nitrate concentration of 80 mg/l NO3- -N. When the total inorganic nitrogen (TIN) removal efficiency of both reactors reached 60%, aerobic denitrifier accumulation was regarded completed. By bromthymol blue (BTB) medium, 20 bacteria were isolated from the two SBRs and DNA samples of 8 of these 20 strains were amplified by PCR and processed for 16SrRNA sequencing. The obtained results were analysed by a Blast similarity search of the GenBank database, and constructing a phylogenetic tree for identification by comparison. The 8 bacteria were found to belong to the genera Pseudomonas, Delftia, Herbaspirillum and Comamonas. At present, no Delftia has been reported to be an aerobic denitrifier.

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.

Biogas production from municipal solid wastes using an integrated rotary drum and anaerobic-phased solids digester system.

PubMed

Zhu, Baoning; Zhang, Ruihong; Gikas, Petros; Rapport, Joshua; Jenkins, Bryan; Li, Xiujin

2010-08-01

This research was conducted to develop an integrated rotary drum reactor (RDR)-anaerobic-phased solids (APS) digester system for the treatment of municipal solid waste (MSW) to produce biogas energy and achieve waste reduction. A commercial RDR facility was used to provide a 3-d pretreatment and sufficient separation of the organics from MSW and then the organics were digested in a laboratory APS-digester system for biogas production. The organics generated from the RDR contained 50% total solids (TS) and 36% volatile solids (VS) on wet basis. The APS-digester was started at an organic loading rate (OLR) of 3.1 gVS L(-1) d(-1) and operated at three higher OLRs of 4.6, 7.7 and 9.2 gVS L(-1) d(-1). At the OLR of 9.2 gVS L(-1) d(-1) the system biogas production rate was 3.5 L L(-1) d(-1) and the biogas and methane yields were 0.38 and 0.19 L gVS(-1), respectively. Anaerobic digestion resulted in 38% TS reduction and 53% VS reduction in the organic solids. It was found that the total VFA concentration reached a peak value of 15,000 mg L(-1) as acetic acid in the first 3d of batch digestion and later decreased to about 500 mg L(-1). The APS-digester system remained stable at each OLRs for over 100d with the pH in the hydrolysis reactors in the range of 7.3-7.8 and the pH in the biogasification reactor in 7.9-8.1. The residual solids after the digestion had a high heating value of 14.7 kJ gTS(-1). Copyright 2010 Elsevier Ltd. All rights reserved.

Design analysis and risk assessment for a single stage to orbit nuclear thermal rocket

NASA Astrophysics Data System (ADS)

Labib, Satira I.

Recent advances in high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This thesis describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1-15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 700 seconds. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. At the same power level, the 40 cm reactor results in the lowest radiation dose rate of the three reactors. Radiation dose rates decrease to background levels ~3.5 km from the launch site. After a one-year decay time, all of the activated materials produced by an NTR launch would be classified as Class A low-level waste. The activation of air produces significant amounts of argon-41 and nitrogen-16 within 100 m of the launch. The derived air concentration, DAC, from the activation products decays to less than unity within two days, with only argon-41 remaining. After 10 minutes of full power operation the 120 cm core corresponding to a 15 MT payload contains 2.5 x 1013, 1.4 x 1012, 1.5 x 1012, and 7.8 x 10 7 Bq of 131I, 137Cs, 90Sr, and 239Pu respectively. The decay heat after shutdown increases with increasing reactor power with a maximum decay heat of 108 kW immediately after shutdown for the 15 MT payload.

[Comparison between porous polymer carrier and activated carbon carrier used for treating organic wastewater in anaerobic fluidized-bed reactor].

PubMed

Yang, P; Fang, Z; Shi, Y

2001-01-01

A comparative performance between porous polymer carriers (HP) and granular activated carbon carriers (GAC) in anaerobic fluidied-bed reactors was undertaken to evaluate their characters. The results showed that the COD removal and the biogas volume yield rate were 84% and 16.5 m3/(m3.d) respectively when HP was used as carrier to treat synthetic wastewater, at the top COD organic load rate of 65.5 kg/(m3.d), however those were 74.2% and 14.5% respectively for GAC carrier at the top load rate of 63.25 kg/(m3.d). The COD removal and biogas volume yield rate were 64.7%-54.5% and 1.89-2.7 m3/(m3.d) respectively when HP was used as carriers to treat straw pulping wastewater, at the load rate of 14.5-36.15 kg/(m3.d), and those were 61.0%-52.1% and 0.73-2.0 m3/(m3.d) respectively for GAC carriers at the load rate 9.16-19.06 kg/(m3.d). The study revealed that the HP carriers reactor is more efficient than the GAC carriers reactor in microbial immobilization and the wastewater treatment.

Long-term effects of operating temperature and sulphate addition on the methanogenic community structure of anaerobic hybrid reactors.

PubMed

Pender, Seán; Toomey, Margaret; Carton, Micheál; Eardly, Dónal; Patching, John W; Colleran, Emer; O'Flaherty, Vincent

2004-02-01

The diversity, population dynamics, and activity profiles of methanogens in anaerobic granular sludges from two anaerobic hybrid reactors treating a molasses wastewater both mesophilically (37 degrees C) and thermophilically (55 degrees C) during a 1081 day trial were determined. The influent to one of the reactors was supplemented with sulphate, after an acclimation period of 112 days, to determine the effect of competition with sulphate-reducing bacteria on the methanogenic community structure. Sludge samples were removed from the reactors at intervals throughout the operational period and examined by amplified ribosomal DNA (rDNA) restriction analysis (ARDRA) and partial sequencing of 16S rRNA genes. In total, 18 operational taxonomic units (OTUs) were identified, 12 of which were sequenced. The methanogenic communities in both reactors changed during the operational period. The seed sludge and the reactor biomass sampled during mesophilic operation, both in the presence and absence of sulphate, was characterised by a predominance of Methanosaeta spp. Following temperature elevation, the dominant methanogenic sequences detected in the non-sulphate supplemented reactor were closely related to Methanocorpusculum parvum. By contrast, the dominant OTUs detected in the sulphate-supplemented reactor upon temperature increase were related to the hydrogen-utilising methanogen, Methanobacterium thermoautotrophicum. The observed methanogenic community structure in the reactors correlated with the operational performance of the reactors during the trial and with physiological measurements of the reactor biomass. Both reactors achieved chemical oxygen demand (COD) removal efficiencies of over 90% during mesophilic operation, with or without sulphate supplementation. During thermophilic operation, the presence of sulphate resulted in decreased reactor performance (effluent acetate concentrations of >3000 mg/l and biogas methane content of <25%). It was demonstrated that methanogenic conversion of acetate at 55 degrees C was extremely sensitive to inhibition by sulphide (50% inhibition at 8-17 mg/l unionised sulphide at pH 7.6-8.0), while the conversion of H(2)/CO(2) methanogenically was favoured. The combination of experiments carried out demonstrated the presence of specific methanogenic populations during periods of successful operational performance.

PLUG STORAGE BUILDING, TRA611, AWAITS SHIELDING SOIL TO BE PLACED ...

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

PLUG STORAGE BUILDING, TRA-611, AWAITS SHIELDING SOIL TO BE PLACED OVER PLUG STORAGE TUBES. WING WALLS WILL SUPPORT EARTH FILL. MTR, PROCESS WATER BUILDING, AND WORKING RESERVOIR IN VIEW BEYOND PLUG STORAGE. CAMERA FACES NORTHEAST. INL NEGATIVE NO. 2949. Unknown Photographer, 7/30/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ELECTRICAL LINES ARRIVE FROM CENTRAL FACILITIES AREA, SOUTH OF MTR. ...

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

ELECTRICAL LINES ARRIVE FROM CENTRAL FACILITIES AREA, SOUTH OF MTR. EXCAVATION RUBBLE IN FOREGROUND. CONTRACTOR CRAFT SHOPS, CRANES, AND OTHER MATERIALS ON SITE. CAMERA FACES EAST, WITH LITTLE BUTTE AND MIDDLE BUTTE IN DISTANCE. INL NEGATIVE NO. 335. Unknown Photographer, 7/1/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

88. ARAIII. "Petrochem" heater is hoisted over south exterior wall ...

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

88. ARA-III. "Petro-chem" heater is hoisted over south exterior wall of heater pit in GCRE reactor building (ARA-608). Printing on heater says, "Petro-chem iso-flow furnace; American industrial fabrications, inc." Camera facing north. January 7, 1959. Ineel photo no. 529-124. Photographer: Ken Mansfield. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

ETR, TRA642, CAMERA IS BELOW, BUT NEAR THE CEILING OF ...

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

ETR, TRA-642, CAMERA IS BELOW, BUT NEAR THE CEILING OF THE GROUND FLOOR, AND LOOKS DOWN TOWARD THE CONSOLE FLOOR. CAMERA FACES WESTERLY. THE REACTOR PIT IS IN THE CENTER OF THE VIEW. BEYOND IT TO THE LEFT IS THE SOUTH SIDE OF THE WORKING CANAL. IN THE FOREGROUND ON THE RIGHT IS THE SHIELDING FOR THE PROCESS WATER TUNNEL AND PIPING. SPIRAL STAIRCASE AT LEFT OF VIEW. INL NEGATIVE NO. 56-2237. Jack L. Anderson, Photographer, 7/6/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Reactor on-off antineutrino measurement with KamLAND

NASA Astrophysics Data System (ADS)

Gando, A.; Gando, Y.; Hanakago, H.; Ikeda, H.; Inoue, K.; Ishidoshiro, K.; Ishikawa, H.; Koga, M.; Matsuda, R.; Matsuda, S.; Mitsui, T.; Motoki, D.; Nakamura, K.; Obata, A.; Oki, A.; Oki, Y.; Otani, M.; Shimizu, I.; Shirai, J.; Suzuki, A.; Takemoto, Y.; Tamae, K.; Ueshima, K.; Watanabe, H.; Xu, B. D.; Yamada, S.; Yamauchi, Y.; Yoshida, H.; Kozlov, A.; Yoshida, S.; Piepke, A.; Banks, T. I.; Fujikawa, B. K.; Han, K.; O'Donnell, T.; Berger, B. E.; Learned, J. G.; Matsuno, S.; Sakai, M.; Efremenko, Y.; Karwowski, H. J.; Markoff, D. M.; Tornow, W.; Detwiler, J. A.; Enomoto, S.; Decowski, M. P.

2013-08-01

The recent long-term shutdown of Japanese nuclear reactors has resulted in a significantly reduced reactor ν¯e flux at KamLAND. This running condition provides a unique opportunity to confirm and constrain backgrounds for the reactor ν¯e oscillation analysis. The data set also has improved sensitivity for other ν¯e signals, in particular ν¯e’s produced in β-decays from U238 and Th232 within the Earth’s interior, whose energy spectrum overlaps with that of reactor ν¯e’s. Including constraints on θ13 from accelerator and short-baseline reactor neutrino experiments, a combined three-flavor analysis of solar and KamLAND data gives fit values for the oscillation parameters of tan⁡2θ12=0.436-0.025+0.029, Δm212=7.53-0.18+0.18×10-5eV2, and sin⁡2θ13=0.023-0.002+0.002. Assuming a chondritic Th/U mass ratio, we obtain 116-27+28 ν¯e events from U238 and Th232, corresponding to a geo ν¯e flux of 3.4-0.8+0.8×106cm-2s-1 at the KamLAND location. We evaluate various bulk silicate Earth composition models using the observed geo ν¯e rate.

Design test request No. 1263 K Reactor graphite key and VSR channel sleeve test

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

Kempf, F.J.

1964-12-10

The objectives of this test were: (1) Determine the coefficient of friction between two adjacent layers of K Reactor graphite at room temperature. (2) Determine the average load required to cause failure of an unirradiated K Reactor side reflector bar, when subjected to tensile loading applied through the reflector keys. (3) Determine the average load at failure and the average deflection at failure of a single VSR channel key when loaded in keyways with clearances equal to those used in original stack construction. (4) Determine the average load and deflection required to break the four K Reactor VSR keys whenmore » loaded simultaneously in both `3-layer` and `7-layer` mockups. Also determine the mode of key failure; i.e., shear, flexure or combined compression and bending. Following these key rupture tests, determine the strength and deflection characteristics of the proposed K Reactor VSR channel sleeve when loaded in a manner identical to that used to fracture the keys. (5) Determine the average load and deflection at failure of both the proposed K Reactor VSR channel sleeves and the proposed C Reactor sleeves when subjected to crushing loads. (6) Determine the extent of damage to the proposed K Reactor VSR channel sleeve when subjected to the following vertical rod loading conditions. (a) Full rod drop in a channel mockup which has been misaligned 2 1/2 inches. (b) Full rod drop in a channel which has been misaligned an amount equal to the maximum flexibility of a `universal` VSR.« less

Impact of Fission Neutron Energies on Reactor Antineutrino Spectra

NASA Astrophysics Data System (ADS)

Hermanek, Keith; Littlejohn, Bryce; Gustafson, Ian

2017-09-01

Recent measurements of the reactor antineutrino spectra (Double Chooz, Reno, and Daya Bay) have shown a discrepancy in the 5-7 MeV region when compared to current theoretical models (Vogel and Huber-Mueller). There are numerous theories pertaining to this antineutrino anomaly, including theories that point to new physics beyond the standard model. In the paper ``Possible Origins and Implications of the Shoulder in Reactor Neutrino Spectra'' by A. Hayes et al., explanations for this anomaly are suggested. One theory is that there are interactions from fast and epithermal incident neutrons which are significant enough to create more events in the 5-7 MeV by a noticeable amount. In our research, we used the Oklo software network created by Dan Dwyer. This generates ab initio antineutrino and beta decay spectra based on standard fission yield databases ENDF, JENDL, JEFF, and the beta decay transition database ENSDF-6. Utilizing these databases as inputs, we show with reasonable assumptions one can prove contributions of fast and epithermal neutrons is less than 3% in the 5-7 MeV region. We also discovered rare isotopes are present in beta decay chains but not well measured and have no corresponding database information, and studied its effect onto the spectrum.

Measurement of neutrino oscillation with KamLAND: evidence of spectral distortion.

PubMed

Araki, T; Eguchi, K; Enomoto, S; Furuno, K; Ichimura, K; Ikeda, H; Inoue, K; Ishihara, K; Iwamoto, T; Kawashima, T; Kishimoto, Y; Koga, M; Koseki, Y; Maeda, T; Mitsui, T; Motoki, M; Nakajima, K; Ogawa, H; Owada, K; Ricol, J-S; Shimizu, I; Shirai, J; Suekane, F; Suzuki, A; Tada, K; Tajima, O; Tamae, K; Tsuda, Y; Watanabe, H; Busenitz, J; Classen, T; Djurcic, Z; Keefer, G; McKinny, K; Mei, D-M; Piepke, A; Yakushev, E; Berger, B E; Chan, Y D; Decowski, M P; Dwyer, D A; Freedman, S J; Fu, Y; Fujikawa, B K; Goldman, J; Gray, F; Heeger, K M; Lesko, K T; Luk, K-B; Murayama, H; Poon, A W P; Steiner, H M; Winslow, L A; Horton-Smith, G A; Mauger, C; McKeown, R D; Vogel, P; Lane, C E; Miletic, T; Gorham, P W; Guillian, G; Learned, J G; Maricic, J; Matsuno, S; Pakvasa, S; Dazeley, S; Hatakeyama, S; Rojas, A; Svoboda, R; Dieterle, B D; Detwiler, J; Gratta, G; Ishii, K; Tolich, N; Uchida, Y; Batygov, M; Bugg, W; Efremenko, Y; Kamyshkov, Y; Kozlov, A; Nakamura, Y; Gould, C R; Karwowski, H J; Markoff, D M; Messimore, J A; Nakamura, K; Rohm, R M; Tornow, W; Wendell, R; Young, A R; Chen, M-J; Wang, Y-F; Piquemal, F

2005-03-04

We present results of a study of neutrino oscillation based on a 766 ton/year exposure of KamLAND to reactor antineutrinos. We observe 258 nu (e) candidate events with energies above 3.4 MeV compared to 365.2+/-23.7 events expected in the absence of neutrino oscillation. Accounting for 17.8+/-7.3 expected background events, the statistical significance for reactor nu (e) disappearance is 99.998%. The observed energy spectrum disagrees with the expected spectral shape in the absence of neutrino oscillation at 99.6% significance and prefers the distortion expected from nu (e) oscillation effects. A two-neutrino oscillation analysis of the KamLAND data gives Deltam(2)=7.9(+0.6)(-0.5)x10(-5) eV(2). A global analysis of data from KamLAND and solar-neutrino experiments yields Deltam(2)=7.9(+0.6)(-0.5)x10(-5) eV(2) and tan((2)theta=0.40(+0.10)(-0.07), the most precise determination to date.

76 FR 67232 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-31

..., 2011--8:30 a.m. Until 5 p.m. The Subcommittee will review Chapters 7, ``Instrumentation and Controls... Chapter 7, ``Instrumentation and Controls,'' of the Calvert Cliffs RCOL SER with Open Items. The...

Biomass characteristics and simultaneous nitrification-denitrification under long sludge retention time in an integrated reactor treating rural domestic sewage.

PubMed

Gong, Lingxiao; Jun, Li; Yang, Qing; Wang, Shuying; Ma, Bin; Peng, Yongzhen

2012-09-01

In this work, a novel integrated reactor incorporating anoxic fixed bed biofilm reactor (FBBR), oxic moving bed biofilm reactor (MBBR) and settler sequentially was proposed for nitrogen removal from rural domestic sewage. For purposes of achieving high efficiency, low costs and easy maintenance, biomass characteristics and simultaneous nitrification-denitrification (SND) were investigated under long sludge retention time during a 149-day period. The results showed that enhanced SND with proportions of 37.7-42.2% tapped the reactor potentials of efficiency and economy both, despite of C/N ratio of 2.5-4.0 in influent. TN was removed averagely by 69.3% at least, even under internal recycling ratio of 200% and less proportions of biomass assimilation (<3%). Consequently, lower internal recycle and intermittent wasted sludge discharge were feasible to save costs, together with cancellations of sludge return and anoxic stir. Furthermore, biomass with low observed heterotrophic yields (0.053 ± 0.035 g VSS/g COD) and VSS/TSS ratio (<0.55) in MBBR, simplified wasted sludge disposal. Copyright © 2012 Elsevier Ltd. All rights reserved.

Radiation absorption and optimization of solar photocatalytic reactors for environmental applications.

PubMed

Colina-Márquez, Jose; Machuca-Martínez, Fiderman; Li Puma, Gianluca

2010-07-01

This study provides a systematic and quantitative approach to the analysis and optimization of solar photocatalytic reactors utilized in environmental applications such as pollutant remediation and conversion of biomass (waste) to hydrogen. Ray tracing technique was coupled with the six-flux absorption scattering model (SFM) to analyze the complex radiation field in solar compound parabolic collectors (CPC) and tubular photoreactors. The absorption of solar radiation represented by the spatial distribution of the local volumetric rate of photon absorption (LVRPA) depends strongly on catalyst loading and geometry. The total radiation absorbed in the reactors, the volumetric rate of absorption (VRPA), was analyzed as a function of the optical properties (scattering albedo) of the photocatalyst. The VRPA reached maxima at specific catalyst concentrations in close agreement with literature experimental studies. The CPC has on average 70% higher photon absorption efficiency than a tubular reactor and requires 39% less catalyst to operate under optimum conditions. The "apparent optical thickness" is proposed as a new dimensionless parameter for optimization of CPC and tubular reactors. It removes the dependence of the optimum catalyst concentration on tube diameter and photocatalyst scattering albedo. For titanium dioxide (TiO(2)) Degussa P25, maximum photon absorption occurs at apparent optical thicknesses of 7.78 for CPC and 12.97 for tubular reactors.

UF6 breeder reactor power plants for electric power generation

NASA Technical Reports Server (NTRS)

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

1976-01-01

The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

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

Indah Rosidah, M., E-mail: indah.maymunah@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id; Yazid, Putranto Ilham

The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With themore » tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature and high pressure condition for more than ten years.« less

Central nervous system radiation syndrome in mice from preferential 10B(n, alpha)7Li irradiation of brain vasculature

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

Slatkin, D.N.; Stoner, R.D.; Rosander, K.M.

1988-06-01

Ionizing radiations were directed at the heads of anesthetized mice in doses that evoked the acute central nervous system (CNS) radiation syndrome. Irradiations were done using either a predominantly thermal neutron field at a nuclear reactor after intraperitoneal injection of 10B-enriched boric acid or 250-kilovolt-peak x-rays with and without previous intraperitoneal injection of equivalent unenriched boric acid. Since 10B concentrations were approximately equal to 3-fold higher in blood than in cerebral parenchyma during the reactor irradiations, more radiation from alpha and 7Li particles was absorbed by brain endothelial cells than by brain parenchymal cells. Comparison of the LD50 dose formore » CNS radiation lethality from the reactor experiments with the LD50 dose from the x-ray experiments gives results compatible with morphologic evidence that endothelial cell damage is a major determinant of acute lethality from the CNS radiation syndrome. It was also observed that boric acid is a low linear energy transfer radiation-enhancement agent in vivo.« less

Anaerobic treatment of rice winery wastewater in an upflow filter packed with steel slag under different hydraulic loading conditions.

PubMed

Jo, Yeadam; Kim, Jaai; Hwang, Seokhwan; Lee, Changsoo

2015-10-01

Rice-washing drainage (RWD), a strong organic wastewater, was anaerobically treated using an upflow filter filled with blast-furnace slag. The continuous performance of the reactor was examined at varying hydraulic retention times (HRTs). The reactor achieved 91.7% chemical oxygen demand removal (CODr) for a 10-day HRT (0.6 g COD/Ld organic loading rate) and maintained fairly stable performance until the HRT was shortened to 2.2 days (CODr > 84%). Further decreases in HRT caused process deterioration (CODr < 50% and pH < 5.5 for a 0.7-day HRT). The methane production rate increased with decreasing HRT to reach the peak level for a 1.3-day HRT, whereas the yield was significantly greater for 3.4-day or longer HRTs. The substrate removal and methane production kinetics were successfully evaluated, and the generated kinetic models produced good performance predictions. The methanogenic activity of the reactor likely relies on the filter biofilm, with Methanosaeta being the main driver. Copyright © 2015 Elsevier Ltd. All rights reserved.

Comprehensive study for Anammox process via multistage anaerobic baffled reactors

NASA Astrophysics Data System (ADS)

Ismail, Sherif; Tawfik, Ahmed

2017-11-01

Continuous anaerobic ammonia oxidation (Anammox) process in multistage anaerobic baffled (MABR) reactor was investigated. The reactor was operated for approximately 150 days at constant hydraulic retention time (HRT) of 48 h and was fed with synthetic wastewater containing nitrite and ammonium as main substrates. The MABR was inoculated with mixed culture bacteria collected from activated sludge plant (41.6 g MLSS/L and 19.1 g MLVSS/L). The MABR reactor exhibited excellent performance for the start-up of Anammox process within a period of 35 days. The start-up period was divided into four successive phases; cell lysis, lag, activity elevation and steady state. Total inorganic nitrogen (TIN) removal efficiency of 96.8± 0.9% was achieved at steady state conditions, corresponding to nitrogen removal rate (NRR) of 50.2±1.7 mg N/L·d. Moreover, the effect of HRT on the Anammox process was assessed with applying five different HRTs of (48, 38.4, 28.8, 19.2 and 9.6 h). Decreasing HRT from 48 to 9.6 h reduced the removal efficiencies of NH4-N, NO2-N and TIN from 97.7±2.2 to 49.0±9.8%, from 95.7±1.9 to 71.0±8.5% and from 96.8±0.9 to 57.9±9.1%, respectively, that corresponding to reduction in NRR from 50.8±1.2 mg N/L·d at HRT of 48 h to 32.5±5.0 mg N/L·d at HRT of 9.6 h.

Treatment of corn ethanol distillery wastewater using two-stage anaerobic digestion.

PubMed

Ráduly, B; Gyenge, L; Szilveszter, Sz; Kedves, A; Crognale, S

In this study the mesophilic two-stage anaerobic digestion (AD) of corn bioethanol distillery wastewater is investigated in laboratory-scale reactors. Two-stage AD technology separates the different sub-processes of the AD in two distinct reactors, enabling the use of optimal conditions for the different microbial consortia involved in the different process phases, and thus allowing for higher applicable organic loading rates (OLRs), shorter hydraulic retention times (HRTs) and better conversion rates of the organic matter, as well as higher methane content of the produced biogas. In our experiments the reactors have been operated in semi-continuous phase-separated mode. A specific methane production of 1,092 mL/(L·d) has been reached at an OLR of 6.5 g TCOD/(L·d) (TCOD: total chemical oxygen demand) and a total HRT of 21 days (5.7 days in the first-stage, and 15.3 days in the second-stage reactor). Nonetheless the methane concentration in the second-stage reactor was very high (78.9%); the two-stage AD outperformed the reference single-stage AD (conducted at the same reactor loading rate and retention time) by only a small margin in terms of volumetric methane production rate. This makes questionable whether the higher methane content of the biogas counterbalances the added complexity of the two-stage digestion.

Hydrogen sulfide removal from air by Acidithiobacillus thiooxidans in a trickle bed reactor.

PubMed

Ramirez, M; Gómez, J M; Cantero, D; Páca, J; Halecký, M; Kozliak, E I; Sobotka, M

2009-09-01

A strain of Acidithiobacillus thiooxidans immobilized in polyurethane foam was utilized for H(2)S removal in a bench-scale trickle-bed reactor, testing the limits of acidity and SO(4) (2-) accumulation. The use of this acidophilic strain resulted in remarkable stability in the performance of the system. The reactor maintained a >98-99 % H(2)S removal efficiency for c of up to 66 ppmv and empty bed residence time 98 % H(2)S was achieved under steady-state conditions, over the pH range of 0.44-7.30. Despite the accumulation of acidity and SO(4) (2-) (up to 97 g/L), the system operated without inhibition.

Evaluation of different types of anaerobic seed sludge for the high rate anaerobic digestion of pig slurry in UASB reactors.

PubMed

Rico, Carlos; Montes, Jesús A; Rico, José Luis

2017-08-01

Three different types of anaerobic sludge (granular, thickened digestate and anaerobic sewage) were evaluated as seed inoculum sources for the high rate anaerobic digestion of pig slurry in UASB reactors. Granular sludge performance was optimal, allowing a high efficiency process yielding a volumetric methane production rate of 4.1LCH 4 L -1 d -1 at 1.5days HRT (0.248LCH 4 g -1 COD) at an organic loading rate of 16.4gCODL -1 d -1 . The thickened digestate sludge experimented flotation problems, thus resulting inappropriate for the UASB process. The anaerobic sewage sludge reactor experimented biomass wash-out, but allowed high process efficiency operation at 3days HRT, yielding a volumetric methane production rate of 1.7LCH 4 L -1 d -1 (0.236LCH 4 g -1 COD) at an organic loading rate of 7.2gCODL -1 d -1 . To guarantee the success of the UASB process, the settleable solids of the slurry must be previously removed. Copyright © 2017 Elsevier Ltd. All rights reserved.

Evolution of microorganisms in thermophilic-dry anaerobic digestion.

PubMed

Montero, B; Garcia-Morales, J L; Sales, D; Solera, R

2008-05-01

Microbial population dynamics were studied during the start-up and stabilization periods in thermophilic-dry anaerobic digestion at lab-scale. The experimental protocol was defined to quantify Eubacteria and Archaea using Fluorescent in situ hybridization (FISH) in a continuously stirred tank reactor (CSTR), without recycling solids. The reactor was subjected to a programme of steady-state operation over a range of the retention times from 40 to 25 days, with an organic loading rate between 4.42 and 7.50 kg volatile solid/m3/day. Changes in microbial concentrations were linked to traditional performance parameters such as biogas production and VS removal. The relations of Eubacteria:Archaea and H2-utilising methanogens:acetate-utilising methanogens were 88:12 and 11:1, respectively, during start-up stage. Hydrogenotrophic methanogens, although important in the initial phase of the reactor start-up, were displaced by acetoclastic methanogens at steady-state, thus their relation were 7:32, respectively. The methane yield coefficient, the methane content in the biogas and VS removal were stabilized around 0.30 LCH4/gCOD, 50% and 80%, respectively. Methanogenic population correlated well with performance measurements.

Low acetate concentrations favor polyphosphate-accumulating organisms over glycogen-accumulating organisms in enhanced biological phosphorus removal from wastewater.

PubMed

Tu, Yunjie; Schuler, Andrew J

2013-04-16

Glycogen-accumulating organisms (GAOs) are thought to compete with polyphosphate-accumulating organisms (PAOs) in enhanced biological phosphorus removal (EBPR) wastewater treatment systems. A laboratory sequencing batch reactor (SBR) was operated for one year to test the hypothesis that PAOs have a competitive advantage at low acetate concentrations, with a focus on low pH conditions previously shown to favor GAOs. PAOs dominated the system under conventional SBR operation with rapid acetate addition (producing high in-reactor concentrations) and pH values of 7.4-8.4. GAOs dominated when the pH was decreased (6.4-7.0). Decreasing the acetate addition rate led to very low reactor acetate concentrations, and PAOs recovered, supporting the study hypothesis. When the acetate feed rate was increased, EBPR failed again. Dominant PAOs and GAOs were Candidatus Accumulibacter phosphatis and Defluviicoccus Cluster 2, respectively, according to fluorescent in situ hybridization and 454 pyrosequencing. Surprisingly, GAOs were not the immediate causes of PAO failures, based on functional and population measurements. Pyrosequencing results suggested Dechloromonas and Tetrasphaera spp. may have also been PAOs, and additional potential GAOs were also identified. Full-scale systems typically have lower in-reactor acetate concentrations than laboratory SBRs, and so, previous laboratory studies may have overestimated the practical importance of GAOs as causes of EBPR failure.

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

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

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

7. EXTERIOR VIEW TO THE SOUTHWEST OF THE NORTH AND ...

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

7. EXTERIOR VIEW TO THE SOUTHWEST OF THE NORTH AND EAST ELEVATIONS. - Nevada Test Site, Reactor Maintenance Assembly & Dissassembly Facility, Area 25, Jackass Flats, Junction of Roads F & G, Mercury, Nye County, NV

Reactivity change in a fast-spectrum space power reactor due to a 328-meter-per-second (1075-ft/sec) impact

NASA Technical Reports Server (NTRS)

Peoples, J. A., Jr.; Puthoff, R. L.

1973-01-01

Application of nuclear reactors in space will present operational problems. One such problem is the possibility of an earth impact at velocities in excess of 305 m/sec (1000 ft/sec). This report shows the results of an impact against concrete at 328 m/sec (1075 ft/sec) and examines the deformed core to estimate the range of activity inserted as a result of the impact. The results of this examination are that the deformation of the reactor core within the containment vessel left only an estimated 2.7 percent void in the core and that the reactivity inserted due to this impact deformation could be from 4.0 to 10.25 dollars.

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

Pentachlorophenol (PCP) dechlorination in horizontal-flow anaerobic immobilized biomass (HAIB) reactors.

PubMed

Damianovic, M H R Z; Moraes, E M; Zaiat, M; Foresti, E

2009-10-01

This study verifies the potential applicability of horizontal-flow anaerobic immobilized biomass (HAIB) reactors to pentachlorophenol (PCP) dechlorination. Two bench-scale HAIB reactors (R1 and R2) were filled with cubic polyurethane foam matrices containing immobilized anaerobic sludge. The reactors were then continuously fed with synthetic wastewater consisting of PCP, glucose, acetic acid, and formic acid as co-substrates for PCP anaerobic degradation. Before being immobilized in polyurethane foam matrices, the biomass was exposed to wastewater containing PCP in reactors fed at a semi-continuous rate of 2.0 microg PCP g(-1) VS. The applied PCP loading rate was increased from 0.05 to 2.59 mg PCP l(-1)day(-1) for R1, and from 0.06 to 4.15 mg PCP l(-1)day(-1) for R2. The organic loading rates (OLR) were 1.1 and 1.7 kg COD m(-3)day(-1) at hydraulic retention times (HRT) of 24h for R1 and 18 h for R2. Under such conditions, chemical oxygen demand (COD) removal efficiencies of up to 98% were achieved in the HAIB reactors. Both reactors exhibited the ability to remove 97% of the loaded PCP. Dichlorophenol (DCP) was the primary chlorophenol detected in the effluent. The adsorption of PCP and metabolites formed during PCP degradation in the packed bed was negligible for PCP removal efficiency.

Equalization of energy density in boiling water reactors (as exemplified by WB-50). Development and testing of WB -50 computational model on the basis of MCU-RR code

NASA Astrophysics Data System (ADS)

Chertkov, Yu B.; Disyuk, V. V.; Pimenov, E. Yu; Aksenova, N. V.

2017-01-01

Within the framework of research in possibility and prospects of power density equalization in boiling water reactors (as exemplified by WB-50) a work was undertaken to improve prior computational model of the WB-50 reactor implemented in MCU-RR software. Analysis of prior works showed that critical state calculations have deviation of calculated reactivity exceeding ±0.3 % (ΔKef/Kef) for minimum concentrations of boric acid in the reactor water and reaching 2 % for maximum concentration values. Axial coefficient of nonuniform burnup distribution reaches high values in the WB-50 reactor. Thus, the computational model needed refinement to take into account burnup inhomogeneity along the fuel assembly height. At this stage, computational results with mean square deviation of less than 0.7 % (ΔKef/Kef) and dispersion of design values of ±1 % (ΔK/K) shall be deemed acceptable. Further lowering of these parameters apparently requires root cause analysis of such large values and paying more attention to experimental measurement techniques.

Counteracting ammonia inhibition in anaerobic digestion by removal with a hollow fiber membrane contactor.

PubMed

Lauterböck, B; Ortner, M; Haider, R; Fuchs, W

2012-10-01

The aim of the current study was to investigate the feasibility of membrane contactors for continuous ammonia (NH₃-N) removal in an anaerobic digestion process and to counteract ammonia inhibition. Two laboratory anaerobic digesters were fed slaughterhouse wastes with ammonium (NH₄⁺) concentrations ranging from 6 to 7.4 g/L. One reactor was used as reference reactor without any ammonia removal. In the second reactor, a hollow fiber membrane contactor module was used for continuous ammonia removal. The hollow fiber membranes were directly submerged into the digestate of the anaerobic reactor. Sulfuric acid was circulated in the lumen as an adsorbent solution. Using this set up, the NH₄⁺-N concentration in the membrane reactor was significantly reduced. Moreover the extraction of ammonia lowered the pH by 0.2 units. In combination that led to a lowering of the free NH₃-N concentration by about 70%. Ammonia inhibition in the reference reactor was observed when the concentration exceeded 6 g/L NH₄⁺-N or 1-1.2 g/L NH₃-N. In contrast, in the membrane reactor the volatile fatty acid concentration, an indicator for process stability, was much lower and a higher gas yield and better degradation was observed. The chosen approach offers an appealing technology to remove ammonia directly from media having high concentrations of solids and it can help to improve process efficiency in anaerobic digestion of ammonia rich substrates. Copyright © 2012 Elsevier Ltd. All rights reserved.

MTR, TRA603. SECOND FLOOR PLAN. OFFICES AND INSTRUMENT ROOM. STEEL ...

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

MTR, TRA-603. SECOND FLOOR PLAN. OFFICES AND INSTRUMENT ROOM. STEEL PARTITIONS ON EAST SIDE OF INSTRUMENT ROOM. DETAIL OF COLUMN ENCASEMENTS. STAIRWAYS IN NORTH AND SOUTH CORNERS. PASSENGER ELEVATION. BLAW-KNOX 3150-803-3, 7/1950. INL INDEX NO. 531-0603-00-098-100562, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Production of fumaric acid by immobilized Rhizopus arrhizus RH 7-13-9# on loofah fiber in a stirred-tank reactor.

PubMed

Liu, Huan; Zhao, Shijie; Jin, Yuhan; Yue, Xuemin; Deng, Li; Wang, Fang; Tan, Tianwei

2017-11-01

Fumaric acid is an important building-block chemical. The production of fumaric acid by fermentation is possible. Loofah fiber is a natural, biodegradable, renewable polymer material with highly sophisticated and pore structure. This work investigated a new immobilization method using loofah fiber as carrier to produce fumaric acid in a stirred-tank reactor. Compared with other carriers, loofah fiber was proven to be efficiently and successfully used in the reactor. After the optimization process, 20g addition of loofah fiber and 400rpm agitation speed were chosen as the most suitable process conditions. 30.3g/L fumaric acid in the broth as well as 19.16g fumaric acid in the precipitation of solid was achieved, while the yield from glucose reached 0.211g/g. Three batches of fermentation using the same loofah fiber carrier were conducted successfully, which meant it provided a new method to produce fumaric acid in a stirred-tank reactor. Copyright © 2017. Published by Elsevier Ltd.

Remedying acidification and deterioration of aerobic post-treatment of digested effluent by using zero-valent iron.

PubMed

Wang, Shen; Zheng, Dan; Wang, Shuang; Wang, Lan; Lei, Yunhui; Xu, Ze; Deng, Liangwei

2018-01-01

This study presents a novel strategy for remedying acidification and improving the removal efficiency of pollutants from digested effluent by using Zero-Valent Iron (iron scraps) in a sequencing batch reactor. Through this strategy, the pH increased from 5.7 (mixed liquid in the reactor without added ZVI) to 7.8 (reactors with added ZVI) because of Fe 0 oxidation and NO 3 - reduction. The removal efficiencies of COD increased from 11.5% to 77.5% because of oxidation of ferric ion and OH produced in chemical reactions of ZVI with oxygen and because of flocculation of iron ions. The removal efficiencies of total nitrogen rose from 1.83% to 93.3% probably because of autotrophic denitrification using electron donors produced by the corrosion of iron, as well as the favorable conditions for anammox due to iron ions. Total phosphorus increased from -25.8% to 77.1% because of the increase in pH and the precipitation with iron ions. Copyright © 2017 Elsevier Ltd. All rights reserved.

Bioremoval of trivalent chromium using Bacillus biofilms through continuous flow reactor.

PubMed

Sundar, K; Sadiq, I Mohammed; Mukherjee, Amitava; Chandrasekaran, N

2011-11-30

Present study deals with the applicability of bacterial biofilms for the bioremoval of trivalent chromium from tannery effluents. A continuous flow reactor was designed for the development of biofilms on different substrates like glass beads, pebbles and coarse sand. The parameters for the continuous flow reactor were 20 ml/min flow rate at 30°C, pH4. Biofilm biomass on the substrates was in the following sequence: coarse sand>pebbles>glass beads (4.8 × 10(7), 4.5 × 10(7) and 3.5 × 10(5)CFU/cm(2)), which was confirmed by CLSM. Biofilms developed using consortium of Bacillus subtilis and Bacillus cereus on coarse sand had more surface area and was able to remove 98% of Cr(III), SEM-EDX proved 92.60% Cr(III) adsorption on biofilms supported by coarse sand. Utilization of Bacillus biofilms for effective bioremoval of Cr(III) from chrome tanning effluent could be a better option for tannery industry, especially during post chrome tanning operation. Copyright © 2011 Elsevier B.V. All rights reserved.

Treatment of low-strength soluble wastewater using an anaerobic baffled reactor (ABR).

PubMed

Gopala Krishna, G V T; Kumar, Pramod; Kumar, Pradeep

2009-01-01

Treatment of low-strength soluble wastewater (COD approximately 500 mg/L) was studied using an eight chambered anaerobic baffled reactor (ABR). At pseudo steady-state (PSS), the average total and soluble COD values (COD(T) and COD(S)) at 8h hydraulic retention time (HRT) were found to be around 50 and 40 mg/L, respectively, while at 10h HRT average COD(T) and COD(S) values were of the order of 47 and 37 mg/L, respectively. COD and BOD (3 day, 27 degrees C) removal averaged more than 90%. Effluent conformed to Indian standards laid down for BOD (less than 30 mg/L). Reactor effluent characteristics exhibited very low values of standard deviation indicating excellent reactor stability at PSS in terms of effluent characteristics. Based on mass balance calculations, more than 60% of raw wastewater COD was estimated to be recovered as CH(4) in the gas phase. Compartment-wise profiles indicated that most of the BOD and COD got reduced in the initial compartments only. Sudden drop in pH (7.8-6.7) and formation of volatile fatty acids (VFA) (53-85 mg/L) were observed in the first compartment due to acidogenesis and acetogenesis. The pH increased and VFA concentration decreased longitudinally down the reactor. Residence time distribution (RTD) studies revealed that the flow pattern in the ABR was neither completely plug-flow nor perfectly mixed. Observations from scanning electron micrographs (SEM) suggest that distinct phase separation takes place in an ABR.

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

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

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

1978-07-01

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

Shear compression testing of glass-fibre steel specimens after 4K reactor irradiation: Present status and facility upgrade

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

Gerstenberg, H.; Kraehling, E.; Katheder, H.

1997-06-01

The shear strengths of various fibre reinforced resins being promising candidate insulators for superconducting coils to be used tinder a strong radiation load, e.g. in future fusion reactors were investigated prior and subsequent to reactor in-core irradiation at liquid helium temperature. A large number of sandwich-like (steel-bonded insulation-steel) specimens representing a widespread variety of materials and preparation techniques was exposed to irradiation doses of up to 5 x 10{sup 7} Gy in form of fast neutrons and {gamma}-radiation. In a systematic study several experimental parameters including irradiation dose, postirradiation storage temperature and measuring temperature were varied before the determination ofmore » the ultimate shear strength. The results obtained from the different tested materials are compared. In addition an upgrade of the in-situ test rig installed at the Munich research reactor is presented, which allows combined shear/compression loading of low temperature irradiated specimens and provides a doubling of the testing rate.« less

Assessment of quasi-linear effect of RF power spectrum for enabling lower hybrid current drive in reactor plasmas

NASA Astrophysics Data System (ADS)

Cesario, Roberto; Cardinali, Alessandro; Castaldo, Carmine; Amicucci, Luca; Ceccuzzi, Silvio; Galli, Alessandro; Napoli, Francesco; Panaccione, Luigi; Santini, Franco; Schettini, Giuseppe; Tuccillo, Angelo Antonio

2017-10-01

The main research on the energy from thermonuclear fusion uses deuterium plasmas magnetically trapped in toroidal devices. To suppress the turbulent eddies that impair thermal insulation and pressure tight of the plasma, current drive (CD) is necessary, but tools envisaged so far are unable accomplishing this task while efficiently and flexibly matching the natural current profiles self-generated at large radii of the plasma column [1-5]. The lower hybrid current drive (LHCD) [6] can satisfy this important need of a reactor [1], but the LHCD system has been unexpectedly mothballed on JET. The problematic extrapolation of the LHCD tool at reactor graded high values of, respectively, density and temperatures of plasma has been now solved. The high density problem is solved by the FTU (Frascati Tokamak Upgrade) method [7], and solution of the high temperature one is presented here. Model results based on quasi-linear (QL) theory evidence the capability, w.r.t linear theory, of suitable operating parameters of reducing the wave damping in hot reactor plasmas. Namely, using higher RF power densities [8], or a narrower antenna power spectrum in refractive index [9,10], the obstacle for LHCD represented by too high temperature of reactor plasmas should be overcome. The former method cannot be used for routinely, safe antenna operations, Thus, only the latter key is really exploitable in a reactor. The proposed solutions are ultimately necessary for viability of an economic reactor.

Cadmium removal using Cladophora in batch, semi-batch and flow reactors.

PubMed

Sternberg, Steven P K; Dorn, Ryan W

2002-02-01

This study presents the results of using viable algae to remove cadmium from a synthetic wastewater. In batch and semi-batch tests, a local strain of Cladophora algae removed 80-94% of the cadmium introduced. The flow experiments that followed were conducted using non-local Cladophora parriaudii. Results showed that the alga removed only 12.7(+/-6.4)% of the cadmium introduced into the reactor. Limited removal was the result of insufficient algal quantities and poor contact between the algae and cadmium solution.

7. INTERIOR OF BUILDING 514. VIEW TO WEST. Rocky ...

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

7. INTERIOR OF BUILDING 514. VIEW TO WEST. - Rocky Mountain Arsenal, Lewisite Reactor & Distilled Mustard Distillation Building, 420 feet South of December Seventh Avenue; 1070 feet East of D Street, Commerce City, Adams County, CO

Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania`s 14-MW TRIGA reactor

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

Bretscher, M.M.; Snelgrove, J.L.; Ciocanescu, M.

1992-12-01

The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, eachmore » containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations.« less

The effect of microwave electromagnetic radiation on organic compounds removal efficiency in a reactor with a biofilm.

PubMed

Zielinski, M; Krzemieniewski, M

2007-01-01

This article shows the results of research on microwave radiation as a factor affecting organic compounds removal in a reactor with a biofilm. In the experiment a bioreactor was situated inside a microwave tube and there exposed to radiation. Municipal wastes were supplied to the bioreactor from a retention tank, to which they returned having passed through the reactor's packing. The whole system operated in a time cycle comprising a 24-hour detention of the wastewaters supply. The research was based on the specific properties of microwave heating, i.e. their ability to heat only the substances of appropriate dielectric properties. As the reactor was properly constructed and the microwave generator work was synchronised with that of the volumetric pump, microwave energy was directed mostly to the biofilm. It was observed that as a result of microwave radiation the process of organic compounds removal, defined as Chemical Oxygen Demand COD, increased its rate nearly by half. Simultaneously the process efficiency increased by 7.7% at the maximum. While analysing the changes the organic compounds underwent it was revealed that the load in-built in the biomass decreased by over half as a result of microwave radiation input at 2.5 W s(-1), which was optimal under the experimental conditions. Similarly the amount of pollutant remaining in the treated effluent decreased nearly by half, whereas the role of oxidation in removing organic pollutant increased in excess of 25% when compared to the control system.

Leachate flush strategies for managing volatile fatty acids accumulation in leach-bed reactors.

PubMed

Riggio, S; Torrijos, M; Vives, G; Esposito, G; van Hullebusch, E D; Steyer, J P; Escudié, R

2017-05-01

In anaerobic leach-bed reactors (LBRs) co-digesting an easily- and a slowly-degradable substrate, the importance of the leachate flush both on extracting volatile fatty acids (VFAs) at the beginning of newly-started batches and on their consumption in mature reactors was tested. Regarding VFA extraction three leachate flush-rate conditions were studied: 0.5, 1 and 2Lkg -1 TSd -1 . Results showed that increasing the leachate flush-rate during the acidification phase is essential to increase degradation kinetics. After this initial phase, leachate injection is less important and the flush-rate could be reduced. The injection in mature reactors of leachate with an acetic acid concentration of 5 or 10gL -1 showed that for an optimized VFA consumption in LBRs, VFAs should be provided straight after the methane production peak in order to profit from a higher methanogenic activity, and every 6-7h to maintain a high biogas production rate. Copyright © 2017 Elsevier Ltd. All rights reserved.

High-solid mesophilic methane fermentation of food waste with an emphasis on Iron, Cobalt, and Nickel requirements.

PubMed

Qiang, Hong; Lang, Dong-Li; Li, Yu-You

2012-01-01

The effect of trace metals on the mesophilic methane fermentation of high-solid food waste was investigated using both batch and continuous experiments. The continuous experiment was conducted by using a CSTR-type reactor with three run. During the first run, the HRT of the reactor was stepwise decreased from 100 days to 30 days. From operation day 50, the reactor efficiency deteriorated due to the lack of trace metals. The batch experiment showed that iron, cobalt, and nickel combinations had a significant effect on food waste. According to the results of the batch experiment, a combination of iron, cobalt, and nickel was added into the CSTR reactor by two different methods at run II, and III. Based on experimental results and theoretical calculations, the most suitable values of Fe/COD, Co/COD, and Ni/COD in the substrate were identified as 200, 6.0, and 5.7 mg/kg COD, respectively. Copyright © 2011 Elsevier Ltd. All rights reserved.

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

Preliminary neutronic analysis of a cavity test reactor

NASA Technical Reports Server (NTRS)

Whitmarsh, C. L., Jr.

1973-01-01

A reference configuration was calculated for a cavity test reactor to be used for testing the gascore nuclear rocket concept. A thermal flux of 4.1 x 10 to the 14th power neutrons per square centimeter per second in the cavity was provided by a driver fuel loading of 6.4 kg of enriched uranium in MTR fuel elements. The reactor was moderated and cooled by heavy water and reflected with 25.4 cm of beryllium. Power generation of 41.3 MW in the driver fuel is rejected to a heat sink. Design effort was directed toward minimization of driver power while maintaining 2.7 MW in the cavity during a test run. Ancillary data on material reactivity worths, reactivity coefficients, flux spectra, and power distributions are reported.

Determine Operating Reactor to Use for the 2016 PCI Level 1 Milestone

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

Clarno, Kevin T.

2016-01-30

The Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) Level 1 milestone to “Assess the analysis capability for core-wide [pressurized water reactor] PWR Pellet- Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region” (L1:CASL.P13.03) requires a particular type of nuclear power plant for the assessment. This report documents the operating reactor and cycles chosen for this assessment in completion of the physics integration (PHI) milestone to “Determine Operating Reactor to use for PCI L1 Milestone” (L3:PHI.CMD.P12.02). Watts Bar Unit 1 experienced (at least) one fuel rod failure in each of cycles 6 and 7, andmore » at least one was deemed to be duty related rather than being primarily related to a manufacturing defect or grid effects. This brief report documents that the data required to model cycles 1–12 of Watts Bar Unit 1 using VERA-CS contains sufficient data to model the PHI portion of the PCI challenge problem. A list of additional data needs is also provided that will be important for verification and validation of the BISON results.« less

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

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

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

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

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

DOE PAGES

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

2017-03-11

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

Characterization and anaerobic treatment of the sanitary landfill leachate in Istanbul.

PubMed

Inanc, B; Calli, B; Saatci, A

2000-01-01

In this study, characterization and anaerobic treatability of leachate from Komurcuoda Sanitary Landfill located on the Asian part of Istanbul were investigated. Time based fluctuations in characteristics of leachate were monitored for an 8 month period. Samples were taken from a 200 m3 holding tank located at the lowest elevation of the landfill. COD concentrations have ranged between 18,800 and 47,800 mg/l while BOD5 between 6820 and 38,500 mg/L. COD and BOD5 values were higher in summer and lower in winter due to dilution by precipitation. On the other hand, it was quite interesting that such a dilution effect was not observed for ammonia. The highest ammonia concentration, 2690 mg/L was in November 1998. BOD5/COD ratio was larger than 0.7 for most samples indicating high biodegradability, and acidic phase of decomposition in the landfill. For anaerobic treatability, three different reactors, namely an upflow anaerobic sludge bed reactor, an anaerobic upflow filter and a hybrid bed reactor, were used. The anaerobic reactors were operated for more than 230 days and were continuing operation when this paper was prepared. Organic loading was increased gradually from 1.3 kg COD/m3.day to 8.2 kg COD/m3.day while hydraulic retention time was reduced from 2.4 days to 2.0 days. All the reactors showed similar performances against organic loadings with efficiencies between 80% and 90%. However the reactors have experienced high ammonia concentrations several times throughout the experimental period, and showed different inhibition levels. Anaerobic filter was the least affected reactor while UASB was the most. Hybrid bed reactor has exhibited a similar performance to anaerobic filter although not to the same degree.

Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

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

Mandelli, Diego; Prescott, Steven R; Smith, Curtis L

2011-07-01

In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of amore » power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.« less

Development of downflow hanging sponge (DHS) reactor as post treatment of existing combined anaerobic tank treating natural rubber processing wastewater.

PubMed

Watari, Takahiro; Cuong Mai, Trung; Tanikawa, Daisuke; Hirakata, Yuga; Hatamoto, Masashi; Syutsubo, Kazuaki; Fukuda, Masao; Nguyen, Ngoc Bich; Yamaguchi, Takashi

2017-01-01

Conventional aerated tank technology is widely applied for post treatment of natural rubber processing wastewater in Southeast Asia; however, a long hydraulic retention time (HRT) is required and the effluent standards are exceeded. In this study, a downflow hanging sponge (DHS) reactor was installed as post treatment of anaerobic tank effluent in a natural rubber factory in South Vietnam and the process performance was evaluated. The DHS reactor demonstrated removal efficiencies of 64.2 ± 7.5% and 55.3 ± 19.2% for total chemical oxygen demand (COD) and total nitrogen, respectively, with an organic loading rate of 0.97 ± 0.03 kg-COD m -3 day -1 and a nitrogen loading rate of 0.57 ± 0.21 kg-N m -3 day -1 . 16S rRNA gene sequencing analysis of the sludge retained in the DHS also corresponded to the result of reactor performance, and both nitrifying and denitrifying bacteria were detected in the sponge carrier. In addition, anammox bacteria was found in the retained sludge. The DHS reactor reduced the HRT of 30 days to 4.8 h compared with the existing algal tank. This result indicates that the DHS reactor could be an appropriate post treatment for the existing anaerobic tank for natural rubber processing wastewater treatment.

Combustion of Na 2B 4O 7 + Mg + C to synthesis B 4C powders

NASA Astrophysics Data System (ADS)

Guojian, Jiang; Jiayue, Xu; Hanrui, Zhuang; Wenlan, Li

2009-09-01

Boron carbide powder was fabricated by combustion synthesis (CS) method directly from mixed powders of borax (Na 2B 4O 7), magnesium (Mg) and carbon. The adiabatic temperature of the combustion reaction of Na 2B 4O 7 + 6 Mg + C was calculated. The control of the reactions was achieved by selecting reactant composition, relative density of powder compact and gas pressure in CS reactor. The effects of these different influential factors on the composition and morphologies of combustion products were investigated. The results show that, it is advantageous for more Mg/Na 2B 4O 7 than stoichiometric ratio in Na 2B 4O 7 + Mg + C system and high atmosphere pressure in the CS reactor to increase the conversion degree of reactants to end product. The final product with the minimal impurities' content could be fabricated at appropriate relative density of powder compact. At last, boron carbide without impurities could be obtained after the acid enrichment and distilled water washing.

Post-Irradiation Non-Destructive Analyses of the AFIP-7 Experiment

NASA Astrophysics Data System (ADS)

Williams, W. J.; Robinson, A. B.; Rabin, B. H.

2017-12-01

This article reports the results and interpretation of post-irradiation non-destructive examinations performed on four curved full-size fuel plates that comprise the AFIP-7 experiment. These fuel plates, having a U-10 wt.%Mo monolithic design, were irradiated under moderate operating conditions in the Advanced Test Reactor to assess fuel performance for geometries that are prototypic of research reactor fuel assemblies. Non-destructive examinations include visual examination, neutron radiography, profilometry, and precision gamma scanning. This article evaluates the qualitative and quantitative data taken for each plate, compares corresponding data sets, and presents the results of swelling analyses. These characterization results demonstrate that the fuel meets established irradiation performance requirements for mechanical integrity, geometric stability, and stable and predictable behavior.

Neem leaves as a source of fertilizer-cum-pesticide vermicompost.

PubMed

Gajalakshmi, S; Abbasi, S A

2004-05-01

Vermicomposting of neem (Azadirachta indica A. Juss) was accomplished in "high-rate" reactors operated at the earthworm (Eudrilus eugeniae) densities of 62.5 and 75 animals per litre of reactor volume. Contrary to the fears that neem--a powerful nematicide--might not be palatable to the annelids, the earthworms fed voraciously on the neem compost, converting upto 7% of the feed into vermicompost per day. Indeed the worms grew faster and reproduced more rapidly in the neem-fed vermireactors than in the reactors fed with mango leaf litter earlier studied by the authors (Gajalakshmi et al., 2003). Another set of experiments on the growth, flowering, and fruition of brinjal (Solanum melongena) plants with and without fertilization with vermicompost, revealed that the vermicompost had a significantly beneficial impact.

A modified biotrickling filter for nitrification-denitrification in the treatment of an ammonia-contaminated air stream.

PubMed

Raboni, Massimo; Torretta, Vincenzo

2016-12-01

A conventional biotrickling filter for airborne ammonia nitrification has been modified, by converting the liquid sump into a biological denitrifying reactor. The biotrickling filter achieves an average ammonia removal efficiency of 92.4 %, with an empty bed retention time (EBRT) equal to 36 s and an average ammonia concentration of 54.7 mg Nm -3 in the raw air stream. The denitrification reactor converts ammonia into inert gas N 2 , in addition to other important advantages connected to the alkaline character of the biochemical pathway of the denitrifying bacteria. Firstly, the trickling water crossing the denitrification reactor underwent a notable pH increase from 7.3 to 8.0 which prevented the acidic inhibition of the nitrifying bacteria due to the buildup of nitric and nitrous acids. Secondly, the pH increase created the ideal conditions for the autotrophic nitrifying bacteria. The tests proved that an ammonia removal efficiency of above 90 % can be achieved with an EBRT greater than 30 s and a volumetric load lower than 200 g NH 3  m -3  day -1 . The results of the biofilm observation by using a scanning confocal laser microscope are reported together with the identification of degrading bacteria genera in the biotrickling filter. The efficiency of the plant and its excellent operational stability highlight the effectiveness of the synergistic action between the denitrification reactor and the biotrickling filter in removing airborne ammonia.

PROCESS WATER BUILDING, TRA605. SECTIONS B, C AND D SHOW ...

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

PROCESS WATER BUILDING, TRA-605. SECTIONS B, C AND D SHOW RELATIONSHIP BETWEEN FLASH EVAPORATORS (ABOVE) AND SEAL AND SUMP TANKS (BELOW). BASEMENT FLOOR IS BELOW GRADE; FIRST FLOOR, ABOVE GRADE. SHIELDING TOLERANCES. BLAW-KNOX 3150-5-7, 8/1950. INL INDEX NO. 531-605-00-098-100012, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

The influence of pH adjustment on kinetics parameters in tapioca wastewater treatment using aerobic sequencing batch reactor system

NASA Astrophysics Data System (ADS)

Mulyani, Happy; Budianto, Gregorius Prima Indra; Margono, Kaavessina, Mujtahid

2018-02-01

The present investigation deals with the aerobic sequencing batch reactor system of tapioca wastewater treatment with varying pH influent conditions. This project was carried out to evaluate the effect of pH on kinetics parameters of system. It was done by operating aerobic sequencing batch reactor system during 8 hours in many tapioca wastewater conditions (pH 4.91, pH 7, pH 8). The Chemical Oxygen Demand (COD) and Mixed Liquor Volatile Suspended Solids (MLVSS) of the aerobic sequencing batch reactor system effluent at steady state condition were determined at interval time of two hours to generate data for substrate inhibition kinetics parameters. Values of the kinetics constants were determined using Monod and Andrews models. There was no inhibition constant (Ki) detected in all process variation of aerobic sequencing batch reactor system for tapioca wastewater treatment in this study. Furthermore, pH 8 was selected as the preferred aerobic sequencing batch reactor system condition in those ranging pH investigated due to its achievement of values of kinetics parameters such µmax = 0.010457/hour and Ks = 255.0664 mg/L COD.

A REACTOR DESIGN PARAMETER STUDY

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

Fox, A.H.; LaVerne, M.E.; Burtnette, C.S.

1954-06-25

Multigroup calculations were performed on reflectormoderated systems to establish some of the nuclear characteristics of various reflector geometries and materials. C, Li/sup 7/, Li/sup 7/OD, and NaOD moderators were used with NaF-UF/ sub 4/ fuel. The results are tabulated for 57 moderator and dimensional variations. (D.E.B.)

Phenol removal from hypersaline wastewaters in a Membrane Biological Reactor (MBR): operation and microbiological characterisation.

PubMed

Dosta, J; Nieto, J M; Vila, J; Grifoll, M; Mata-Álvarez, J

2011-03-01

In this study, two Membrane Biological Reactors (MBR) with submerged flat membranes, one at lab-scale conditions and the other at pilot-plant conditions, were operated at environmental temperature to treat an industrial wastewater characterised by low phenol concentrations (8-16 mg L(-1)) and high salinity (∼ 150-160 mS cm(-1)). During the operation of both reactors, the phenol loading rate was progressively increased and less than 1mg phenol L(-1) was detected even at very low HRTs (0.5-0.7 days). Membrane fouling was minimized by the cross flow aeration rate inside the MBRs and by intermittent permeation. Microbial community analysis of both reactors revealed that members of the genera Halomonas and Marinobacter (gammaproteobacteria) were major components. Growth-linked phenol degradation by pure cultures of Marinobacter isolates demonstrated that this bacterium played a major role in the removal of phenol from the bioreactors. Copyright © 2010 Elsevier Ltd. All rights reserved.

Modeling and Design Optimization of Multifunctional Membrane Reactors for Direct Methane Aromatization

PubMed Central

Fouty, Nicholas J.; Carrasco, Juan C.; Lima, Fernando V.

2017-01-01

Due to the recent increase of natural gas production in the U.S., utilizing natural gas for higher-value chemicals has become imperative. Direct methane aromatization (DMA) is a promising process used to convert methane to benzene, but it is limited by low conversion of methane and rapid catalyst deactivation by coking. Past work has shown that membrane separation of the hydrogen produced in the DMA reactions can dramatically increase the methane conversion by shifting the equilibrium toward the products, but it also increases coke production. Oxygen introduction into the system has been shown to inhibit this coke production while not inhibiting the benzene production. This paper introduces a novel mathematical model and design to employ both methods in a multifunctional membrane reactor to push the DMA process into further viability. Multifunctional membrane reactors, in this case, are reactors where two different separations occur using two differently selective membranes, on which no systems studies have been found. The proposed multifunctional membrane design incorporates a hydrogen-selective membrane on the outer wall of the reaction zone, and an inner tube filled with airflow surrounded by an oxygen-selective membrane in the middle of the reactor. The design is shown to increase conversion via hydrogen removal by around 100%, and decrease coke production via oxygen addition by 10% when compared to a tubular reactor without any membranes. Optimization studies are performed to determine the best reactor design based on methane conversion, along with coke and benzene production. The obtained optimal design considers a small reactor (length = 25 cm, diameter of reaction tube = 0.7 cm) to subvert coke production and consumption of the product benzene as well as a high permeance (0.01 mol/s·m2·atm1/4) through the hydrogen-permeable membrane. This modeling and design approach sets the stage for guiding further development of multifunctional membrane reactor models and designs for natural gas utilization and other chemical reaction systems. PMID:28850068

MTRETR MAINTENANCE SHOP, TRA653. FLOOR PLAN FOR FIRST FLOOR: MACHINE ...

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

MTR-ETR MAINTENANCE SHOP, TRA-653. FLOOR PLAN FOR FIRST FLOOR: MACHINE SHOP, ELECTRICAL AND INSTRUMENT SHOP, TOOL CRIB, ELECTRONIC SHOP, LOCKER ROOM, SPECIAL TEMPERATURE CONTROLLED ROOM, AND OFFICES. "NEW" ON DRAWING REFERS TO REVISION OF 11/1956 DRAWING ON WHICH AREAS WERE DESIGNATED AS "FUTURE." HUMMEL HUMMEL & JONES 810-MTR-ETR-653-A-7, 5/1957. INL INDEX NO. 532-0653-00-381-101839, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Stable partial nitritation for low-strength wastewater at low temperature in an aerobic granular reactor.

PubMed

Isanta, Eduardo; Reino, Clara; Carrera, Julián; Pérez, Julio

2015-09-01

Partial nitritation for a low-strength wastewater at low temperature was stably achieved in an aerobic granular reactor. A bench-scale granular sludge bioreactor was operated in continuous mode treating an influent of 70 mg N-NH4(+) L(-1) to mimic pretreated municipal nitrogenous wastewater and the temperature was progressively decreased from 30 to 12.5 °C. A suitable effluent nitrite to ammonium concentrations ratio to a subsequent anammox reactor was maintained stable during 300 days at 12.5 °C. The average applied nitrogen loading rate at 12.5 °C was 0.7 ± 0.3 g N L(-1) d(-1), with an effluent nitrate concentration of only 2.5 ± 0.7 mg N-NO3(-) L(-1). The biomass fraction of nitrite-oxidizing bacteria (NOB) in the granular sludge decreased from 19% to only 1% in 6 months of reactor operation at 12.5 °C. Nitrobacter spp. where found as the dominant NOB population, whereas Nitrospira spp. were not detected. Simulations indicated that: (i) NOB would only be effectively repressed when their oxygen half-saturation coefficient was higher than that of ammonia-oxidizing bacteria; and (ii) a lower specific growth rate of NOB was maintained at any point in the biofilm (even at 12.5 °C) due to the bulk ammonium concentration imposed through the control strategy. Copyright © 2015 Elsevier Ltd. All rights reserved.

Single-phase and two-phase anaerobic digestion of fruit and vegetable waste: Comparison of start-up, reactor stability and process performance

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

Ganesh, Rangaraj; Torrijos, Michel, E-mail: michel.torrijos@supagro.inra.fr; Sousbie, Philippe

Highlights: • Single-phase and two-phase systems were compared for fruit and vegetable waste digestion. • Single-phase digestion produced a methane yield of 0.45 m{sup 3} CH{sub 4}/kg VS and 83% VS removal. • Substrate solubilization was high in acidification conditions at 7.0 kg VS/m{sup 3} d and pH 5.5–6.2. • Energy yield was lower by 33% for two-phase system compared to the single-phase system. • Simple and straight-forward operation favored single phase process over two-phase process. - Abstract: Single-phase and two-phase digestion of fruit and vegetable waste were studied to compare reactor start-up, reactor stability and performance (methane yield, volatilemore » solids reduction and energy yield). The single-phase reactor (SPR) was a conventional reactor operated at a low loading rate (maximum of 3.5 kg VS/m{sup 3} d), while the two-phase system consisted of an acidification reactor (TPAR) and a methanogenic reactor (TPMR). The TPAR was inoculated with methanogenic sludge similar to the SPR, but was operated with step-wise increase in the loading rate and with total recirculation of reactor solids to convert it into acidification sludge. Before each feeding, part of the sludge from TPAR was centrifuged, the centrifuge liquid (solubilized products) was fed to the TPMR and centrifuged solids were recycled back to the reactor. Single-phase digestion produced a methane yield of 0.45 m{sup 3} CH{sub 4}/kg VS fed and VS removal of 83%. The TPAR shifted to acidification mode at an OLR of 10.0 kg VS/m{sup 3} d and then achieved stable performance at 7.0 kg VS/m{sup 3} d and pH 5.5–6.2, with very high substrate solubilization rate and a methane yield of 0.30 m{sup 3} CH{sub 4}/kg COD fed. The two-phase process was capable of high VS reduction, but material and energy balance showed that the single-phase process was superior in terms of volumetric methane production and energy yield by 33%. The lower energy yield of the two-phase system was due to the loss of energy during hydrolysis in the TPAR and the deficit in methane production in the TPMR attributed to COD loss due to biomass synthesis and adsorption of hard COD onto the flocs. These results including the complicated operational procedure of the two-phase process and the economic factors suggested that the single-phase process could be the preferred system for FVW.« less

Needs of Accurate Prompt and Delayed γ-spectrum and Multiplicity for Nuclear Reactor Designs

NASA Astrophysics Data System (ADS)

Rimpault, G.; Bernard, D.; Blanchet, D.; Vaglio-Gaudard, C.; Ravaux, S.; Santamarina, A.

The local energy photon deposit must be accounted accurately for Gen-IV fast reactors, advanced light-water nuclear reactors (Gen-III+) and the new experimental Jules Horowitz Reactor (JHR). The γ energy accounts for about 10% of the total energy released in the core of a thermal or fast reactor. The γ-energy release is much greater in the core of the reactor than in its structural sub-assemblies (such as reflector, control rod followers, dummy sub-assemblies). However, because of the propagation of γ from the core regions to the neighboring fuel-free assemblies, the contribution of γ energy to the total heating can be dominant. For reasons related to their performance, power reactors require a 7.5% (1σ) uncertainty for the energy deposition in non-fuelled zones. For the JHR material-testing reactor, a 5% (1 s) uncertainty is required in experimental positions. In order to verify the adequacy of the calculation of γ-heating, TLD and γ-fission chambers were used to derive the experimental heating values. Experimental programs were and are still conducted in different Cadarache facilities such as MASURCA (for SFR), MINERVE and EOLE (for JHR and Gen-III+ reactors). The comparison of calculated and measured γ-heating values shows an underestimation in all experimental programs indicating that for the most γ-production data from 239Pu in current nuclear-data libraries is highly suspicious.The first evaluation priority is for prompt γ-multiplicity for U and Pu fission but similar values for otheractinides such as Pu and U are also required. The nuclear data library JEFF3.1.1 contains most of the photon production data. However, there are some nuclei for which there are missing or erroneous data which need to be completed or modified. A review of the data available shows a lack of measurements for conducting serious evaluation efforts. New measurements are needed to guide new evaluation efforts which benefit from consolidated modeling techniques.

Characterization of elemental release during microbe-basalt interactions

NASA Astrophysics Data System (ADS)

Wu, L.; Jacobson, A. D.; Hausner, M.

2006-12-01

This study used batch reactors to characterize the rates, mechanisms, and stoichiometry of elemental release during the interaction of Burkholderia fungorum, a common soil microbe, with Columbia River Flood Basalt at 28°C for 36 d. We especially focused on the release of Ca, Mg, P, Si, and Sr under a variety of biotic and abiotic conditions with the ultimate aim of evaluating how actively metabolizing bacteria might influence basalt weathering on the continents. Four days after inoculating P-limited reactors (those lacking P in the growth medium), pH decreased from ~7 to 4, and glucose was depleted. Theoretical calculations suggest that the lowered pH resulted from the release of organic acids and/or CO2. Purely abiotic control reactors as well as control reactors containing nonviable cells showed constant glucose concentrations and near-neutral pH. Over the entire 36 day period, the P-limited reactors yielded Ca, Mg, Si, and Sr release rates several times higher than those observed in the P-bearing biotic reactors and the abiotic controls. Release rates directly correlate with pH, indicating that proton-promoted dissolution was the dominant reaction mechanism. Ligand- promoted dissolution was probably less important because the P-limited and P-bearing reactors experienced nearly identical rates of microbial growth, but the P-bearing reactors displayed overall lower dissolution rates at near-neutral pH, where presumably, the effect of ligand-promoted dissolution would be most evident. Chemical analyses of bacteria collected at the end of the experiments, combined with mass-balances between the biological and fluid phases, demonstrate that the low P concentration in the biotic reactors was an artifact of P uptake during microbial growth. These findings suggest that when bacteria utilize basalt as a nutrient source, they can potentially elevate the rate of long-term atmospheric CO2 consumption by Ca-Mg silicate weathering by a factor of 5 over the corresponding inorganic rate.

Characterization of elemental release during microbe basalt interactions at T = 28 °C

NASA Astrophysics Data System (ADS)

Wu, Lingling; Jacobson, Andrew D.; Chen, Hsin-Chieh; Hausner, Martina

2007-05-01

This study used batch reactors to characterize the rates and mechanisms of elemental release during the interaction of a single bacterial species ( Burkholderia fungorum) with Columbia River Flood Basalt at T = 28 °C for 36 days. We primarily examined the release of Ca, Mg, P, Si, and Sr under a variety of biotic and abiotic conditions with the aim of evaluating how actively metabolizing bacteria might influence basalt weathering on the continents. Four days after inoculating P-limited reactors (those lacking P in the growth medium), the concentration of viable planktonic cells increased from ˜10 4 to 10 8 CFU (Colony Forming Units)/mL, pH decreased from ˜7 to 4, and glucose decreased from ˜1200 to 0 μmol/L. Mass-balance and acid-base equilibria calculations suggest that the lowered pH resulted from either respired CO 2, organic acids released during biomass synthesis, or H + extrusion during NH4+ uptake. Between days 4 and 36, cell numbers remained constant at ˜10 8 CFU/mL and pH increased to ˜5. Purely abiotic control reactors as well as control reactors containing inert cells (˜10 8 CFU/mL) showed constant glucose concentrations, thus confirming the absence of biological activity in these experiments. The pH of all control reactors remained near-neutral, except for one experiment where the pH was initially adjusted to 4 but rapidly rose to 7 within 2 days. Over the entire 36 day period, P-limited reactors containing viable bacteria yielded the highest Ca, Mg, Si, and Sr release rates. Release rates inversely correlate with pH, indicating that proton-promoted dissolution was the dominant reaction mechanism. Both biotic and abiotic P-limited reactors displayed low P concentrations. Chemical analyses of bacteria collected at the end of the experiments, combined with mass-balances between the biological and fluid phases, demonstrate that the absence of dissolved P in the biotic reactors resulted from microbial P uptake. The only P source in the basalt is a small amount of apatite (˜1.2%), which occurs as needles within feldspar grains and glass. We therefore conclude that B. fungorum utilized apatite as a P source for biomass synthesis, which stimulated elemental release from coexisting mineral phases via pH lowering. The results of this study suggest that actively metabolizing bacteria have the potential to influence elemental release from basalt in continental settings.

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

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

Dunn, Floyd E.; Hu, Lin-wen; Wilson, Erik

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings onmore » avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.« less

Biogeochemical controls on interactions of microbial iron and sulfate reduction

NASA Astrophysics Data System (ADS)

Kirk, M. F.; Paper, J. M.; Haller, B. R.; Shodunke, G. O.; Marquart, K. A.; Jin, Q.

2016-12-01

Although iron and sulfate reduction are two of the most common microbial electron accepting processes in anoxic settings, the relative influences of environmental factors that guide interactions between each are poorly known. Identifying these factors is a key to predicting how those interactions will respond to future environmental changes. In this study, we used semi-continuous bioreactors to examine the influence of pH, electron donor flux, and sulfate availability. The reactors contained 100 mL of aqueous media and 1 g of marsh sediment amended with goethite (1 mmol). One set of reactors received acidic media (pH 6) while the other set received alkaline media (pH 7.5). Media for both sets of reactors included acetate (0.25 and 1 mM), which served as an electron donor, and sulfate (2.5 mM). We also included sets of sulfate-deficient and acetate-deficient control reactors. We maintained a fluid residence time of 35 days in the reactors by sampling and feeding them every seven days during the 91-day incubation. Our results show that, under the conditions tested, pH had a larger influence on the balance between each reaction than acetate concentration. In acidic reactors, the molar amount of iron reduced exceeded the amount of sulfate reduced by a factor of 3 in reactors receiving media with 0 and 0.25 mM acetate and a factor of 2 in reactors receiving 1 mM acetate. Under alkaline conditions, iron and sulfate were reduced in nearly equal proportions, regardless of influent acetate concentration. Results from sulfate-deficient control reactors show that the presence of sulfate reduction increased the extent of iron reduction in all reactors, but particularly those with alkaline pH. Under acidic conditions, the amount of iron reduced was greater by a factor of 1.2 if sulfate reduction occurred simultaneously than if it did not. Under alkaline conditions, that factor increased to 8.2. Hence, pH influenced the extent to which sulfate reduction promoted iron reduction.

Science And Technology In Rapid Crisis Response. Volume 7, September 2011

DTIC Science & Technology

2011-09-01

Nuclear Reactor Accident at Fukushima Daiichi .... 16 From the Eyes of the Marines: III MEF’s Assistance in Japan Relief Efforts ............ 20 Chilean...by the radiological hazards at the Fukushima Daiichi Nuclear Power Plant. Although this project is in its early stages, the team hopes to have...OBSERVATIONS ON THE U.S. GOVERNMENT RESPONSE TO THE “GREAT EAST JAPAN EARTHQUAKE” AND THE NUCLEAR REACTOR ACCIDENT AT FUKUSHIMA DAIICHI CAPT Jim

Long-term stability of thermophilic co-digestion submerged anaerobic membrane reactor encountering high organic loading rate, persistent propionate and detectable hydrogen in biogas.

PubMed

Qiao, Wei; Takayanagi, Kazuyuki; Niu, Qigui; Shofie, Mohammad; Li, Yu You

2013-12-01

The performance of thermophilic anaerobic co-digestion of coffee grounds and sludge using membrane reactor was investigated for 148 days, out of a total research duration of 263 days. The OLR was increased from 2.2 to 33.7 kg-COD/m(3)d and HRT was shortened from 70 to 7 days. A significant irreversible drop in pH confirmed the overload of reactor. Under a moderately high OLR of 23.6 kg-COD/m(3)d, and with HRT and influent total solids of 10 days and 150 g/L, respectively, the COD removal efficiency was 44.5%. Hydrogen in biogas was around 100-200 ppm, which resulted in the persistent propionate of 1.0-3.2g/L. The VFA consumed approximately 60% of the total alkalinity. NH4HCO3 was supplemented to maintain alkalinity. The stability of system relied on pH management under steady state. The 16SrDNA results showed that hydrogen-utilizing methanogens dominates the archaeal community. The propionate-oxidizing bacteria in bacterial community was insufficient. Copyright © 2013 Elsevier Ltd. All rights reserved.

HOT CELL BUILDING, TRA632. CONTEXTUAL AERIAL VIEW OF HOT CELL ...

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

HOT CELL BUILDING, TRA-632. CONTEXTUAL AERIAL VIEW OF HOT CELL BUILDING, IN VIEW AT LEFT, AS YET WITHOUT ROOF. PLUG STORAGE BUILDING LIES BETWEEN IT AND THE SOUTH SIDE OF THE MTR BUILDING AND ITS WING. NOTE CONCRETE DRIVE BETWEEN ROLL-UP DOOR IN MTR BUILDING AND CHARGING FACE OF PLUG STORAGE. REACTOR SERVICES BUILDING (TRA-635) WILL COVER THIS DRIVE AND BUTT UP TO CHARGING FACE. DOTTED LINE IS ON ORIGINAL NEGATIVE. TRA PARKING LOT IN LEFT CORNER OF THE VIEW. CAMERA FACING NORTHWESTERLY. INL NEGATIVE NO. 8274. Unknown Photographer, 7/2/1953 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

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

David Andrs; Ray Berry; Derek Gaston

The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7more » is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to evolve with time. RELAP-7 is a MOOSE-based application. MOOSE (Multiphysics Object-Oriented Simulation Environment) is a framework for solving computational engineering problems in a well-planned, managed, and coordinated way. By leveraging millions of lines of open source software packages, such as PETSC (a nonlinear solver developed at Argonne National Laboratory) and LibMesh (a Finite Element Analysis package developed at University of Texas), MOOSE significantly reduces the expense and time required to develop new applications. Numerical integration methods and mesh management for parallel computation are provided by MOOSE. Therefore RELAP-7 code developers only need to focus on physics and user experiences. By using the MOOSE development environment, RELAP-7 code is developed by following the same modern software design paradigms used for other MOOSE development efforts. There are currently over 20 different MOOSE based applications ranging from 3-D transient neutron transport, detailed 3-D transient fuel performance analysis, to long-term material aging. Multi-physics and multiple dimensional analyses capabilities can be obtained by coupling RELAP-7 and other MOOSE based applications and by leveraging with capabilities developed by other DOE programs. This allows restricting the focus of RELAP-7 to systems analysis-type simulations and gives priority to retain and significantly extend RELAP5's capabilities.« less

Biodegradation of methyl t-butyl ether by aerobic granules under a cosubstrate condition.

PubMed

Zhang, L L; Chen, J M; Fang, F

2008-03-01

Aerobic granules efficient at degrading methyl tert-butyl ether (MTBE) with ethanol as a cosubstrate were successfully developed in a well-mixed sequencing batch reactor (SBR). Aerobic granules were first observed about 100 days after reactor startup. Treatment efficiency of MTBE in the reactor during stable operation exceeded 99.9%, and effluent MTBE was in the range of 15-50 microg/L. The specific MTBE degradation rate was observed to increase with increasing MTBE initial concentration from 25 to 500 mg/L, which peaked at 22.7 mg MTBE/g (volatile suspended solids).h and declined with further increases in MTBE concentration as substrate inhibition effects became significant. Microbial-community deoxyribonucleic acid profiling was carried out using denaturing gradient gel electrophoresis of polymerase chain reaction-amplified 16S ribosomal ribonucleic acid. The reactor was found to be inhabited by several diverse bacterial species, most notably microorganisms related to the genera Sphingomonas, Methylobacterium, and Hyphomicrobium vulgare. These organisms were previously reported to be associated with MTBE biodegradation. A majority of the bands in the reactor represented a group of organisms belonging to the Flavobacteria-Proteobacteria-Actinobacteridae class of bacteria. This study demonstrates that MTBE can be effectively degraded by aerobic granules under a cosubstrate condition and gives insight into the microorganisms potentially involved in the process.

Biomethanation of poultry litter leachate in UASB reactor coupled with ammonia stripper for enhancement of overall performance.

PubMed

Gangagni Rao, A; Sasi Kanth Reddy, T; Surya Prakash, S; Vanajakshi, J; Joseph, Johny; Jetty, Annapurna; Rajashekhara Reddy, A; Sarma, P N

2008-12-01

In the present study possibility of coupling stripper to remove ammonia to the UASB reactor treating poultry litter leachate was studied to enhance the overall performance of the reactor. UASB reactor with stripper as ammonia inhibition control mechanism exhibited better performance in terms of COD reduction (96%), methane yield (0.26m(3)CH(4)/kg COD reduced), organic loading rate (OLR) (18.5kg COD m(-3)day(-1)) and Hydraulic residence time (HRT) (12h) compared to the UASB reactor without stripper (COD reduction: 92%; methane yield: 0.21m(3)CH(4)/kg COD reduced; OLR: 13.6kg CODm(-3)day(-1); HRT: 16h). The improved performance was due to the reduction of total ammonia nitrogen (TAN) and free ammonia nitrogen (FAN) in the range of 75-95% and 80-95%, respectively by the use of stripper. G/L (air flow rate/poultry leachate flow rate) in the range of 60-70 and HRT in the range of 7-9min are found to be optimum parameters for the operation of the stripper.

Exploratory screening tests of several alloys and coatings for automobile thermal reactors

NASA Technical Reports Server (NTRS)

Oldrieve, R. E.

1971-01-01

A total of 23 materials (including uncoated ferritic and austenitic iron-base alloys, uncoated nickel and cobalt-base superalloys, and several different coatings on AISI 304 stainless steel) were screened as test coupons on a rack in an automobile thermal reactor. Test exposures were generally 51 hours including 142 thermal cycles of 10 minutes at 1010 + or - 30 C test coupon temperature and 7-minutes cool-down to about 510 C. Materials that exhibited corrosion resistance better than that of Hastelloy X include: a ferritic iron alloy with 6 weight percent aluminum; three nickel-base superalloys; two diffused-aluminum coatings on AISI 304; and a Ni-Cr slurry-sprayed coating on AISI 304. Preliminary comparison is made on the performance of the directly impinged coupons and a reactor core of the same material.

Summary of Apollo; A D- sup 3 He tokamak reactor design

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

Kulcinski, G.L.; Blanchard, T.P.; El-Guebaly, L.A.

1992-07-01

In this paper, the key features of Apollo, a conceptual D-{sup 3}He tokamak reactor for commercial electricity production, are summarized. The 1000-MW (electric) design utilizes direct conversion of transport, neutron, and bremsstrahlung radiation power. The direct conversion method uses reactants, and the thermal conversion cycle uses an organic coolant. Apollo operates in the first-stability regime, with a major radius of 7.89 m, a peak magnetic field on the toroidal field coils of 19.3 T, a 53-MA plasma current, and a 6.7% beta value. The low neutron production of the D-{sup 3}He fuel cycle greatly reduces the radiation damage rate andmore » allows a full-lifetime first wall and structure made of standard steels with only slight modifications to reduce activation levels.« less

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.

Temporal Microbial Community Dynamics in Microbial Electrolysis Cells – Influence of Acetate and Propionate Concentration

PubMed Central

Hari, Ananda Rao; Venkidusamy, Krishnaveni; Katuri, Krishna P.; Bagchi, Samik; Saikaly, Pascal E.

2017-01-01

Microbial electrolysis cells (MECs) are widely considered as a next generation wastewater treatment system. However, fundamental insight on the temporal dynamics of microbial communities associated with MEC performance under different organic types with varied loading concentrations is still unknown, nevertheless this knowledge is essential for optimizing this technology for real-scale applications. Here, the temporal dynamics of anodic microbial communities associated with MEC performance was examined at low (0.5 g COD/L) and high (4 g COD/L) concentrations of acetate or propionate, which are important intermediates of fermentation of municipal wastewaters and sludge. The results showed that acetate-fed reactors exhibited higher performance in terms of maximum current density (I: 4.25 ± 0.23 A/m2), coulombic efficiency (CE: 95 ± 8%), and substrate degradation rate (98.8 ± 1.2%) than propionate-fed reactors (I: 2.7 ± 0.28 A/m2; CE: 68 ± 9.5%; substrate degradation rate: 84 ± 13%) irrespective of the concentrations tested. Despite of the repeated sampling of the anodic biofilm over time, the high-concentration reactors demonstrated lower and stable performance in terms of current density (I: 1.1 ± 0.14 to 4.2 ± 0.21 A/m2), coulombic efficiency (CE: 44 ± 4.1 to 103 ± 7.2%) and substrate degradation rate (64.9 ± 6.3 to 99.7 ± 0.5%), while the low-concentration reactors produced higher and dynamic performance (I: 1.1 ± 0.12 to 4.6 ± 0.1 A/m2; CE: 52 ± 2.5 to 105 ± 2.7%; substrate degradation rate: 87.2 ± 0.2 to 99.9 ± 0.06%) with the different substrates tested. Correlating reactor’s performance with temporal dynamics of microbial communities showed that relatively similar anodic microbial community composition but with varying relative abundances was observed in all the reactors despite differences in the substrate and concentrations tested. Particularly, Geobacter was the predominant bacteria on the anode biofilm of all MECs over time suggesting its possible role in maintaining functional stability of MECs fed with low and high concentrations of acetate and propionate. Taken together, these results provide new insights on the microbial community dynamics and its correlation to performance in MECs fed with different concentrations of acetate and propionate, which are important volatile fatty acids in wastewater. PMID:28775719

Measurements of effective delayed neutron fraction in a fast neutron reactor using the perturbation method

NASA Astrophysics Data System (ADS)

Zhou, Hao-Jun; Yin, Yan-Peng; Fan, Xiao-Qiang; Li, Zheng-Hong; Pu, Yi-Kang

2016-06-01

A perturbation method is proposed to obtain the effective delayed neutron fraction β eff of a cylindrical highly enriched uranium reactor. Based on reactivity measurements with and without a sample at a specified position using the positive period technique, the reactor reactivity perturbation Δρ of the sample in β eff units is measured. Simulations of the perturbation experiments are performed using the MCNP program. The PERT card is used to provide the difference dk of effective neutron multiplication factors with and without the sample inside the reactor. Based on the relationship between the effective multiplication factor and the reactivity, the equation β eff = dk/Δρ is derived. In this paper, the reactivity perturbations of 13 metal samples at the designable position of the reactor are measured and calculated. The average β eff value of the reactor is given as 0.00645, and the standard uncertainty is 3.0%. Additionally, the perturbation experiments for β eff can be used to evaluate the reliabilities of the delayed neutron parameters. This work shows that the delayed neutron data of 235U and 238U from G.R. Keepin’s publication are more reliable than those from ENDF-B6.0, ENDF-B7.0, JENDL3.3 and CENDL2.2. Supported by Foundation of Key Laboratory of Neutron Physics, China Academy of Engineering Physics (2012AA01, 2014AA01), National Natural Science Foundation (11375158, 91326104)

Thermal analysis of cylindrical natural-gas steam reformer for 5 kW PEMFC

NASA Astrophysics Data System (ADS)

Jo, Taehyun; Han, Junhee; Koo, Bonchan; Lee, Dohyung

2016-11-01

The thermal characteristics of a natural-gas based cylindrical steam reformer coupled with a combustor are investigated for the use with a 5 kW polymer electrolyte membrane fuel cell. A reactor unit equipped with nickel-based catalysts was designed to activate the steam reforming reaction without the inclusion of high-temperature shift and low-temperature shift processes. Reactor temperature distribution and its overall thermal efficiency depend on various inlet conditions such as the equivalence ratio, the steam to carbon ratio (SCR), and the fuel distribution ratio (FDR) into the reactor and the combustor components. These experiments attempted to analyze the reformer's thermal and chemical properties through quantitative evaluation of product composition and heat exchange between the combustor and the reactor. FDR is critical factor in determining the overall performance as unbalanced fuel injection into the reactor and the combustor deteriorates overall thermal efficiency. Local temperature distribution also influences greatly on the fuel conversion rate and thermal efficiency. For the experiments, the operation conditions were set as SCR was in range of 2.5-4.0 and FDR was in 0.4-0.7 along with equivalence ratio of 0.9-1.1; optimum results were observed for FDR of 0.63 and SCR of 3.0 in the cylindrical steam reformer.

Contributions of each isotope in some fluids on neutronic performance in a fusion-fission hybrid reactor: a Monte Carlo method

NASA Astrophysics Data System (ADS)

Günay, M.; Şarer, B.; Kasap, H.

2014-08-01

In the present investigation, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa ferritic steel structural material and 99-95 % Li20Sn80-1-5 % SFG-Pu, 99-95 % Li20Sn80-1-5 % SFG-PuF4, 99-95 % Li20Sn80-1-5 % SFG-PuO2 the molten salt-heavy metal mixtures, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium zone with the width of 3 cm was used for the neutron multiplicity between liquid first wall and blanket. The contributions of each isotope in fluids on the nuclear parameters of a fusion-fission hybrid reactor such as tritium breeding ratio, energy multiplication factor, heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional analyses were performed by using Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

Selective synthesis of human milk fat-style structured triglycerides from microalgal oil in a microfluidic reactor packed with immobilized lipase

DOE PAGES

Wang, Jun; Liu, Xi; Wang, Xu -Dong; ...

2016-08-18

Human milk fat-style structured triacylglycerols were produced from microalgal oil in a continuous microfluidic reactor packed with immobilized lipase for the first time. A remarkably high conversion efficiency was demonstrated in the microreactor with reaction time being reduced by 8 times, Michaelis constant decreased 10 times, the lipase reuse times increased 2.25-fold compared to those in a batch reactor. In addition, the content of palmitic acid at sn-2 position (89.0%) and polyunsaturated fatty acids at sn-1, 3 positions (81.3%) are slightly improved compared to the product in a batch reactor. The increase of melting points (1.7 °C) and decrease ofmore » crystallizing point (3 °C) implied higher quality product was produced using the microfluidic technology. The main cost can be reduced from 212.3 to 14.6 per batch with the microreactor. Altogether, the microfluidic bioconversion technology is promising for modified functional lipids production allowing for cost-effective approach to produce high-value microalgal coproducts.« less

Selective synthesis of human milk fat-style structured triglycerides from microalgal oil in a microfluidic reactor packed with immobilized lipase

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

Wang, Jun; Liu, Xi; Wang, Xu -Dong

Human milk fat-style structured triacylglycerols were produced from microalgal oil in a continuous microfluidic reactor packed with immobilized lipase for the first time. A remarkably high conversion efficiency was demonstrated in the microreactor with reaction time being reduced by 8 times, Michaelis constant decreased 10 times, the lipase reuse times increased 2.25-fold compared to those in a batch reactor. In addition, the content of palmitic acid at sn-2 position (89.0%) and polyunsaturated fatty acids at sn-1, 3 positions (81.3%) are slightly improved compared to the product in a batch reactor. The increase of melting points (1.7 °C) and decrease ofmore » crystallizing point (3 °C) implied higher quality product was produced using the microfluidic technology. The main cost can be reduced from 212.3 to 14.6 per batch with the microreactor. Altogether, the microfluidic bioconversion technology is promising for modified functional lipids production allowing for cost-effective approach to produce high-value microalgal coproducts.« less

Selective synthesis of human milk fat-style structured triglycerides from microalgal oil in a microfluidic reactor packed with immobilized lipase.

PubMed

Wang, Jun; Liu, Xi; Wang, Xu-Dong; Dong, Tao; Zhao, Xing-Yu; Zhu, Dan; Mei, Yi-Yuan; Wu, Guo-Hua

2016-11-01

Human milk fat-style structured triacylglycerols were produced from microalgal oil in a continuous microfluidic reactor packed with immobilized lipase for the first time. A remarkably high conversion efficiency was demonstrated in the microreactor with reaction time being reduced by 8 times, Michaelis constant decreased 10 times, the lipase reuse times increased 2.25-fold compared to those in a batch reactor. In addition, the content of palmitic acid at sn-2 position (89.0%) and polyunsaturated fatty acids at sn-1, 3 positions (81.3%) are slightly improved compared to the product in a batch reactor. The increase of melting points (1.7°C) and decrease of crystallizing point (3°C) implied higher quality product was produced using the microfluidic technology. The main cost can be reduced from $212.3 to $14.6 per batch with the microreactor. Overall, the microfluidic bioconversion technology is promising for modified functional lipids production allowing for cost-effective approach to produce high-value microalgal coproducts. Copyright © 2016 Elsevier Ltd. All rights reserved.

The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

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

Ariani, Menik; Su'ud, Zaki; Waris, Abdul

2012-06-06

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it ismore » shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

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

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Removal of human pathogenic viruses in a down-flow hanging sponge (DHS) reactor treating municipal wastewater and health risks associated with utilization of the effluent for agricultural irrigation.

PubMed

Kobayashi, Naohiro; Oshiki, Mamoru; Ito, Toshihiro; Segawa, Takahiro; Hatamoto, Masashi; Kato, Tsuyoshi; Yamaguchi, Takashi; Kubota, Kengo; Takahashi, Masanobu; Iguchi, Akinori; Tagawa, Tadashi; Okubo, Tsutomu; Uemura, Shigeki; Harada, Hideki; Motoyama, Toshiki; Araki, Nobuo; Sano, Daisuke

2017-03-01

A down-flow hanging sponge (DHS) reactor has been developed as a cost-effective wastewater treatment system that is adaptable to local conditions in low-income countries. A pilot-scale DHS reactor previously demonstrated stable reduction efficiencies for chemical oxygen demand (COD) and ammonium nitrogen over a year at ambient temperature, but the pathogen reduction efficiency of the DHS reactor has yet to be investigated. In the present study, the reduction efficiency of a pilot-scale DHS reactor fed with municipal wastewater was investigated for 10 types of human pathogenic viruses (norovirus GI, GII and GIV, aichivirus, astrovirus, enterovirus, hepatitis A and E viruses, rotavirus, and sapovirus). DHS influent and effluent were collected weekly or biweekly for 337 days, and concentrations of viral genomes were determined by microfluidic quantitative PCR. Aichivirus, norovirus GI and GII, enterovirus, and sapovirus were frequently detected in DHS influent, and the log 10 reduction (LR) of these viruses ranged from 1.5 to 3.7. The LR values for aichivirus and norovirus GII were also calculated using a Bayesian estimation model, and the average LR (±standard deviation) values for aichivirus and norovirus GII were estimated to be 1.4 (±1.5) and 1.8 (±2.5), respectively. Quantitative microbial risk assessment was conducted to calculate a threshold reduction level for norovirus GII that would be required for the use of DHS effluent for agricultural irrigation, and it was found that LRs of 2.6 and 3.7 for norovirus GII in the DHS effluent were required in order to not exceed the tolerable burden of disease at 10 -4 and 10 -6 disability-adjusted life years loss per person per year, respectively, for 95% of the exposed population during wastewater reuse for irrigation. Copyright © 2016 Elsevier Ltd. All rights reserved.

Development of a New Transportation/Storage Cask System for Use by the DOE Russian Research Reactor Fuel Return Program

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

Michael J. Tyacke; Frantisek Svitak; Jiri Rychecky

2007-10-01

The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions at these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design, licensing,more » testing, and delivery of this new cask system result from a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: 1) Introduction; 2) VPVR/M Cask Description; 3) Ancillary Equipment, 4) Cask Licensing; 5) Cask Demonstration and Operations; 6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, 7) Conclusions.« less

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

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

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

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

Purification and Chemical Control of Molten Li2BeF 4 for a Fluoride Salt Cooled Reactor

NASA Astrophysics Data System (ADS)

Kelleher, Brian Christopher

Out of the many proposed generation IV, high-temperature reactors, the molten salt reactor (MSR) is one of the most promising. The first large scale MSR, the molten salt reactor experiment (MSRE), operated from 1965 to 1969 using Li2BeF4, or flibe, as a coolant and solvent for uranium fluoride fuel, at maximum temperatures of 654°C, for over 15000 hours. The MSRE experienced no concept breaking surprises and was considered a success. Newly proposed designs of molten salt reactors use solid fuels, making them less exotic compared to the MSRE. However, any molten salt reactor will require a great deal of research pertaining to the chemical and mechanical mastery of molten salts in order to prepare it for commercialization. To supplement the development of new molten salt reactors, approximately 100 kg of flibe was purified using the standard hydrofluorination process. Roughly half of the purified salt was lithium-7 enriched salt from the secondary loop of the MSRE. Purification rids the salt of impurities and reduces its capacity for corrosion, also known as the redox potential. The redox potential of flibe was measured at various stages of purification for the first time using a dynamic beryllium reference electrode. These redox measurements have been superimposed with metal impurities measurements found by neutron activation analysis. Lastly, reductions of flibe with beryllium metal have been investigated. Over reductions have been performed, which have shown to decrease redox potential while seemingly creating a beryllium-beryllium halide system. Recommendations of the lowest advisable redox potential for corrosion tests are included along with suggestions for future work.

Nitrogen source effects on the denitrifying anaerobic methane oxidation culture and anaerobic ammonium oxidation bacteria enrichment process.

PubMed

Fu, Liang; Ding, Jing; Lu, Yong-Ze; Ding, Zhao-Wei; Zeng, Raymond J

2017-05-01

The co-culture system of denitrifying anaerobic methane oxidation (DAMO) and anaerobic ammonium oxidation (Anammox) has a potential application in wastewater treatment plant. This study explored the effects of permutation and combination of nitrate, nitrite, and ammonium on the culture enrichment from freshwater sediments. The co-existence of NO 3 - , NO 2 - , and NH 4 + shortened the enrichment time from 75 to 30 days and achieved a total nitrogen removal rate of 106.5 mg/L/day on day 132. Even though ammonium addition led to Anammox bacteria increase and a higher nitrogen removal rate, DAMO bacteria still dominated in different reactors with the highest proportion of 64.7% and the maximum abundance was 3.07 ± 0.25 × 10 8 copies/L (increased by five orders of magnitude) in the nitrite reactor. DAMO bacteria showed greater diversity in the nitrate reactor, and one was similar to M. oxyfera; DAMO bacteria in the nitrite reactor were relatively unified and similar to M. sinica. Interestingly, no DAMO archaea were found in the nitrate reactor. This study will improve the understanding of the impact of nitrogen source on DAMO and Anammox co-culture enrichment.

Modelling the radiolysis of RSG-GAS primary cooling water

NASA Astrophysics Data System (ADS)

Butarbutar, S. L.; Kusumastuti, R.; Subekti, M.; Sunaryo, G. R.

2018-02-01

Water chemistry control for light water coolant reactor required a reliable understanding of radiolysis effect in mitigating corrosion and degradation of reactor structure material. It is known that oxidator products can promote the corrosion, cracking and hydrogen pickup both in the core and in the associated piping components of the reactor. The objective of this work is to provide the radiolysis model of RSG GAS cooling water and further more to predict the oxidator concentration which can lead to corrosion of reactor material. Direct observations or measurements of the chemistry in and around the high-flux core region of a nuclear reactor are difficult due to the extreme conditions of high temperature, pressure, and mixed radiation fields. For this reason, chemical models and computer simulations of the radiolysis of water under these conditions are an important route of investigation. FACSIMILE were used to calculate the concentration of O2 formed at relatively long-time by the pure water γ and neutron irradiation (pH=7) at temperature between 25 and 50 °C. This simulation method is based on a complex chemical reaction kinetic. In this present work, 300 MeV-proton were used to mimic γ-rays radiolysis and 2 MeV fast neutrons. Concentration of O2 were calculated at 10-6 - 106 s time scale.

7. Another picture of workers laying up the graphite core ...

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

7. Another picture of workers laying up the graphite core of the 105-B pile. This view is towards the rear of the pile. The gun barrels can be seen protruding into the pile. D-3047 - B Reactor, Richland, Benton County, WA

Development of New Transportation/Storage Cask System for Use by DOE Russian Research Reactor Fuel Return Program

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

Michael Tyacke; Frantisek Svitak; Jiri Rychecky

2010-04-01

The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian supplied high-enriched uranium (HEU) fuel currently stored at Russian-designed research reactors throughout the world to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions for these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design,more » licensing, testing, and delivery of this new cask system are the results of a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: (1) Introduction/Background; (2) VPVR/M Cask Description; (3) Ancillary Equipment, (4) Cask Licensing; (5) Cask Demonstration and Operations; (6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, (7) Summary and Conclusions.« less

Feasibility of Rectangular Concrete Pressure Vessels for Human Occupancy

DTIC Science & Technology

1990-07-01

incorporated: into the s 1.71 ! -1 rule-. 20 DISTRIBUTION/ AVAILABILIT ’- ,"r ABSTRACT 2’ ABSTRACT SECURITY CLASSIFICATION 7 JUNCLASSIFIED/UNLIMITED 0 SAME AS RPT...carried the end loads. Gas cooled reactors were never very popular in the US. Domestic utilities preferred boiling water reactors that operated at...Point Tower); 1975 - 9,000 psi in Chicago ( Water Tower Place); 1984 - 10,000 psi in Seattle (Century Square Bldg.); 1987 - 10,000 psi in Toronto

Microstructural characterization of a thin film ZrN diffusion barrier in an As-fabricated U-7Mo/Al matrix dispersion fuel plate

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Perez, Emmanuel; Wiencek, Tom; Leenaers, Ann; Van den Berghe, Sven

2015-03-01

The United States High Performance Research Reactor Fuel Development program is developing low enriched uranium fuels for application in research and test reactors. One concept utilizes U-7 wt.% Mo (U-7Mo) fuel particles dispersed in Al matrix, where the fuel particles are coated with a 1 μm-thick ZrN coating. The ZrN serves as a diffusion barrier to eliminate a deleterious reaction that can occur between U-7Mo and Al when a dispersion fuel is irradiated under aggressive reactor conditions. To investigate the final microstructure of a physically-vapor-deposited ZrN coating in a dispersion fuel plate after it was fabricated using a rolling process, characterization samples were taken from a fuel plate that was fabricated at 500 °C using ZrN-coated U-7Mo particles, Al matrix and AA6061 cladding. Scanning electron and transmission electron microscopy analysis were performed. Data from these analyses will be used to support future microstructural examinations of irradiated fuel plates, in terms of understanding the effects of irradiation on the ZrN microstructure, and to determine the role of diffusion barrier microstructure in eliminating fuel/matrix interactions during irradiation. The as-fabricated coating was determined to be cubic-ZrN (cF8) phase. It exhibited a columnar microstructure comprised of nanometer-sized grains and a region of relatively high porosity, mainly near the Al matrix. Small impurity-containing phases were observed at the U-7Mo/ZrN interface, and no interaction zone was observed at the ZrN/Al interface. The bonding between the U-7Mo and ZrN appeared to be mechanical in nature. A relatively high level of oxygen was observed in the ZrN coating, extending from the Al matrix in the ZrN coating in decreasing concentration. The above microstructural characteristics are discussed in terms of what may be most optimal for a diffusion barrier in a dispersion fuel plate application.

Improved hydrocracker temperature control: Mobil quench zone technology

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

Sarli, M.S.; McGovern, S.J.; Lewis, D.W.

1993-01-01

Hydrocracking is a well established process in the oil refining industry. There are over 2.7 million barrels of installed capacity world-wide. The hydrocracking process comprises several families of highly exothermic reactions and the total adiabatic temperature rise can easily exceed 200 F. Reactor temperature control is therefore very important. Hydrocracking reactors are typically constructed with multiple catalyst beds in series. Cold recycle gas is usually injected between the catalyst beds to quench the reactions, thereby controlling overall temperature rise. The design of this quench zone is the key to good reactor temperature control, particularly when processing poorer quality, i.e., highermore » heat release, feeds. Mobil Research and Development Corporation (MRDC) has developed a robust and very effective quench zone technology (QZT) package, which is now being licensed to the industry for hydrocracking applications.« less

Aggregate Size and Architecture Determine Microbial Activity Balance for One-Stage Partial Nitritation and Anammox ▿

PubMed Central

Vlaeminck, Siegfried E.; Terada, Akihiko; Smets, Barth F.; De Clippeleir, Haydée; Schaubroeck, Thomas; Bolca, Selin; Demeestere, Lien; Mast, Jan; Boon, Nico; Carballa, Marta; Verstraete, Willy

2010-01-01

Aerobic ammonium-oxidizing bacteria (AerAOB) and anoxic ammonium-oxidizing bacteria (AnAOB) cooperate in partial nitritation/anammox systems to remove ammonium from wastewater. In this process, large granular microbial aggregates enhance the performance, but little is known about granulation so far. In this study, three suspended-growth oxygen-limited autotrophic nitrification-denitrification (OLAND) reactors with different inoculation and operation (mixing and aeration) conditions, designated reactors A, B, and C, were used. The test objectives were (i) to quantify the AerAOB and AnAOB abundance and the activity balance for the different aggregate sizes and (ii) to relate aggregate morphology, size distribution, and architecture putatively to the inoculation and operation of the three reactors. A nitrite accumulation rate ratio (NARR) was defined as the net aerobic nitrite production rate divided by the anoxic nitrite consumption rate. The smallest reactor A, B, and C aggregates were nitrite sources (NARR, >1.7). Large reactor A and C aggregates were granules capable of autonomous nitrogen removal (NARR, 0.6 to 1.1) with internal AnAOB zones surrounded by an AerAOB rim. Around 50% of the autotrophic space in these granules consisted of AerAOB- and AnAOB-specific extracellular polymeric substances. Large reactor B aggregates were thin film-like nitrite sinks (NARR, <0.5) in which AnAOB were not shielded by an AerAOB layer. Voids and channels occupied 13 to 17% of the anoxic zone of AnAOB-rich aggregates (reactors B and C). The hypothesized granulation pathways include granule replication by division and budding and are driven by growth and/or decay based on species-specific physiology and by hydrodynamic shear and mixing. PMID:19948857

Enrichment of specific electro-active microorganisms and enhancement of methane production by adding granular activated carbon in anaerobic reactors.

PubMed

Lee, Jung-Yeol; Lee, Sang-Hoon; Park, Hee-Deung

2016-04-01

Direct interspecies electron transfer (DIET) via conductive materials can provide significant benefits to anaerobic methane formation in terms of production amount and rate. Although granular activated carbon (GAC) demonstrated its applicability in facilitating DIET in methanogenesis, DIET in continuous flow anaerobic reactors has not been verified. Here, evidences of DIET via GAC were explored. The reactor supplemented with GAC showed 1.8-fold higher methane production rate than that without GAC (35.7 versus 20.1±7.1mL-CH4/d). Around 34% of methane formation was attributed to the biomass attached to GAC. Pyrosequencing of 16S rRNA gene demonstrated the enrichment of exoelectrogens (e.g. Geobacter) and hydrogenotrophic methanogens (e.g. Methanospirillum and Methanolinea) from the biomass attached to GAC. Furthermore, anodic and cathodic currents generation was observed in an electrochemical cell containing GAC biomass. Taken together, GAC supplementation created an environment for enriching the microorganisms involved in DIET, which increased the methane production rate. Copyright © 2016 Elsevier Ltd. All rights reserved.

System for fuel rod removal from a reactor module

DOEpatents

Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

1988-07-28

A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

Characteristics of DO, organic matter, and ammonium profile for practical-scale DHS reactor under various organic load and temperature conditions.

PubMed

Nomoto, Naoki; Ali, Muntjeer; Jayaswal, Komal; Iguchi, Akinori; Hatamoto, Masashi; Okubo, Tsutomu; Takahashi, Masanobu; Kubota, Kengo; Tagawa, Tadashi; Uemura, Shigeki; Yamaguchi, Takashi; Harada, Hideki

2018-04-01

Profile analysis of the down-flow hanging sponge (DHS) reactor was conducted under various temperature and organic load conditions to understand the organic removal and nitrification process for sewage treatment. Under high organic load conditions (3.21-7.89 kg-COD m -3  day -1 ), dissolved oxygen (DO) on the upper layer of the reactor was affected by organic matter concentration and water temperature, and sometimes reaches around zero. Almost half of the COD Cr was removed by the first layer, which could be attributed to the adsorption of organic matter on sponge media. After the first layer, organic removal proceeded along the first-order reaction equation from the second to the fourth layers. The ammoniacal nitrogen removal ratio decreased under high organic matter concentration (above 100 mg L -1 ) and low DO (less than 1 mg L -1 ) condition. Ammoniacal nitrogen removal proceeded via a zero-order reaction equation along the reactor height. In addition, the profile results of DO, COD Cr , and NH 3 -N were different in the horizontal direction. Thus, it is thought the concentration of these items and microbial activities were not in a uniform state even in the same sponge layer of the DHS reactor.

Preparation of dilute magnetic semiconductor films by metalorganic chemical vapor deposition

NASA Technical Reports Server (NTRS)

Nouhi, Akbar (Inventor); Stirn, Richard J. (Inventor)

1988-01-01

A method for preparation of a dilute magnetic semiconductor (DMS) film is provided, in which a Group II metal source, a Group VI metal source and a transition metal magnetic ion source are pyrolyzed in the reactor of a metalorganic chemical vapor deposition (MOCVD) system by contact with a heated substrate. As an example, the preparation of films of Cd(sub 1-x)Mn(sub x)Te, in which 0 is less than or equal to x less than or equal to 0.7, on suitable substrates (e.g., GaAs) is described. As a source of manganese, tricarbonyl (methylcyclopentadienyl) manganese (TCPMn) is employed. To prevent TCPMn condensation during its introduction into the reactor, the gas lines, valves and reactor tubes are heated. A thin-film solar cell of n-i-p structure, in which the i-type layer comprises a DMS, is also described; the i-type layer is suitably prepared by MOCVD.

Preparation of dilute magnetic semiconductor films by metalorganic chemical vapor deposition

NASA Technical Reports Server (NTRS)

Nouhi, Akbar (Inventor); Stirn, Richard J. (Inventor)

1990-01-01

A method for preparation of a dilute magnetic semiconductor (DMS) film is provided, wherein a Group II metal source, a Group VI metal source and a transition metal magnetic ion source are pyrolyzed in the reactor of a metalorganic chemical vapor deposition (MOCVD) system by contact with a heated substrate. As an example, the preparation of films of Cd.sub.1-x Mn.sub.x Te, wherein 0.ltoreq..times..ltoreq.0.7, on suitable substrates (e.g., GaAs) is described. As a source of manganese, tricarbonyl (methylcyclopentadienyl) maganese (TCPMn) is employed. To prevent TCPMn condensation during the introduction thereof int the reactor, the gas lines, valves and reactor tubes are heated. A thin-film solar cell of n-i-p structure, wherein the i-type layer comprises a DMS, is also described; the i-type layer is suitably prepared by MOCVD.

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

Bily, T.

Thermoluminescent dosimeters represent very useful tool for gamma fields parameters measurements at nuclear research reactors, especially at zero power ones. {sup 7}LiF:Mg,Ti and {sup 7}LiF:Mg,Cu,P type TL dosimeters enable determination of only gamma component in mixed neutron - gamma field. At VR-1 reactor operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague the integral characteristics of gamma rays field were investigated, especially its spatial distribution and time behaviour, i.e. the non-saturated delayed gamma ray emission influence. Measured spatial distributions were compared with monte carlo code MCNP5 calculations. Although MCNP cannot generate delayedmore » gamma rays from fission, the relative gamma dose rate distribution is within {+-} 15% with measured values. The experiments were carried out with core configuration C1 consisting of LEU fuel IRT-4M (19.7 %). (author)« less

Characterization and treatment of Denizli landfill leachate using anaerobic hybrid/aerobic CSTR systems.

PubMed

Ağdağ, Osman Nuri

2011-01-01

Leachate generated in municipal solid waste landfill contains large amounts of organic and inorganic contaminants. In the scope of the study, characterization and anaerobic/aerobic treatability of leachate from Denizli (Turkey) Sanitary Landfill were investigated. Time-based fluctuations in characteristics of leachate were monitored during a one-year period. In characterization study; chemical oxygen demand (COD), biochemical oxygen demand (BOD) dissolved oxygen, temperature, pH, alkalinity, volatile fatty acids, total nitrogen, NH4-N, BOD5/COD ratio, suspended solid, inert COD, anaerobic toxicity assay and heavy metals concentrations in leachate were monitored. Average COD, BOD and NH4-N concentration in leachate were measured as 18034 mg/l, 11504 mg/l and 454 mg/l, respectively. Generally, pollution parameters in leachate were higher in summer and relatively lower in winter due to dilution by precipitation. For treatment of leachate, two different reactors, namely anaerobic hybrid and aerobic completely stirred tank reactor (CSTR) having effective volumes of 17.7 and 10.5 litres, respectively, were used. After 41 days of start-up period, leachate was loaded to hybrid reactor at 10 different organic loading rates (OLRs). OLR was increased by increasing COD concentrations. COD removal efficiency of hybrid reactor was carried out at a maximum of 91%. A percentage of 96% of residual COD was removed in the aerobic reactor. NH4-N removal rate in CSTR was quite high. In addition, high methane content was obtained as 64% in the hybrid reactor. At the end of the study, after 170 operation days, it can be said that the hybrid reactor and CSTR were very effective for leachate treatment.

Treatment of raw and ozonated oil sands process-affected water under decoupled denitrifying anoxic and nitrifying aerobic conditions: a comparative study.

PubMed

Xue, Jinkai; Zhang, Yanyan; Liu, Yang; Gamal El-Din, Mohamed

2016-11-01

Batch experiments were performed to evaluate biodegradation of raw and ozonated oil sands process-affected water (OSPW) under denitrifying anoxic and nitrifying aerobic conditions for 33 days. The results showed both the anoxic and aerobic conditions are effective in degrading OSPW classical and oxidized naphthenic acids (NAs) with the aerobic conditions demonstrating higher removal efficiency. The reactors under nitrifying aerobic condition reduced the total classical NAs of raw OSPW by 69.1 %, with better efficiency for species of higher hydrophobicity. Compared with conventional aerobic reactor, nitrifying aerobic condition substantially shortened the NA degradation half-life to 16 days. The mild-dose ozonation remarkably accelerated the subsequent aerobic biodegradation of classical NAs within the first 14 days, especially for those with long carbon chains. Moreover, the ozone pretreatment enhanced the biological removal of OSPW classical NAs by leaving a considerably lower final residual concentration of 10.4 mg/L under anoxic conditions, and 5.7 mg/L under aerobic conditions. The combination of ozonation and nitrifying aerobic biodegradation removed total classical NAs by 76.5 % and total oxy-NAs (O3-O6) by 23.6 %. 454 Pyrosequencing revealed that microbial species capable of degrading recalcitrant hydrocarbons were dominant in all reactors. The most abundant genus in the raw and ozonated anoxic reactors was Thauera (~56 % in the raw OSPW anoxic reactor, and ~65 % in the ozonated OSPW anoxic reactor); whereas Rhodanobacter (~40 %) and Pseudomonas (~40 %) dominated the raw and ozonated aerobic reactors, respectively. Therefore, the combination of mild-dose ozone pretreatment and subsequent biological process could be a competent choice for OSPW treatment.

Engineered heat treated methanogenic granules: a promising biotechnological approach for extreme thermophilic biohydrogen production.

PubMed

Abreu, Angela A; Alves, Joana I; Pereira, M Alcina; Karakashev, Dimitar; Alves, M Madalena; Angelidaki, Irini

2010-12-01

In the present study, two granular systems were compared in terms of hydrogen production rate, stability and bacterial diversity under extreme thermophilic conditions (70 degrees C). Two EGSB reactors were individually inoculated with heat treated methanogenic granules (HTG) and HTG amended with enrichment culture with high capacity of hydrogen production (engineered heat treated methanogenic granules - EHTG), respectively. The reactor inoculated with EHTG (R(EHTG)) attained a maximum production rate of 2.7l H(2)l(-1)day(-1) in steady state. In comparison, the R(HTG) containing the HTG granules was very unstable, with low hydrogen productions and only two peaks of hydrogen (0.8 and 1.5l H(2)l(-1)day(-1)). The presence of active hydrogen producers in the R(EHTG) system during the reactor start-up resulted in the development of an efficient H(2)-producing bacterial community. The results showed that "engineered inocula" where known hydrogen producers are co-inoculated with HTG is an efficient way to start up biohydrogen-producing reactors. Copyright (c) 2010 Elsevier Ltd. All rights reserved.

Degradation characteristics of polylactide in thermophilic anaerobic digestion with hyperthermophilic solubilization condition.

PubMed

Wang, F; Hidaka, T; Oishi, T; Osumi, S; Tsubota, J; Tsuno, H

2011-01-01

To test whether hyperthermophilic treatment promotes polylactide (PLA) dissolution and methane conversion under anaerobic digestion conditions, a single thermophilic control reactor (55 °C) and a two-phase system consisting of a hyperthermophilic reactor (80 °C) and a thermophilic reactor (55 °C) were continuously fed with a mixture of PLA and artificial kitchen garbage. In Runs 1 and 2, the PLA dissolution ratios in the two-phase system were 79.2 ± 6.5% and 85.2 ± 7.0%, respectively, higher than those of the control. Batch experimental results indicated that hyperthermophilic treatment could promote PLA dissolution to a greater degree as compared with single thermophilic treatment and that ammonia addition also had a promotional effect on PLA dissolution. In the two-phase system, after hyperthermophilic treatment, dissolved PLA was converted to methane gas under the subsequent thermophilic condition.

The Application of U-Np Fuel and {sup 6}Li Burnable Poison for Space Reactors

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

Nikitin, Konstantin L.; Saito, Masaki; Artisyuk, Vladimir V.

2003-11-15

The possible application of {sup 6}Li as a burnable poison and U-Np nitride as a fuel for space nuclear reactors has been studied. The analysis was performed for an infinite lattice with a leakage in the form of buckling and (U-Np)N fuel with 20% uranium enrichment. The combination of {sup 7}Li as a coolant and {sup 6}Li as a burnable poison results in a favorable criticality behavior during burnup. The parameters taken into consideration include the different fuel and coolant compositions, the form of absorber material, and the various absorber mass and concentrations. It was found that absorption properties ofmore » {sup 6}Li allow reaching the burnup value up to 67 GWd/tHM while reactivity swing is comparable with {beta}{sub eff}. The corresponding reactor lifetime is {approx}10 to 30 yr.« less

Radiological characterization of the pressure vessel internals of the BNL High Flux Beam Reactor.

PubMed

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

2004-08-01

In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, measurements and calculations of the decay gamma-ray dose-rate were performed in the reactor pressure vessel and on vessel internal structures such as the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. Measurements of gamma-ray dose rates were made using Red Perspex polymethyl methacrylate high-dose film, a Radcal "peanut" ion chamber, and Eberline's RO-7 high-range ion chamber. As a comparison, the Monte Carlo MCNP code and MicroShield code were used to model the gamma-ray transport and dose buildup. The gamma-ray dose rate at 8 cm above the center of the Transition Plate was measured to be 160 Gy h (using an RO-7) and 88 Gy h at 8 cm above and about 5 cm lateral to the Transition Plate (using Red Perspex film). This compares with a calculated dose rate of 172 Gy h using Micro-Shield. The gamma-ray dose rate was 16.2 Gy h measured at 76 cm from the reactor core (using the "peanut" ion chamber) and 16.3 Gy h at 87 cm from the core (using Red Perspex film). The similarity of dose rates measured with different instruments indicates that using different methods and instruments is acceptable if the measurement (and calculation) parameters are well defined. Different measurement techniques may be necessary due to constraints such as size restrictions.

Functionally gradient material for membrane reactors to convert methane gas into value-added products

DOEpatents

Balachandran, U.; Dusek, J.T.; Kleefisch, M.S.; Kobylinski, T.P.

1996-11-12

A functionally gradient material for a membrane reactor for converting methane gas into value-added-products includes an outer tube of perovskite, which contacts air; an inner tube which contacts methane gas, of zirconium oxide, and a bonding layer between the perovskite and zirconium oxide layers. The bonding layer has one or more layers of a mixture of perovskite and zirconium oxide, with the layers transitioning from an excess of perovskite to an excess of zirconium oxide. The transition layers match thermal expansion coefficients and other physical properties between the two different materials. 7 figs.

Removal of 2,4-dinitrophenol using hybrid methods based on ultrasound at an operating capacity of 7 L.

PubMed

Bagal, Manisha V; Lele, Bhagyashree J; Gogate, Parag R

2013-09-01

Sonochemical removal of 2,4-dinitrophenol (DNP) has been investigated using ultrasonic bath, with an operating capacity of 7 L, fitted with a large transducer with longitudinal vibrations having a 1 kW rated power output and operating frequency of 25 kHz. It has been revealed from calorimetric studies that maximum power is dissipated at a capacity of 7 L. The concentration of DNP has been monitored with an objective of evaluation of the efficacy of ultrasonic reactor in combination with process intensifying approaches for the removal of DNP. The effect of operating pH and additives such as hydrogen peroxide and ferrous iron activated persulfate on the extent of removal of DNP has been investigated. It has been observed that the extent of removal is greater at lower pH (pH 2.5 and 4) than at higher pH (pH 10). The combined treatment strategies such as ultrasound (US)/Fenton, US/advanced Fenton and US/CuO/H2O2 have also been investigated with an objective of obtaining complete removal of DNP using hybrid treatment strategies. The extent of removal has been found to increase significantly in US/Fenton process (98.7%) as compared to that using US alone (5.8%) which demonstrates the efficacy of the combined process. First order kinetics has been fitted for all the approaches investigated in the work. Calculations of cavitational yield indicated the superiority of the reactor design as compared to the conventional ultrasonic horn type reactors. The main intermediates formed during the process of removal of DNP have been identified. Copyright © 2013 Elsevier B.V. All rights reserved.

Indirect Liquefaction of Coal-Biomass Mixture for Production of Jet Fuel with High Productivity and Selectivity

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

Gangwal, Santosh K; McCabe, Kevin

Coal to liquids (CTL) and coal-biomass to liquids (CBTL) processes were advanced by testing and demonstrating Southern Research’s sulfur tolerant nickel-based reforming catalyst and Chevron’s highly selective and active cobalt-zeolite hybrid Fischer-Tropsch (FT) catalyst to clean, upgrade and convert syngas predominantly to jet fuel range hydrocarbon liquids, thereby minimizing expensive cleanup and wax upgrading operations. The National Carbon Capture Center (NCCC) operated by Southern Company (SC) at Wilsonville, Alabama served as the host site for the gasifier slip-stream and simulated syngas testing/demonstration. Reformer testing was performed to (1) reform tar and light hydrocarbons, (2) decompose ammonia in the presence H2S,more » and (3) deliver the required H2 to CO ratio for FT synthesis. FT Testing was performed to produce a product primarily containing C5-C20 liquid hydrocarbons and no C21+ waxy hydrocarbons with productivity greater than 0.7 gC5+/g catalyst/h, and at least 70% diesel and jet fuel range (C8-C20) hydrocarbon selectivity in the liquid product. A novel heat-exchange reactor system was employed to enable the use of the highly active FT catalyst and larger diameter reactors that results in cost reduction for commercial systems. Following laboratory development and testing, SR’s laboratory reformer was modified to operate in a Class 1 Div. 2 environment, installed at NCCC, and successfully tested for 125 hours using raw syngas. The catalyst demonstrated near equilibrium reforming (~90%) of methane and complete reforming/decomposition of tar and ammonia in the presence of up to 380 ppm H2S. For FT synthesis, SR modified and utilized a bench scale skid mounted FT reactor system (SR-CBTL test rig) that was fully integrated with a slip stream from SC/NCCC’s transport gasifier (TRIG). The test-rig developed in a previous project (DE-FE0010231) was modified to receive up to 7.5 lb/h raw syngas augmented with bottled syngas to adjust the H2/CO molar ratio to 2, clean it to cobalt FT catalyst specifications, and produce liquid FT products at the design capacity of up to 6 L/day. Promising Chevron catalyst candidates in the size range from 70-200 μm were loaded onto SR’s 2-inch ID and 4-inch ID bench-scale reactors utilizing IntraMicron’s micro-fiber entrapped catalyst (MFEC) heat exchange reactor technology. During 2 test campaigns, the FT reactors were successfully demonstrated at NCCC using syngas for ~420 hours. The catalyst did not experience deactivation during the tests. SR’s thermo-syphon heat removal system maintained reactor operating temperature along the axis to within ±4 °C. The experiments gave a steady catalyst productivity of 0.7-0.8 g/g catalyst/h, liquid hydrocarbon selectivity of ~75%, and diesel and jet fuel range hydrocarbon selectivity in the liquid product as high as 85% depending on process conditions. A preliminary techno-economic evaluation showed that the SR technology-based 50,000 bpd plant had a 10 % lower total plant cost compared to a conventional slurry reactor based plant. Furthermore, because of the modular nature of the SR technology, it was shown that the total plant cost advantage increases to >35 % as the plant is scaled down to 1000 bpd.« less

Sensitivity and Uncertainty Analysis of Plutonium and Cesium Isotopes in Modeling of BR3 Reactor Spent Fuel

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

Conant, Andrew; Erickson, Anna; Robel, Martin

Nuclear forensics has a broad task to characterize recovered nuclear or radiological material and interpret the results of investigation. One approach to isotopic characterization of nuclear material obtained from a reactor is to chemically separate and perform isotopic measurements on the sample and verify the results with modeling of the sample history, for example, operation of a nuclear reactor. The major actinide plutonium and fission product cesium are commonly measured signatures of the fuel history in a reactor core. This study investigates the uncertainty of the plutonium and cesium isotope ratios of a fuel rod discharged from a research pressurizedmore » water reactor when the location of the sample is not known a priori. A sensitivity analysis showed overpredicted values for the 240Pu/ 239Pu ratio toward the axial center of the rod and revealed a lower probability of the rod of interest (ROI) being on the periphery of the assembly. The uncertainty analysis found the relative errors due to only the rod position and boron concentration to be 17% to 36% and 7% to 15% for the 240Pu/ 239Pu and 137Cs/ 135Cs ratios, respectively. Lastly, this study provides a method for uncertainty quantification of isotope concentrations due to the location of the ROI. Similar analyses can be performed to verify future chemical and isotopic analyses.« less

Sensitivity and Uncertainty Analysis of Plutonium and Cesium Isotopes in Modeling of BR3 Reactor Spent Fuel

DOE PAGES

Conant, Andrew; Erickson, Anna; Robel, Martin; ...

2017-02-03

Nuclear forensics has a broad task to characterize recovered nuclear or radiological material and interpret the results of investigation. One approach to isotopic characterization of nuclear material obtained from a reactor is to chemically separate and perform isotopic measurements on the sample and verify the results with modeling of the sample history, for example, operation of a nuclear reactor. The major actinide plutonium and fission product cesium are commonly measured signatures of the fuel history in a reactor core. This study investigates the uncertainty of the plutonium and cesium isotope ratios of a fuel rod discharged from a research pressurizedmore » water reactor when the location of the sample is not known a priori. A sensitivity analysis showed overpredicted values for the 240Pu/ 239Pu ratio toward the axial center of the rod and revealed a lower probability of the rod of interest (ROI) being on the periphery of the assembly. The uncertainty analysis found the relative errors due to only the rod position and boron concentration to be 17% to 36% and 7% to 15% for the 240Pu/ 239Pu and 137Cs/ 135Cs ratios, respectively. Lastly, this study provides a method for uncertainty quantification of isotope concentrations due to the location of the ROI. Similar analyses can be performed to verify future chemical and isotopic analyses.« less

Autism and heritable bone fragility: A true association?

PubMed

Balasubramanian, Meena; Jones, Rebecca; Milne, Elizabeth; Marshall, Charlotte; Arundel, Paul; Smith, Kath; Bishop, Nicholas J

2018-06-01

Osteogenesis Imperfecta (OI) is a heterogeneous condition mainly characterised by bone fragility; intelligence is reported to be normal. However, a minority of children seen also show symptomology consistent with an 'Autism Spectrum Disorder'. A joint genetics and psychology research study was undertaken to identify these patients using 'Gold Standard' research tools: Autism Diagnostic Inventory Revised (ADI-R); Autism Diagnostic Observation Schedule (ADOS) and undertake genetic analyses in them. A cohort of n  = 7 children with autistic traits and severe/complex OI were recruited to the study. The study was set-up to explore whether there was a genetic link between bone fragility and autism in a sub-set of patients with bone fragility identified with autism traits in our complex/severe OI clinic. This was not set-up as a prevalence study but rather an exploration of genetics in association with ADI/ADOS confirmed ASD and bone fragility. Standardised tools were used to confirm autism diagnosis. ADI and ADOS were completed by the Clinical Psychologist; ADI comprises a 93 item semi-structured clinical review with a diagnostic algorithm diagnosing Autism; ADOS is a semi-structured assessment of socialisation, communication and play/imagination which also provides a diagnostic algorithm. In patients recruited, those that fulfilled research criteria for diagnosis of autism using above tools were recruited to trio whole exome sequencing (WES). one patient had compound heterozygous variants in NBAS ; one patient had a variant in NRX1 ; one patient had a maternally inherited PLS3 variant; all the other patients in this cohort had pathogenic variants in COL1A1/COL1A2 . Although, not set out as an objective, we were able to establish that identifying autism had important clinical and social benefits for patients and their families in ensuring access to services, appropriate schooling, increased understanding of behaviour and support. It is important for clinicians looking after children with brittle bone disease, also referred to as Osteogenesis Imperfecta (OI) to be aware of early features of developmental delay/autistic traits especially with severe forms of OI as the emphasis is on their mobility and bone health. Ensuring appropriate assessment and access to services early-on will enable these patients to achieve their potential. Further investigations of genomics in bone fragility in relation to autism are required and dual diagnosis is essential for high quality clinical and educational provision.

A microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1985-02-01

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. To improve the analytical extrapolation of test results to full-size assembly bundles, the facility upgrade will increase the maximum size of the test bundle from 7 to 37 fuel pins. By creating a core convertor zone around the test location, the neutron spectrum incident on the test assembly will be hardened and the maximum energy deposited in the sample will be increased. In addition, a programmable Automated Reactor Control System (ARCS) willmore » 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 system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations. A quantitative reliability analysis of the RTS shows that the unreliability, that is, the probability of failure, is acceptable for a 10 hour mission time or risk interval.« less

Development of a carbon formation reactor for carbon dioxide reduction

NASA Technical Reports Server (NTRS)

Noyes, G.

1985-01-01

Applied research, engineering development, and performance evaluation were conducted on a process for formation of dense carbon by pyrolysis of methane. Experimental research showed that dense (0.7 to 1.6 g/cc bulk density and 1.6 to 2.2 g/cc solid density) carbon can be produced by methane pyrolysis in quartzwool-packed quartz tubes at temperatrues of 1100 to 1300 C. This result supports the condensation theory of pyrolytic carbon formation from gaseous hydrocarbons. A full-scale Breadboard Carbon Formation Reactor (CFR) was designed, fabricated, and tested at 1100 to 1200 C with 380 to 2280 sccm input flows of methane. Single-pass conversion of methane to carbon ranged from 60 to 100 percent, with 89 percent average conversion. Performance was projected for an Advanced Carbon Reactor Subsystem (ACRS) which indicated that the ACRS is a viable option for management of metabolic carbon on long-duration space missions.

Technical assumption for Mo-99 production in the MARIA reactor. Feasibility study

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

Jaroszewicz, J.; Pytel, K.; Dabkowski, L.

2008-07-15

The main objective of U-235 irradiation is to obtain the Tc-99m isotope which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short life time, is a reaction of radioactive decay of Mo-99 into Tc- 99m. One of the possible sources of molybdenum can be achieved in course of the U-235 fission reaction. The paper presents activities and the calculations results obtained upon the feasibility study on irradiation of U-235 targets for production of molybdenum in the MARIA reactor. The activities including technical assumption were focused on performing calculation for modelling ofmore » the target and irradiation device as well as adequate equipment and tools for processing in reactor. It has been assumed that the basic component of fuel charge is an aluminium cladded plate with dimensions of 40x230x1.45 containing 4.7 g U-235. The presumed mode of the heat removal generated in the fuel charge of the reactor primary cooling circuit influences the construction of installation to be used for irradiation and the technological instrumentation. The outer channel construction for irradiation has to be identical as the standard fuel channel construction of the MARIA reactor. It enables to use the existing slab and reactor mounting sockets for the fastening of the molybdenum channel as well as the cooling water delivery system. The measurement of water temperature cooling a fuel charge and control of water flow rate in the channel can also be carried out be means of the standard instrumentation of the reactor. (author)« less

Acidogenic fermentation of iron-enhanced primary sedimentation sludge under different pH conditions for production of volatile fatty acids.

PubMed

Lin, Lin; Li, Xiao-Yan

2018-03-01

Iron-based chemically enhanced primary sedimentation (CEPS) is increasingly adopted for wastewater treatment in mega cities, producing a large amount of sludge (Fe-sludge) with a high content of organics for potential organic resource recovery. In this experimental study, acidogenic fermentation was applied treat FeCl 3 -based CEPS sludge for production of volatile fatty acids (VFAs) at different pHs. Batch fermentation tests on the Fe-sludge with an organic content of 10 g-COD/L showed that the maximum VFAs production reached 2782.2 mg-COD/L in the reactor without pH control, and it reached 688.4, 3095.3, and 2603.7 mg-COD/L in reactors with pHs kept at 5.0, 6.0 and 8.0, respectively. Analysis of the acidogenesis kinetics and enzymatic activity indicated that the alkaline pH could accelerate the rate of organic hydrolysis but inhibited the further organic conversion to VFAs. In semi-continuous sludge fermentation tests, the VFAs yield in the pH6 reactor was 20% higher than that in the control reactor without pH regulation, while the VFAs yield in the pH8 reactor was 10% lower than the control. Illumina MiSeq sequencing revealed that key functional microorganisms known for effective sludge fermentation, including Bacteroidia and Erysipelotrichi, were enriched in the pH6 reactor with an enhanced VFAs production, while Clostridia became more abundant in the pH8 reactor to stand the unfavorable pH condition. The research presented acidogenic fermentation as an effective process for CEPS sludge treatment and organic resource recovery and provided the first insight into the related microbial community dynamics. Copyright © 2017 Elsevier Ltd. All rights reserved.

Reduction of precursors of chlorination by-products in drinking water using fluidized-bed biofilm reactor at low temperature.

PubMed

Xie, Shu-Guang; Wen, Dong-Hui; Shi, Dong-Wen; Tang, Xiao-Yan

2006-10-01

To investigate the reduction of chlorination by-products (CBPs) precursors using the fluidized-bed biofilm reactor (FBBR). Reduction of total organic carbon (TOC), ultraviolet absorbance (UV254), trihalomethane (THM) formation potential (THMFP), haloacetic acid (HAA) formation potential (HAAFP), and ammonia in FBBR were evaluated in detail. Results The reduction of TOC or UV254 was low, on average 12.6% and 4.7%, respectively, while the reduction of THMFP and HAAFP was significant. The reduction of ammonia was 30%-40% even below 3 degrees C, however, it could quickly rise to over 50% above 3degrees C. Conclusions The FBBR effectively reduces CBPs and ammonia in drinking water even at low temperature and seems to be a very promising and competitive drinking water reactor for polluted surface source waters, especially in China.

Continuous reduction of tellurite to recoverable tellurium nanoparticles using an upflow anaerobic sludge bed (UASB) reactor.

PubMed

Ramos-Ruiz, Adriana; Sesma-Martin, Juan; Sierra-Alvarez, Reyes; Field, Jim A

2017-01-01

According to the U.S. Department of Energy and the European Union, tellurium is a critical element needed for energy and defense technology. Thus methods are needed to recover tellurium from waste streams. The objectives of this study was to determine the feasibility of utilizing upflow anaerobic sludge bed (UASB) reactors to convert toxic tellurite (Te IV ) oxyanions to non-toxic insoluble elemental tellurium (Te 0 ) nanoparticles (NP) that are amendable to separation from aqueous effluents. The reactors were supplied with ethanol as the electron donating substrate to promote the biological reduction of Te IV . One reactor was additionally amended with the redox mediating flavonoid compound, riboflavin (RF), with the goal of enhancing the bioreduction of Te IV . Its performance was compared to a control reactor lacking RF. The continuous formation of Te 0 NPs using the UASB reactors was found to be feasible and remarkably improved by the addition of RF. The presence of this flavonoid was previously shown to enhance the conversion rate of Te IV by approximately 11-fold. In this study, we demonstrated that this was associated with the added benefit of reducing the toxic impact of Te IV towards the methanogenic consortium in the UASB and thus enabled a 4.7-fold higher conversion rate of the chemical oxygen demand. Taken as a whole, this work demonstrates the potential of a methanogenic granular sludge to be applied as a bioreactor technology producing recoverable Te 0 NPs in a continuous fashion. Copyright © 2016 Elsevier Ltd. All rights reserved.

Dismantling the nuclear research reactor Thetis

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

Michiels, P.

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

Chernobyl Studies Project: Working group 7.0, Environmental transport and health effects. Progress report, March--September 1994

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

Anspaugh, L.R.; Hendrickson, S.M.

1994-12-01

In April 1988, the US and the former-USSR signed a Memorandum of Cooperation (MOC) for Civilian Nuclear Reactor Safety; this MOC was a direct result of the accident at the Chernobyl Nuclear Power Plant Unit 4 and the following efforts by the two countries to implement a joint program to improve the safety of nuclear power plants and to understand the implications of environmental releases. A Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS) was formed to implement the MOC. The JCCCNRS established many working groups; most of these were the responsibility of the Nuclear Regulatory Commission, as farmore » as the US participation was concerned. The lone exception was Working Group 7 on Environmental Transport and Health Effects, for which the US participation was the responsibility of the US Department of Energy (DOE). The purpose of Working Group 7 was succintly stated to be, ``To develop jointly methods to project rapidly the health effects of any future nuclear reactor accident.`` To implement the work DOE then formed two subworking groups: 7.1 to address Environmental Transport and 7.2 to address Health Effects. Thus, the DOE-funded Chernobyl Studies Project began. The majority of the initial tasks for this project are completed or near completion. The focus is now turned to the issue of health effects from the Chernobyl accident. Currently, we are involved in and making progress on the case-control and co-hort studies of thyroid diseases among Belarussian children. Dosimetric aspects are a fundamental part of these studies. We are currently working to implement similar studies in Ukraine. A major part of the effort of these projects is supporting these studies, both by providing methods and applications of dose reconstruction and by providing support and equipment for the medical teams.« less

Focused technology: Nuclear propulsion

NASA Technical Reports Server (NTRS)

Miller, Thomas J.

1991-01-01

The topics presented are covered in viewgraph form and include: nuclear thermal propulsion (NTP), which challenges (1) high temperature fuel and materials, (2) hot hydrogen environment, (3) test facilities, (4) safety, (5) environmental impact compliance, and (6) concept development, and nuclear electric propulsion (NEP), which challenges (1) long operational lifetime, (2) high temperature reactors, turbines, and radiators, (3) high fuel burn-up reactor fuels, and designs, (4) efficient, high temperature power conditioning, (5) high efficiency, and long life thrusters, (6) safety, (7) environmental impact compliance, and (8) concept development.

76 FR 59392 - Notice of Intent To Grant Exclusive Patent License; Enhanced Energy Group, LLC

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-26

... inventions, and they are covered by U.S. Patent No. 7,926,275: Closed Brayton Cycle Direct Contact Reactor/ Storage Tank With Chemical Scrubber.//U.S. Patent No. 7,926,276: Closed Cycle Brayton Propulsion System With Direct Heat Transfer.//U.S. Patent No. 7,937,930: Semiclosed Brayton Cycle Power System With...

76 FR 59391 - Notice of Availability of Government-Owned Inventions; Available for Licensing

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-26

... Issued 4/12/2011//U.S. Patent No. 7,926,275: Closed Brayton Cycle Direct Contact Reactor/Storage Tank... With Precursor Bubble Issued 4/19/2011//U.S. Patent No. 7,929,375: Method and Apparatus for Improved...: Hydrogen Generator Apparatus for an Underwater Vehicle Issued 5/10/2011//U.S. Patent No. 7,940,602...

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.

Effects of plastic composite support and pH profiles on pullulan production in a biofilm reactor.

PubMed

Cheng, Kuan-Chen; Demirci, Ali; Catchmark, Jeffrey M

2010-04-01

Pullulan is a linear homopolysaccharide which is composed of glucose units and often described as alpha-1, 6-linked maltotriose. The applications of pullulan range from usage as blood plasma substitutes to environmental pollution control agents. In this study, a biofilm reactor with plastic composite support (PCS) was evaluated for pullulan production using Aureobasidium pullulans. In test tube fermentations, PCS with soybean hulls, defatted soy bean flour, yeast extract, dried bovine red blood cells, and mineral salts was selected for biofilm reactor fermentation (due to its high nitrogen content, moderate nitrogen leaching rate, and high biomass attachment). Three pH profiles were later applied to evaluate their effects on pullulan production in a PCS biofilm reactor. The results demonstrated that when a constant pH at 5.0 was applied, the time course of pullulan production was advanced and the concentration of pullulan reached 32.9 g/L after 7-day cultivation, which is 1.8-fold higher than its respective suspension culture. The quality analysis demonstrated that the purity of produced pullulan was 95.8% and its viscosity was 2.4 centipoise. Fourier transform infrared spectroscopy spectra also supported the supposition that the produced exopolysaccharide was mostly pullulan. Overall, this study demonstrated that a biofilm reactor can be successfully implemented to enhance pullulan production and maintain its high purity.

Fundamental interactions involving neutrons and neutrinos: reactor-based studies led by Petersburg Nuclear Physics Institute (National Research Centre 'Kurchatov Institute') [PNPI (NRC KI)

NASA Astrophysics Data System (ADS)

Serebrov, A. P.

2015-11-01

Neutrons of very low energy ( ˜ 10-7 eV), commonly known as ultracold, are unique in that they can be stored in material and magnetic traps, thus enhancing methodical opportunities to conduct precision experiments and to probe the fundamentals of physics. One of the central problems of physics, of direct relevance to the formation of the Universe, is the violation of time invariance. Experiments searching for the nonzero neutron electric dipole moment serve as a time invariance test, and the use of ultracold neutrons provides very high measurement precision. Precision neutron lifetime measurements using ultracold neutrons are extremely important for checking ideas on the early formation of the Universe. This paper discusses problems that arise in studies using ultracold neutrons. Also discussed are the currently highly topical problem of sterile neutrinos and the search for reactor antineutrino oscillations at distances of 6-12 meters from the reactor core. The field reviewed is being investigated at multiple facilities globally. The present paper mainly concentrates on the results of PNPI-led studies at WWR-M PNPI (Gatchina), ILL (Grenoble), and SM-3 (Dimitrovgrad) reactors, and also covers the results obtained during preparation for research at the PIK reactor which is under construction.

Behavior of toxic metals and radionuclides during molten salt oxidation of chlorinated plastics.

PubMed

Yang, Hee-Chul; Cho, Yong-Jun; Eun, Hee-Chul; Yoo, Jae-Hyung; Kim, Joon-Hyung

2004-01-01

Molten salt oxidation is one of the promising alternatives to incineration for chlorinated organics without the emission of chlorinated organic pollutants. This study investigated the behavior of three hazardous metals (Cd, Pb, and Cr) and four radioactive metal surrogates (Cs, Ce, Gd, and Sm) in the molten Na2CO3 oxidation reactor during the destruction of PVC plastics. In the tested temperature ranges (1143 1223K) and NaCl content (0-10%), the impact of temperature on the retention of cadmium and lead in the molten salt reactor was very small, but that of the NaCl content for their retention was relatively higher. The influence of NaCl accumulation was, however, proven to be practically negligible due to the low-temperature operating characteristics of the molten salt oxidation system. Neither temperature increase nor chlorine accumulation in the MSO reactor reduced the retention of Cr, Ce, Gd, and Sm. Over 99.98% of these metals remained in the reactor. The influence of the temperature on the cesium behavior is relatively large for a chlorine addition, however, over 99.7% of cesium remained in the reactor throughout the entire test. The experimental metal entrainment rate and the entrained metal particle size distribution agree well with the theoretical equilibrium metal distributions.

Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K.; Ross Finlay, M.

2012-06-01

The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

Structural behavior of the Bitter plate tf magnet for the Zephyr ignition test reactor

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

Bobrov, E.S.; Becker, H.

1981-01-01

This paper discusses methods and results of the computer structural analysis of the Bitter plate toroidal field magnet design for the ZEPHYR Ignition Test Reactor. The magnet provides a field of 7.06 T at the center of the bore which is 1.76 m from the major toroidal axis. The ignited plasma is located at a major radius of 1.36 m where the magnetic field is 9.11 T. The plasma is moved to this final position following compression in the major radius. The horizontal bore of the magnet is 1.8 m.

The Sugar Model: Catalytic Flow Reactor Dynamics of Pyruvaldehyde Synthesis from Triose Catalyzed by Poly-L-Lysine Contained in a Dialyzer

NASA Technical Reports Server (NTRS)

Weber, Arthur L.; DeVincenzi, Donald (Technical Monitor)

2000-01-01

The formation of pyruvaldehyde from triose sugars was catalyzed by poly-L-lysine contained in a small dialyzer (100 MWCO) suspended in a much larger triose substrate reservoir. The polylysine confined in the dialyzer functioned as a catalytic flow reactor that constantly brought in triose from the substrate reservoir by diffusion to offset the drop in triose concentration within the reactor caused by its conversion to pyruvaldehyde. A 400 mM solution of poly-L-lysine contained in a 0.35 ml dialyzer placed in a 120 ml solution of triose substrate (pH 5.5, 40 C) generated pyruvaldehyde 11 -times faster than an a control reaction without the catalytic dialyzer. However, since the catalytic dialyzer's volume was 343-times smaller than the control reaction, the synthetic intensity (rate/volume) of pyruvaldehyde synthesis within the catalytic dialyzer was 3400-times greater than that of the control reaction and substrate solution. A similar result was obtained using a dialyzer with a 500 MWCO value. Acting as a catalytic flow reactor the polylysine catalytic dialyzer synthesized about 3.5 molecules of pyruvaldehyde per lysine residue in 7 days -- an amount of triose equal to twice the weight of the catalyst. At 7 days the catalytic activity of polylysine was 16% of its initial value, a result indicating catalyst-poisoning caused by reaction of pyruvaldehyde with the e-amino groups of polylysine. The dialyzer method of catalyst containment was selected it provides a simple, flexible, and easily manipulated experimental system for studying the dynamics and evolutionary development of confined autocatalytic processes related to the origin of life under anaerobic conditions.

The characteristics of extracellular polymeric substances and soluble microbial products in moving bed biofilm reactor-membrane bioreactor.

PubMed

Duan, Liang; Jiang, Wei; Song, Yonghui; Xia, Siqing; Hermanowicz, Slawomir W

2013-11-01

The characteristics of extracellular polymeric substances (EPS) and soluble microbial products (SMP) in conventional membrane bioreactor (MBR) and in moving bed biofilm reactor-membrane bioreactors (MBBR-MBR) were investigated in long-term (170 days) experiments. The results showed that all reactors had high removal efficiency of ammonium and COD, despite very different fouling conditions. The MBBR-MBR with media fill ratio of 26.7% had much lower total membrane resistance and no obvious fouling were detected during the whole operation. In contrast, MBR and MBBR-MBR with lower and higher media fill experienced more significant fouling. Low fouling at optimum fill ratio may be due to the higher percentage of small molecular size (<1 kDa) and lower percentage of large molecular size (>100 kDa) of EPS and SMP in the reactor. The composition of EPS and SMP affected fouling due to different O-H bonds in hydroxyl functional groups, and less polysaccharides and lipids. Copyright © 2013 Elsevier Ltd. All rights reserved.

Utilization of solid and liquid waste generated during ethanol fermentation process for production of gaseous fuel through anaerobic digestion--a zero waste approach.

PubMed

Narra, Madhuri; Balasubramanian, Velmurugan

2015-03-01

Preliminary investigations were performed in the laboratory using batch reactors at 10% solid concentration for the assessment of the biogas production at thermophilic and mesophilic temperatures using solid residues generated during ethanol fermentation process. One kg of solid residues (left after enzyme extraction and enzymatic hydrolysis) from thermophilic reactors (TR1 and TR2) produced around 131 and 84L of biogas, respectively, whereas biogas production from mesophilic reactors (MR1 and MR2) was 86 and 62L, respectively. After 20 and 35days of retention time, the TS and VS reductions from TR1, TR2 and MR1, MR2 were found to be 39.2% and 35.0%, 67.3% and 61.0%, 21.0% and 18.0%, 34.7% and 27.8%, respectively. Whereas the liquid waste was treated using four laboratory anaerobic hybrid reactors (AHRs) with two different natural and synthetic packing media at 15-3days HRTs. AHRs packed with natural media showed better COD removal efficiency and methane yield. Copyright © 2015 Elsevier Ltd. All rights reserved.

Performance of CSTR-EGSB-SBR system for treating sulfate-rich cellulosic ethanol wastewater and microbial community analysis.

PubMed

Shan, Lili; Zhang, Zhaohan; Yu, Yanling; Ambuchi, John Justo; Feng, Yujie

2017-06-01

Performance and microbial community composition were evaluated in a two-phase anaerobic and aerobic system treating sulfate-rich cellulosic ethanol wastewater (CEW). The system was operated at five different chemical oxygen demand (COD)/SO 4 2- ratios (63.8, 26.3, 17.8, 13.7, and 10.7). Stable performance was obtained for total COD removal efficiency (94.5%), sulfate removal (89.3%), and methane production rate (11.5 L/day) at an organic loading rate of 32.4 kg COD/(m 3 ·day). The acidogenic reactor made a positive contribution to net VFAs production (2318.1 mg/L) and sulfate removal (60.9%). Acidogenic bacteria (Megasphaera, Parabacteroides, unclassified Ruminococcaceae spp., and Prevotella) and sulfate-reducing bacteria (Butyrivibrio, Megasphaera) were rich in the acidogenic reactor. In the methanogenic reactor, high diversity of microorganisms corresponded with a COD removal contribution of 83.2%. Moreover, methanogens (Methanosaeta) were predominant, suggesting that these organisms played an important role in the acetotrophic methanogenesis pathway. The dominant aerobic bacteria (Truepera) appeared to have been responsible for the COD removal of the SBR. These results indicate that dividing the sulfate reduction process could effectively minimize sulfide toxicity, which is important for the successful operation of system treating sulfate-rich CEW.

Bacteria of the Candidate Phylum TM7 are Prevalent in Acidophilic Nitrifying Sequencing-Batch Reactors

PubMed Central

Hanada, Akiko; Kurogi, Takashi; Giang, Nguyen Minh; Yamada, Takeshi; Kamimoto, Yuki; Kiso, Yoshiaki; Hiraishi, Akira

2014-01-01

Laboratory-scale acidophilic nitrifying sequencing-batch reactors (ANSBRs) were constructed by seeding with sewage-activated sludge and cultivating with ammonium-containing acidic mineral medium (pH 4.0) with or without a trace amount of yeast extract. In every batch cycle, the pH varied between 2.7 and 4.0, and ammonium was completely converted to nitrate. Attempts to detect nitrifying functional genes in the fully acclimated ANSBRs by PCR with previously designed primers mostly gave negative results. 16S rRNA gene-targeted PCR and a subsequent denaturating gradient gel electrophoresis analysis revealed that a marked change occurred in the bacterial community during the overall period of operation, in which members of the candidate phylum TM7 and the class Gammaproteobacteria became predominant at the fully acclimated stage. This result was fully supported by a 16S rRNA gene clone library analysis, as the major phylogenetic groups of clones detected (>5% of the total) were TM7 (33%), Gammaproteobacteria (37%), Actinobacteria (10%), and Alphaproteobacteria (8%). Fluorescence in situ hybridization with specific probes also demonstrated the prevalence of TM7 bacteria and Gammaproteobacteria. These results suggest that previously unknown nitrifying microorganisms may play a major role in ANSBRs; however, the ecophysiological significance of the TM7 bacteria predominating in this process remains unclear. PMID:25241805

The effectiveness of using the combined-cycle technology in a nuclear power plant unit equipped with an SVBR-100 reactor

NASA Astrophysics Data System (ADS)

Kasilov, V. F.; Dudolin, A. A.; Gospodchenkov, I. V.

2015-05-01

The design of a modular SVBR-100 reactor with a lead-bismuth alloy liquid-metal coolant is described. The basic thermal circuit of a power unit built around the SVBR-100 reactor is presented together with the results of its calculation. The gross electrical efficiency of the turbine unit driven by saturated steam at a pressure of 6.7 MPa is estimated at η{el/gr} = 35.5%. Ways for improving the efficiency of this power unit and increasing its power output by applying gas-turbine and combined-cycle technologies are considered. With implementing a combined-cycle power-generating system comprising two GE-6101FA gas-turbine units with a total capacity of 140 MW, it becomes possible to obtain the efficiency of the combined-cycle plant equipped with the SVBR-100 reactor η{el/gr} = 45.39% and its electrical power output equal to 328 MW. The heat-recovery boiler used as part of this power installation generates superheated steam with a temperature of 560°C, due to which there is no need to use a moisture separator/steam reheater in the turbine unit thermal circuit.

Core design of a direct-cycle, supercritical-water-cooled fast breeder reactor

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

Jevremovic, T.; Oka, Yoshiaki; Koshizuka, Seiichi

1994-10-01

The conceptual design of a direct-cycle fast breeder reactor (FBR) core cooled by supercritical water is carried out as a step toward a low-cost FBR plant. The supercritical water does not exhibit change of phase. The turbines are directly driven by the core outlet coolant. In comparison with a boiling water reactor (BWR), the recirculation systems, steam separators, and dryers are eliminated. The reactor system is much simpler than the conventional steam-cooled FBRs, which adopted Loeffler boilers and complicated coolant loops for generating steam and separating it from water. Negative complete and partial coolant void reactivity are provided without muchmore » deterioration in the breeding performances by inserting thin zirconium-hydride layers between the seeds and blankets in a radially heterogeneous core. The net electric power is 1245 MW (electric). The estimated compound system doubling time is 25 yr. The discharge burnup is 77.7 GWd/t, and the refueling period is 15 months with a 73% load factor. The thermal efficiency is high (41.5%), an improvement of 24% relative to a BWR's. The pressure vessel is not thick at 30.3 cm.« less